ML20235L389
| ML20235L389 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 09/30/1987 |
| From: | Hunter D, Murphy M NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20235L248 | List: |
| References | |
| 50-313-87-14, 50-368-87-14, NUDOCS 8710050449 | |
| Download: ML20235L389 (58) | |
See also: IR 05000313/1987014
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APPENDIX C
U.S. NUCLEAR REGULATORY COMMISSION
REGION IV
NRC Inspection Report: 50-313/87-14 Licenses: DPR-51
50-368/87-14 NPF-6
Dockets: 50-313
50-368
Licensee: Arkansas Power & Light Company (AP&L) i
P. O. Box 551
Little Rock, Arkansas 72203
Facility Name: Arkansas Nuclear One (ANO), Units 1 and 2
Inspection At: AN0 Site, Russellville, Arkansas
Inspection Conducted: May 4-8 an'd June 8-12, 1987
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Inspector: / 9y77
M'. E. Murphy, Projett Iffspector, Project Dlte /
Section B, Reactor Projects Branch j
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Accompanied
By: D. Kubicki, Office of Nuclear Reactor Regulation
G. Dick, Office of Nuclear Reactor Regulation
R. Lee, Office 01 Nuclear Reactor Regulation l
M. Villaran, Brookhaven National Laboratory t
K. Parkinson, Brookhaven National Laborstory I
Approved: [ w et
D. R. Hunter, Chief, Project Section B Date {
Reactor Projects Branch
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Inspection Summary i
Inspection Conducted May 4-8, 1987 (Report 50-313/87-14)
Areas Inspected: Nonroutine, announced inspection for implementation of and
compliance to the safe shutdown requirements of 10 CFR 50, Appendix R.
Results: Within the areas inspected, one violation was identified (failure to
properly protect structural steel which supports or is framed into fire
barriers, paragraph 3). Two deviations were identified (failure of fire alarm
8710050449 870930 ,
PDR ADOCK 05000313
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fire barriers in accordance with BTP-APCSB 9.5-1, Appendix A, . paragraph 3).
Three unresolved items are identified in paragraphs 3, 8.-a(5), and 10.b(1).
Inspection Conducted June 8-12, 1987 (Report 50-368/87-14)
Areas Inspected: Nonroutine, announced inspection for implementation of and. j
compliance to the safe shutdown requirements of.10 CFR 50, Appendix R. I
Results: Within the areas inspected, one violation was identified (failure to
properly protect structural steel which supports or is framed into fire
barriers, paragraph 3). Two deviations were identified (failure of fire alarm i
system to comply with NFPA Standard No. 720, 1975; and, failure to maintain 1
l fire barriers in accordance with BTP-APCSB 9.5-1, Appendix A, paragraph 3). l
Three unresolved items are identified in paragraphs 3, 8.a(5), and 10.b(1).-
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DETAILS ]
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1. Persons Contacted I
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- E. C. Ewing, General Manager, Plant Support l
- M. M. Coombs, Corporate Fire Protection Specialist !
- D. D. Snellings, Manager, Nuclear Programs. 1
- D. Williams, Senior Engineer J
l *D. Lomax, Team Leader, Appendix R Audit (Plant Licensing' Supervisor)-
- 0. Howard, Team Leader, Appendix R' Audit (Manager, Special Projects)
D. H. Smith, Senior Engineering Technician )
H. Rideout, Engineer {
- J. G.'Dobbs, LR. Electrical Engineering Supervisor ;
- M. C. Moser, Engineer ;j
W. E.' Rogers, LR Mechanical Engineering Supervisor
M. Huff, LR Mechanical Engineering Supervisor
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M. R. Cumbest, Supervisor, Fire Protection (MSU SSI, New Orleans)
- M. C. Snow, Licensing Engineer
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- P. Michalk, Plant Licensing Engineer ;
G. D'Aunoy, Unit 2, Operations i
^S. McGregor, Engineering Services Supervisor !
B. L. Bata, Quality Assurance (QA) Engineer
- D. R. Brown, Nuclear Quality
- G. Higgs, Plant Engineering
B. E. Williams, Plant Engineering
"G. Storey, Engineering Services !
- R. Rispoli, ANO Fire Protection Specialist
K. Dickerson, Plant Engineer
E. Force, Training
R. Shinkowski, Shift Supervisor
S. Feemster, Senior Radiation Officer
l D. Olsen, Radiation Officer
l B. Flake, Radiation Officer
S. Szabo, Shift Administrative Assistant
- C, Zimmerman, U-1 Operations Technical' Support Supervisor
B. L. Garrison, U-1 Operations Technical Support
R. Ashcraft, Electrical Maintenance Supervisor
- J. Johnson, Engineering Technical, Fire Protection
R. Oakley, I&C Engineering Supervisor
- P. Pittman, Electrical Engineer
- D Williams, Engineer-I
W. Cottingham, I&C Engineer
P. Crossland, Training
B. McBride, Shift Supervisor
L. McLerran, Control Room Supervisor
R. Pierce, Radiation Officer
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R. Carter,' Radiation Officer
N. Yockey, Shift Administrative' Assistant
- C. W. Taylor, U-2 Operations Technical Support Supervisor
- T. Robinson, Fire Protection Specialist
- D. Provencher, QA Supervisor.
- J. Taylor-Brown, Quality Control (QC). Superintendent
- R. Lane, Manager,. Engineering
- B. Hinton, Operations Technician Engineer , !
'*S. M. Quennoz, General Manager, Plant Operations
G. W. Woerner, Mechanical Engineer-
J. Lamb, Fire Prevention and Safety Coordinator
J. Waid, Technical Support Training Supervisor
R. Hargrove, General Employee Trainer
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- E. J. Shelton, Nuclear Programs
Bechtel
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T. Gilmartin, Electrical Field Engineer l
0. Barnhouse, Electrical Field Engineer ;
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- Denotes those attending the exit interview conducted on June 12, 1987.
The team members also interviewed other AP&L personnel during the
inspection. '
2. List of Documents Reviewed
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See attachment to this appendix.
3. Fire Protection / Prevention Program Units 1 and 2
This inspection was conducted to determine that the licensee had
established and was implementing a program for fire protection and
prevention in conformance with regulatory requirements and industry guides
and standards.
The NRC inspectors reviewed the documentation constituting the licensee's
approved fire protection program. These documents are referenced in
paragraph 2 of this report. The licensee's program provides for.the
control of combustible materials and housekeeping for reduction of fire
hazards. Administrative controls have been established to handle disarmed
or inoperable fire detection or' suppression systems; provide for
maintenance and surveillance on fire suppression, detection, and
emergency communications equipment; establish personnel fire fighting
qualifications, training, and fire protection staff responsibilities;
provide fire emergency personnel designations as well as plans and
actions; and establish controls for welding, cutting, grinding, and other
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ignition sources.
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'The NRC inspectors conducted a walkdown of the fire suppression water. j
system and verified that it was operable as: required by Technical j
Specifications.
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A tour of' accessible areas of the' plant was conducted to assess general, 1
area condition, work activities.in progress, and visual condition of fire a
protection systems and equipment. Combustible materials and flammable and ;
combustible liquid and gas usage.were restricted or properly controlled in j
areas containing safety-related equipment and components. Items checked;
included positions of selected valves,' fire barrier condition,-' hose i
stations, hose' houses, halon' system lineups,-fire lockers, and fire- "'
extinguishers for type, . location, and condition. 'i
There was'no welding, cutting, or use of-open flame ignition.. sources foun'da
in the areas toured.- There were no construction act_ivities'in progress in .
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.the toured areas. There was'some maintenance work and surveillance
testing noted. General housekeeping conditions were found,to be' goo'd.
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Fire protection systems and equipment installed for protection of .
safety-related. areas'were found'to be functional and. tested in accordance l
with the requirements specified in the Technical Specifications- Fire
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brigade equipment, including. emergency breathing. apparatus,.was found to j
be properly stored and maintained. '
The NRC inspectors'also reviewed fire brigade training and drill records.
The records were iniorder and confirmed that training and drills were .I
being conducted at the specified intervals.- Individual qualifications and :(
training were found to meet BTP-APCSB19.5-1 requirements. A review of:the'
current roster of_ qualified fire brigade members verified that brigade
composition is'in accordance'with Technical Specification requirements.
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By letter dated August 15, 1984, the-licensee requested approval of a. i
number.of exemptions from'Section III.G of Appendix'R to the: extent that.
it requires protection for structural steel which is . framed. into or
supports a fire' barrier. During the Unit.1 inspection,.the NRC inspectors :
discovered unprotected steel in an area'not encompassed by the exemption
request. The licensee provided a copy of a fire hazards' analysis for this ,
area. .It was considered inadequate by the staff because it did not:
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consider the potential'for areas of concentrated combustibles which could
produce localized heating sufficient to cause the steel to fail. During
the Unit 2 inspection, the audit team.also discovered additional areas l
containing unprotected steel ~not encompassed by the licensee's August 15,
1984, exemption request. The' licensee produced a revised fire hazards
analysis which was conducted by'a consultant. . (The. methodology utilized
in this-analysis was one'that had previously been accepted by the staff.)
The results of this analysis-indicate that a localized heating problem
exists in Room 111 in Unit 1 and Rooms 2055 and 2084 in. Unit 2. The
licensee is studying options-to mitigate.this problem but'no actual'
modifications have been completed. .Because the steel in these areas is.
supporting a fire barrier and is l unprotected from the ' effects of a fire, o
and because no exemption for this condition is either pending or approved,:
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this represents a potential violation of.Section III.G of Appendix R.
(313/8714-01; 368/8714-01) ,
During the licensee's review of the AND Fire Protection Program against
the guidelines of Appendix A to BTP-APCSB 9.5-1, the plant was divided ,
into distinct fire areas and significant fire hazards were isolated by !
fire-rated walls and floor / ceiling assemblies. These fire barriers were
to be.surveilled under the provisions of the plant Technical
Specifications. The NRC inspectors discovered that certain fire barriers, !
such as portions of the wall and floor / ceiling surrounding the control J
room were not being surveilled per the Technical Specifications. This j
represents a potential deviation from the licensee's commitment to )
establish and maintain these fire barriers. (313/8714-02; 368/8714-02)
During the review of the ANO Fire Protection Program against the ;
guidelines of Appendix A to BTP-APCSB 9.5-1, the licensee committed to I
design and install a fire alarm system in accordance with the provisions i
of National Fire Protection Association Standard No. 720, 1975. I
Paragraph 2464 of the standard stipulates, "An audible trouble signal at a i
central supervising station may be common to several signaling line l
circuits. The act of silencing it upon operation in connection with one '
signaling line circuit shall not prevent it from operating immediately
upon the occurrence of trouble on other signaling line circuits."
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The fire alarm system at AN0-1 and -2 is not designed to annunciate
subsequent trouble alarms after an initial alarm is silenced. This
condition represents a potential deviation from the licensee's commitment ,
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to conform with NFPA 72.0. (313/8714-03; 368/8714-03)
During the BTP-APCSB 9.5-1, Appendix A review, the licensee submitted to
the staff, for evaluation, a number of fire barrier penetration seal
designs. The staff accepted these asigns as documented in the fire
protection safety evaluations for ANO-1 and -2. Subsequently, the
licensee submitted additional seal designs for staff review. NRR action i
on this issue is pending. During the inspection, the licensee notified
the NRC inspectors that qualification fire tests on three additional seal
designs (not previously reviewed by the staff or pending with NRR) were i
unavailable. The licensee has committed to submit these seal designs to a
standard fire test. Pending transmittal of the test results.and
evaluation by NRR, the adequacy of these seal' designs is considered an
unresolved item. (313/8714-04; 368/8714-04) :
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4. Emergency Lighting System, Units 1 and 2
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The NRC inspectors examined the emergency lighting system required for
( safe shutdown. Section J of Appendix R requires that emergency lighting
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units with at least an 8-hour battery power supply be provided in all
areas needed for operation of safe shutdown equipment and in access and
egress routes thereto'. The licensee has installed all cf the emergency
lighting units required for the operation of safe shutdown equipment in
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the event of a fire that destroys and forces evacuation of the control
room.
During the alternate shutdown procedure walkdown for both units, the team
members observed adequate lighting at work stations and in access. routes
to the components and controllers.
5. Post-Fire Safe Shutdown Capability (Unit 1)
a. Systems Required for Safe Shutdown
The systems are grouped according to the performance goals for PWR
safe shutdown functions to achieve both hot standby and cold
shutdown.
(1) Reactivity Control
Initial reactivity control is provided by operator initiation of
a reactor manual trip from the control room upon notification of
a major fire requiring plant shutdown. The control rods will be
inserted into the reactor core thereby bringing it subcritical.
The reactor may also be tripped from outside the control room by
local trip of the control rod drive system feeder breakers.
Additional boration will be provided during cooldown by j
intermittent operation of-the makeup /high pressure ;
injection (HPI) pumps taking suction from the borated water ;
storage tank (BWST). The BWST has a minimum volume'of 362,000 i !
13,000 gallons of 2470 1 200 ppm boron concentration water per ;
the ANO-1 Technical Specifications. With. letdown paths ,
isolated, the licensee has determined that supplying borated l
makeup water from the BWST to offset RCS inventory shrinkage
during cooldown will provide sufficient boration to assure
better than a 1 percent shutdown margin during the cooldown and
subsequent Xenon decay.
(2) Reactor Coolant Makeup (Level and. Pressure Control)
For a post-fire safe shutdown, reactor coolant system (RCS)
inventory will be controlled by isolating all . reactor coolant
leakage paths and verifying isolation of these paths. Potential ,
loss pathways include' normal letdown, reactor coolant pump (RCP)
seal bleedoff, RCS high point vents, pressurizer power operated
relief valves, RCS sample valves, and decay heat suction valves.
The reactor coolant pumps will be secured for the duration of
the incident to avaid seal failure which might occur while
attempting to restart the pumps.in an' emergency situation.
The only maLeup thus required will be to offset inventory
shrinkage resulting during the cooldown. Steam generator
overcooling pathways will be controlled to regulate RCS
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inventory shrinkage by controlling emergency feedwater (EFW)
flow, closing the main steam isolation valves and turbine. bypass !
valves, tripping and verifying tripped the main turbine, and
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controlling the atmospheric dump valves.
Isolation of the direct RCS leakage paths and control of the ;
steam generator overcooling pathways as described above will l
delay the need for makeup for about 1 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> following a
post-fire reactor trip.. Borated inventory makeup will then be j
provided as required via the makeup /HPI pumps taking suction I
from the BWST and discharging through the HPI valves as i
described in Section (1) above. The makeup /HPI pumps will be i
started and stopped to maintain inventory (as monitored by j
temperature-compensated pressurizer level) and control RCS ]
pressure. j
l RCS pressure may be more' easily controlled by operation of the
vital' bus powered pressurizer heaters if they are available.
However, the pressurizer heaters 'are.not mandatory for achieving i
and maintaining safe shutdown.
(3) Decay Heat Removal I
Decay heat will be removed from the reactor following a reactor
scram via the steam generators by natural circulation cooldown.
Emergency feedwater is supplied to the steam generators by the
emergency feedwater system to provide makeup for the inventory
discharged as steam from the safety relief valves and the
atmospheric dump valves.
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The emergency feedwater system consists of one motor-driven
pump (P7B) and one turbine-driven pump (P7A) interconnected to
permit supply of emergency feedwater to either one or both of
the steam generators from either or both of the pumps. The
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motor-driven pump and its associated components are powered from
the red train, AC engineered safeguards buses. The
turbine-driven pump is powered by steam from the steam
generators and its associated components and controls are
supplied by the green train, battery backed DC buses.
The EFW condensate storage tank (T418) will serve as the normal
water source for the emergency feedwater system. The EFW CST
has a Technical Specification minimum capacity of
107,000 gallons which equates to about a 4 -hour supply for
natural circulation cooldown. The backup source of feedwater is
the service water system. The CST T41 may also be aligned to
the EFW system suction header.
The EFW system automatic operating logic is provided by the
emergency feedwater initiation and control system (EFIC). The
system will auto initiate upon loss of main feedwater, low steam
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generator level or pressure, loss of all RCPs, or ECCS
actuation.
Steam release from the! steam generators will be controlled by
the atmospheric dump valves and/or the mechanical safety. relief
valves. Controlled operation of the atmospheric dump valves
will be utilized to achieve the desired RCS cooldown rate. '
(4) Process Monitoring
The following process monitoring instrumentation is available in
the control room and on~the safety parameter display
system (SPDS) " Alternate Shutdown" display:
Source Range Flux
Reactor Coolant System Pressure ;
Reactor Coolant System Hot Leg Temperature j
Reactor Coolant System Cold Leg Temperature i
Steam Generator Level J
Steam Generator Pressure
Pressurizer Level
CST 141B Level !
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The SPDS at' ANO is a computer-based system used for monitoring
and display of plant safety parameters developed in accordance
with the requirements of NUREG-0737. The SPDS configuration at
ANO is designed to provide redundant. isolated data acquisition,
processing, and display devices. Two redundant display .
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terminals are available in the control room and also-in the TSC ;
for alternate shutdown. Software has been developed to display '
the above parameters as well as provide trending data on an
" Alternate Shutdown" display screen.
In addition, the following local indicators are available:
P7A EFW Pump Discharge Pressure
Decay Heat Pump Discharge Pressure
Decay Heat Pump Suction Temperature
Steam Generator Pressure (at the ADV area) ,
(5) Support Systems
The safe shutdown components and systems described in
Sections (1) through (4) require the operation of several-
critical support systems in order to properly perform.their safe
shutdown function. .The following systems must have one train
operating to support the safe shutdown.
Service Water System
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Diesel Fuel 011 Transfer System l
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Emergency (Engineered. Safeguards) AC Power Distribution
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Uninterruptible DC Power Distribution System !
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HVAC for all Essential Areas j
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Radio Communication System
SPDS j
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(6) Cold Shutdown !
The reactor coolant system temperature and pressure will be )
reduced by natural circulation cooldown using the atmospheric i
dump valves and the emergency feedwater system as described in l
the previous sections. 'Once the reactor coolant system _
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temperature has been lowered to less than 280 F, the RCS'will be
depressurized to e. approximately 225 psig, and the decay heat
removal (OHR) system.will be initiated. The DHR. system will be
used to reduce RCS temperature to 200 F and maintain cold
shutdown. Some minor repairs at motor control center breaker
cubicles in the form of lifted leads and jumpers for control
circuits and control power fuse replacements may be required.
It should be noted that the licensee was granted an exemption
for ANO-1 from the requirement that the plant be capable of j
achieving cold shutdown within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> without the use of '
offsite power. Without offsite power, the RCPs cannot_be -l
operated; therefore, pressurizer auxiliary spray:is unavailable. '
A very conservative controlled cooldown to DHR' system cut-in
conditions without void formation in the reactor vessel would
take at least 135 hours0.00156 days <br />0.0375 hours <br />2.232143e-4 weeks <br />5.13675e-5 months <br /> per licensee's analysis-with
approximately another 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> to reach cold shutdown.- For the
purposes of this audit, the reference documents pertaining to
this exemption were reviewed for baseline information only,
b. Alternate Shutdown
In_the event of a fire in the main control room (Fire Area G,
Zone 129-F) and/or the cable spreading room (Fire Area G, Zone 97-R)
which results in a functional loss of control room instrumentation
and controls, or requiring evacuation of the control room, the
licensee has provided alternate shutdown capability. The
Procedure 1203.02, " Alternate Shutdown," will implement the safe
shutdown from outside the control room.
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, Initial reactivity control will be provided by manually scramming the
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reactor just prior to evacuating the control room. The shift
supervisor will direct the remainder of the shutdown from the
Technical Support Center (3rd floor of the administration' building)
using portable radios to communicate with shift personnel deployed
throughout the plant. Monitoring of plant parameters is provided on
either of the two redundant safety parameter display system terminals
in the technical support center (TSC).
l' It should be noted that AN0-1 and -2 share what.is essentially a
common control room, separated by only a glass wall. Both unit-
control rooms share a common ventilation and air conditioning system.
For this reason, a significant Unit 1 control room fire requiring
evacuation would in all likelihood require a Unit 2 control room
evacuation and shutdown of Unit 2 in accordance with
Procedure 2203.30, " Remote Shutdown." See Section 6.b for further
-discussion.
The alternate shutdown procedure directs reactor coolant system level
and pressure control by locally verifying and manually isolating as
required the RCS letdown pathways and steam generator overcooling
pathways. Operators then verify the makeup /fiPI injection pathway and
locally operate the pumps from the switchgear breakers under
direction of the shift supervisor.
RCS pressure may be more easily controlled by local manual operation
of the vital bus powered pressurizer heaters from the motor control
center breakers under supervision of the shift supervisor, if control
circuits have not been faulted by control room / cable spreading room
fire failing the SCR gating circuits. The pressurizer heaters are
not mandatory, however, for achieving and maintaining safe shutdown.
Decay heat removal by natural circulation is. established by first
verifying the operation of the turbine driven EFW Pump P7A. Steam
supply to the turbine and flow path from CST T41B.to the steam
l generators will be verified, and the turbine manually started if it
did not auto start or had tripped. The EFW flow will be locally
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directed by the shift supervisor to achieve a stable natural
circulation cooldown. As time and additional manpower permit, the
motor-driven EWF Pump P7B flow path will be verified. When AC power
is restored P7B may provide feedwater to the steam generators in
preference to P7A to eliminate the need for P7A trip / throttle
adjustment.
Steam release from the steam generators will initially be provided by
the mechanical safety relief valves. As additional manpower becomes
available, the atmospheric dump valves will be manually operated
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locally (manual operation of the air-operated valves), the operator
will be told to reduce steam header. pressure as required to achieve
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the desired cooldown rate. If the isolation valve control circuit is. .
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undamaged, the isolation valves may be electrically operated at the
MCCs to control RCS cooldown rate.
6. Procedures (Unit 1)
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The procedures reviewed during this inspection include a post-control room
fire alternate shutdown procedure, a remote shutdown procedure, Technical-
Specification surveillance procedures, fire alarm corrective action
procedures, and plant operating procedures. A sampling of fire preplans
from the fire protection program manual was also reviewed.
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a. Alternate Shutdown Procedure 1203.02
In the event of a fire in the main' control room and/or the cable !
spreading room which results in evacuation of the ANO-1 control room, I
l the licensee may shut the reactor down from outside the control room {
in accordance with the Procedure 1203.02, " Alternate Shutdown."
The minimum operating shift complement for Unit I consists of the
following personnel:
Title Control Room Evacuation Assignment
Shift Supervisor 3rd Floor TSC
Shift Admin. Asst. 3rd Floor TSC & plant library
Shift SR0 switchgear, EDG, battery rooms !
Reactor Operator No.1 Various locations in auxiliary / turbine I
buildings per procedure & shift supervisor's
direction
Reactor Operator No. 2 Upper north / lower south piping penetration
rooms, EFW pump room, various locations per
shift supervisor's direction
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Waste Control Operator Fire brigade leader, affected unit
Auxiliary Operator Fire brigade member
The first five people above are required to implement the control
room evacuation procedure. The shift supervisor is in charge of the
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normal operating shift and he continues in that role upon entry into
the procedure. He also serves as emergency director of the emergency
response organization until-the TSC is formally activated. The waste
control operator and auxiliary operator together with other onshift-
personnel will staff the fire brigade. Available additional operators
will be utilized to support the required operators.
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Once the decision to' evacuate the' control room has'been made, th'e.
reactor will be tripped from the control room driving the control 1
rods in for. initial reactivity control. Severa1 'other immediate
- actions will:be attempted from the main control' room pr.ior to-
evacuation. If unsuccessful, however, all are covered procedurally
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from outside the control room soon after the evacuation. The
operating shift personnel then proceed to their post-evacuation
assignments as outlined above, to implement their assigned procedure .j
sections, attachments, and checklists. The remainder of the
L alternate safe shutdown will be directed and-monitored by the' shift l ,
supervisor from the TSC, 3rd floor of the administration building -
b. Procedure Walkdown
The alternate shutdown procedure (1203.02, Revision 4)walkdownwas
initiated at 9:23 a.m. (CDT) on May 7, 1987, using five people from
the licensee's operating staff with the proper training and
qualifications to fill the. control room operating shift positions.
Since the procedure is symptom oriented, the following initial .
conditions were presented:
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Fire in control room of sufficient size to require-
evacuation
Coincident loss of offsite power 4
Botn EFW pumps fail to auto start
Both standby diesel generators failed to auto start !
All system line-ups as normally found at full power
Two NRC inspectors accompanied the shift supervisor and the shift
administrative astistant to the 3rd floor TSC. They remained there
to observe personnel actions, crew direction and leadership, i
communications, use of the SPDS, interface with and initiation of j
emergency plan implementation procedure, and training and familiarity ;
with the alternate shutdown procedure. Reactor. Operator Nos.1 and 2-
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were not accompanied by NRC inspectors during .the walkdown,'so their
actions were' monitored from the TSC as they made their' call-ins to
the shift supervisor.. A third NRC inspector accompanied'the shift
SR0 to observe his actions.at the electrical equipment rooms and
EDGs, observe lighting 'at~ work ' stations and access routes, evaluate
l- communications, and evaluate training and familiarity with the )
procedure. -
The procedure was halted during Step 13, Item No.1, of Shift
Supervisor Section.1A which demonstrated that a stable hot shutdown
could be achieved and a controlled rate of cooldown was to be ;
commenced. Emergency feedwater flow to.the steam generators via the !
turbine-driven EFW Pump P7A was established and verified within -
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12 minutes. Both EDGs K4A and K4B were manually started and loaded
within 20 minutes of start of event. The operating staff
demonstrated adequate training and familiarity with the procedure '
.
throughout the walkdown.
The only procedure deficiency identified was the lack of a. clear i
procedural interface between the Unit I and 2 control rooms for the i
situation where a control room fire requires evacuation of one unit's ;
control room. The.two control rooms share a single ventilation and i
, air conditioning system and are only physically separated by a glass
I
wall. There should be a procedure to provide positive direct a
notification of the other unit's shift supervisor when a control room
fire evacuation in one unit is imminent. The second unit's control
- room response to such a notification should also be established by
l procedure.
Other improvements to the alternate shutdown procedure would be:
Provide a procedure step to perform a radio check at the time
that portable radios are picked up by operations personnel at
the alternate shutdown equipment cabinet. This assures that
working communication will exist during the first critical
minutes of the procedure implementation.
Establish a periodic program to check contents of the alternate
shutdown equipment cabinet and file cabinet. This will assure
that all procedures, tools and equipment used to implement the
procedure will be there if they should ever be needed,
c. Fire Alarm Corrective Action Procedures and Fire Preplans
Procedure 1203.12K, " Annunciation K12 Corrective Action," and l
Procedure 1203.09, " Fire Protection System Corrective Action," were l
reviewed to evaluate the train of responses followed by the control
room operator upon receipt of a fire annunciation alarm. The
procedures were found to adequately direct the operators to identify
the origin of an alarm, verify its validity, and respond accordingly.
The fire preplans were readily available in the control room to guide j
the operators in their response. ;
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d. Technical Specification Surveillance Procedures
Surveillance test procedures covering a sampling of Technical
Specification surveillance requirements for various post-fire safe
shutdown components and systems were reviewed. These included .
procedures for the Makeup /High Pressure Injection Pumps P36A and l
P36B, the motor-driven Feedwater Pump P7B, and the turbine-driven
auxiliary Feedwater Pump P7A. The past two periodic performances of
Procedures 1104.02, Revision 22, " Makeup and Purification System
Operation," (a quarterly Technical Specification surveillance), and
Procedure 1106.06, Revision 31, " Emergency Feedwater Pump Operation," ,
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. . _ . .. . . . . . . . . . . _ _ . . . . .. . . _ . .
. .
.. .
.. ... .
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' Supplement I for P7B and Supplement II for P7A (monthly Technical'
Specification surveillance) were retrieved for inspection of test
results. This review showed the procedures to be adequate and the
performances reviewed were performed on time and documented
adequately.
e. Operator Training on Safe Shutdown Procedures
]
i
In addition to observing the operator's performance during the
'
walkdown of the alternate shutdown procedure, operations technical
support and operator training personnel were interviewed concerning
operator training on Appendix R post-fire safe shutdown procedures !
and equipment. The training covered Appendix R equipment, Alternate I
Shutdown Procedure 1203.02, and the fire protection program manual,
and fire preplans. The program includes classroom instruction, ,
walkdowns, simulation, and hands-on operating experience. A 2 ycar
requalification cycle is being maintained.
.)
Lesson plans related to Appendix R training were provided for J
inspection. Training records for operating shift personnel were also
'
reviewed. These areas' reviewed were found to be adequate.
,
The licensee is presently revising the Appendix-R rehted training to j
enhance the fire preplan instruction and to integrate plant simulator J
exercises into the safe shutdown / alternate shutdown portions of the
course. These improvements are scheduled to be incorporated by late
1987.
l
7. Protection for Associated Circuits (Unit 1)
Common Bus Concern
Spurious Signals Concern
Common Enclosure Concern
a. Common Bus Concern
, The common bus associated circuit concern is found in circuits,
i either nonsafety-related or safety-related, where there is a common
power source with shutdown equipment and the power source is not
electrically protected from the circuit of concern.
The common bus concern is made up of two items:
Circuit Coordination
High Impedance Fault Analysis
(1) Circuit Coordination
Breaker Coordination is audited by reviewing the time current
curves developed during the licensee's bus coordination study.
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At ANO-1, the following circuits were randomly selected for
review:
Circuit Comment
! 4160 V DG1 Coordination Satisfactory
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4160 V-BUS A3 Coordination Satisfactory
4160 V DG2 Coordination Satisfactory
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4160 V BUS A4 Coordination Satisfactory 1
480 V LC B5 Coordination Satisfactory
480 V LC B6 Coordination Satisfactory
480 V MCC B52 Coordination Satisfactory
480 V MCC B62- Coordination Satisfactory
125 VDC PNL 001 Coordination. Satisfactory
125 VDC PNL 002 Coordination Satisfactory
'
125 VDC PNL D11 Coordination Satisfactory l
125 VDC PNL D21 Coordination Satisfactory l
120 VAC PNL RS1 Coordination Satisfactory.
120 VAC PNL RS2 Coordination Satisfactory .I
PNL C539 Coordination Satisfactory
BNL C540 Coordination Satisfactory
The licensee's coordination program was found to be i
satisfactory. )
1
To ensure that the existing satisfactory circuit coordination is
not compromised by future design changes, the licensee has an
established procedure for modification design review,
Procedure 216, Revision-1, dated April 30, 1987, " Guidelines for
Evaluation of Safe Shutdown Capability and Control of Safe
Shutdown Capability Assessment," which provides for reviewing
modification design for Appendix R concerns.
The licensee performs relay testing and maintenance at 18-month
intervals (each refueling outage). Circuit breakers are. tested
and maintained at intervals of up to 60 months. Breaker-and
relay maintenance and testing are currently scheduled manually. 2
The licensee is in the process of converting to an automated !
maintenance scheduling system.
Maintenance records for the following randomly selected circuit
breakers or protective relays were reviewed to verify that
maintenance and testing are being performed at the specified
frequency:
Maintenance Required Comp
Component Title Procedure Frequency Date
BKR A-409 A-4 bus Feeder 1307.07 Refueling 11/12/82
10/29/84
10/02/86
L__ ___---_1---_------
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BKR B-633 Reector Bldg. 1403.81^ 234 Weeks 01/23/83
Cooling Fan 11/26/84
BKR B-723 Reactor 1403.82 234 Weeks 12/15/84
Pressurizer 11/20/86
Heatera
BKR B-512 B-5 Mair, 1403.83 306 Weeks' 12/19/82'
Breaker ;
I
WA302/A- Service Water 1403.150 Refueling' 12/03/84 -l
150/151/M Pump P4A "A" 12/10/86
Phase Relay
I i
E11/A-187 DG1 "A" Phase 1403.151 Refueling 10/26/84 i
DG1 Differential 11/12/86 l
Relay a
The reviewed records documeated compliance with established , ,
maintenance procedures.
'
'j
Control of fuse replacement is required to ensure maintenance of
coordination for circuits protacted by fuses. The licensee's ;
controls for fuse replacement include the following: '
I *
Procedure 1403.85, " Motor Control Center Preventive
Maintenance"
.)
Plant Drawings / Prints -
Technical. Manuals , j
,
1
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1
Job Orders
l {
Material Controls. I
The licensee's fuse replacement control, were found to~be i
satisfactory. )
(2) High Impedance Fault Analys_is ,
The high impedance fault concern is found in the case where
multiple high impedance faults exist as loads -on a safe shutdown-
power supply and cause the loss of the safe shutdown power
supply prior to clearing the high impedance faults.
The licensee's analysis for high impedance faults, EE-87-014,
January 21, 1987, ANO-1 and -2, " Position on Multiple High
Impedance Faults," determined that protection for simultaneous
high impedance faults was provided by the following:
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4.16 KV safety buses are equipped with ground fault relays.
'
The 4.16 KV and 480 V distribution systems are grounded.
High impedance faults would rapidly propagate into low
impedance faults and cause coordinated circuit breaker
tripping.
In the event that a required electrical bus.is lost due to
high impedance faults,.the licensee's procedures provide
for manually tripping all breakers on the faulted bus and-
reenergizing required safe shutdown loads.
w ., i
The' licensee's analysis and protection for high impedance faults l
were :found to be satisfactory. j
The licensee's protection for the' associated circuit coinmon bus
concern was found to be satisfactory. ;
b. Spurious Signals
i
The spurious signal concern is made up of two items:
i
The false motor, cor trol, and instrument readings such as !
,
occurred at the 197S Brown's Ferry fire. These could be caused
l by fire-initiated grounds, short or open circuits.
Spurious operation of safety related or nonsafety related I
components that would adversely affect shutdown capability
(e.g., RHR/RCS isolation valves).
,i
(1) High/ Low Pressure Interfaces l
The licensee has identified the following high/ low pressure
interfaces and methods for controlling the interfaces: ;
Interface Method of Control Status
l- ERV & Associated Block valve CV 1000 has a transfer
Block Valve switch at MCC 861. Separation of cables
is maintained outside of containment. !
Decay Heat Drop Procedure 1102.02, Revision 36, I
Line Valves December 16, 1986, Plant'Startup ;
Step 6.4.9.C, opens CV-1050 breaker 1
after shutting the valve.
Letdown Valves The cables to CV-1214, CV-1216, and !
CV-1221 are routed separately outside of
l the Control Room. .The alternate
shutdown procedure deenergizes MCC B51
to assure that the interface is isolated.
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19
High Point Vents .The design flow 1 capacity at the high
point vent path is_less than the
definition of a LOCA.
'The licensee's protection for. fire induced spurious operation of
high/ low pressure' interfaces was found to be satisfactory.
(2) Current Transformer Secondaries
The licensee's current transformer (CT) analysis, EE-86-0002,
January 10, 1986, ANO-1, "Possible Appendix R Concerns Due to
Open CT Circuits," determined that the CT saturation
characteristics will. limit potential and energy in CT
secondaries such that secondary fires could not be induced by
open CT secondaries.
The licensee's protection for current transformer open secondary
concerns was found to be satisfactory.
(3) Isolation of Fire Instigated Spurious Signals
The licensee has provided isolation for fire instigated spurious
signals by various methods, including:
Administrative controls '
Rerouting of cables
Wrapping cables
Isolation / Transfer switches (redundant fuses used)
Fuses
Signal isolators
During the inspection, all forms of isolation listed above were
observed.
The licensee's methods of fire-instigated spurious signal
isolation were found to be satisfactory.
c. Common Enclosure
The common enclosure associated circuit. concern is found when
redundant circuits are routed together in a ' raceway or enclosure and
they are not electrically protected, or fire can destroy both
circuits due to inadequate fire protection means.
Licensee representatives state that:
Redundant safe shutdown cables are never routed in common
enclosure.
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Nonsafety-related. cables routed in common enclosure with I
redundant safety-related cables are never routed between 'l
redundant trains.
All circuits are electrically protected.
During the' inspection, no exceptions to the above statements were
noted.
I The licensee's protection for the common enclosure associated circuit
concern was found to be satisfactory. j
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d. Cable Routing l
'
Documentation (cable routing) review and physical in plant inspection
were performed on the following: i
Component Type Cable
Primary Makeup Pumps P36A/B/C Power
Fuel Oil Transfer Pumps P16A/B Power
The cabling for th'e redundant' fuel oil transfer pumps has less than
20 feet of horizontal separation in several areas of the plant. The
licensee has provided alternate shutdown capability using Unit 2 fuel ,
oil transfer pumps. I
i The cabling for the redundant primary makeup pumps has been provided
separation by 3-hour fire barriers or has been fire wrapped.
'
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1
, The licensee's separation analysis for components requiring manual ;
I operations in Fire Area B was questioned by'the electrical. inspector ;
and the NRR representative. Licensee representatives had stated that l
redundant cables for manually-operated components in Fire Area B were- i
not completely walked down to verify separation in accordance with
Section III G.2 separation requirements. The licensee's basis for
not fully walking down Fire Area B cable routings was that credit was
being taken for manual operation so that the cables were not
required.
During the inspection, the licensee failed to provide marked up
drawings of cable routings for selected components. The licensee's
documentation to support providing the requested information was
available; however, the licensee did not have personnel resources.
available to produce the requested information during the inspection.
Cable separation / routing was found to be an unresolved item pending
documentation review and physical in plant inspection (to be '
l accomplished during the inspection of Unit 2 in June 1987, see
paragraph 13.b) of the following:
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Components Type Cables ;
.
CV1407 and CV1408 (BWST) Power and Control
s
CV1219 and CV1220 (HPSI) Power and Control
PSV1000 and CV1000 (PZR.PORV) Power and Control
1
CV1228 and CV2618 (SG ATMOS and associated j
block valves)' Power and Control j
P4A/B/C (Service Water Pumps) Power and Control-
LT1001 and LT1002 (PZR Level) Instrumentation j
PT1042 and PT1041 (RCS Press) Instrumentation
TE1144 and TE1147 (RCS Temp) Instrumentation
!
LT2620 and LT2624 (SG Level) Instrumentation
NE501 and NE502 (Source Range) Instrumentation l
l
LT4204 and LT4205 (CST Level) Instrumentation ,
l
, Redundant Components in Fire Area B Power and Control I
requiring manual operation
i
e. Modification Review
The licensee's process for controlling the design and installation of ,
modifications was reviewed for proper review and approval, including
10 CFR 50.59 aspects. The following Appendix R modifications were
reviewed:
Tech Safety
Testing Training Spec Question
Modification Description _Compl Compl Review Review
,
DCP 84-1061 EDG Room Exhaust 05/31/85 01/20/86 11/17/84 11/17/84
Fans Control
Isolation
i
DCP 83-1008 Disconnect 05/01/85 11/06/84 09/26/84 09/26/84
l Switches "B"
Makeup Pump
DCP 83-1012 CV-1000 03/25/85 10/15/84 09/12/84 09/12/84
Modification
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DCP 82-2079 EDG Cross 07/18/83 09/20/82 10/30/82_ 10/30/82
Connect
The control of modification design is governed by Procedure 202,
Revision 10, dated November 7, 1986, " Design Process," which provides
for design review for Appendix R concerns. The administration of
modifications for Appendix R concerns'was found to be satisfactory.
f. Control of Cables
The licensee controls / tracks cables using Procedure 216, Revision 1,
dated April 30, 1987,." Guideline'for. Evaluation of Safe Shutdown
Capability and Control of Safe Shutdown Capability Assessment." This
procedure provides the following forms for maintaining the-cable data
base:
Form Description
216 F2 Used to add new shutdown components and their related
cables to the data base.
216 F3 Used to add new shutdown cables and associated cable
data.
The licensee's control of cables was found to be satisfactory.
8. Post-Fire Safe Shutdown (Unit 2)
a. Systems Required for Safe Shutdown
The systems required for safe shutdown are grouped according to the
performance goals for PWR safe shutdown functions to achieve both hot
l standby and cold shutdown.
(1) Reactivity Control
Upon notification of a major fire, initial. reactivity control is
, provided by inserting the control element assemblies via
l
operator initiation of a reactor manual trip from the main
i_ control room. The reactor may also be tripped from outside the
control room by locally opening reactor protection system
i breakers or tripping the reactor protection system
i
motor generator sets.
1
Additional boration will be provided throughout the cooldown by
intermittent operation of the coolant charging pumps (CCPs)
taking suction from the boric acid makeup (BAM) tanks. The BAM
tanks (2T6A and 2T68) are required by Technical Specifications
to contain a minimum of 8121 gallons (65 percent capacity) each
at a baron concentration of no less than 8750 ppm. With the
letdown paths isolated, the licensee has determined that
- - - _
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23 l
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injection of boric acid from the BAMTs via a CCP for a 3
cumulative run time.of 84 minutes will assure that a '
-5.5 percent shutdown margin will be maintained during the
cooldown. In the event that suction from the BAM tanks isn't
available, the refueling water tank (464,900 gallon minimum
Technical Specification volume of at least 1731 ppm boron
concentration) may be used as a boric acid source. Boron
injection from the. refueling water tank will ensure that the
reactor will remain subcritical with at least a -1 percent 3
shutdown. margin.
(2) Reactor Coolant Makeup (Level and Pressure Control)
For a post-fire safe shutdown, reactor coolant system (RCS) 1'
inventory will:be controlied first by isolating all reactor
coolant leakage paths and' verifying isolation of these paths.
Potential loss pathways include normal letdown, reactor vessel
head vents, reactor coolant pump (RCP). seal bleedoff, RCS
high point vents, and pressurizer vent /LTOP relief isolation
valves.
The only makeup thus required will be to offset inventory
shrinkage resulting during the cooldown. Steam generator l
overcooling pathways will be controlled to regulate RCS
inventory shrinkage by tripping the main feedwater pumps
controlling emergency feedwater (EFW) flow, closing the main
steam isolation valves and turbine bypass valves, tripping and
i
verifying tripped the main turbine, and controlling the l
l
atmospheric dump valves.
'
Isolation of the direct RCS leakage paths and control of the
steam generator overcooling pathways as described above will
limit most of the variations in pressurizer level to inventory
shrinkage due to cooldown. Borated' inventory makeup will be
,
'
provided as required via the CCP taking suction from the BAM
tanks or refueling water tanks. The coolant charging pumps will
be started and stopped to maintain inventory (as monitored by.
temperature-compensated pressurizer level) and control RCS
l
pressure.
RCS pressure may be more easily controlled by operation of the
vital bus powered pressurizer heaters if they are available.
l However, the pressurizer heaters are not mandatory for achieving
and maintaining safe shutdown.
Decay heat will be removed from the reactor following a reactor.
scram via the steam' generators by natural circulation cooldown.
Emergency feedwater is supplied to the "B" steam generator
(2E24B) by the emergency feedwater system to provide makeup for
- _ - - _ _ _ _ _ _ _ _ - -
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1
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the inventury discharged as steam from the safety relief valves I
and.the atmospheric dump valves. Emergency feedater may also ;
be supplied to the "A" steam generator (2E24A); however, the-
instrumentation monitoring this component is not protected.to
the requirements of 10 CFR 50, Appendix R. Only one steam
generator is required to perform a natural circulation cooldown.
The emergency feedwater system consists of one motor-driven j
pump (2P78) and one turbine-driven pump ~(2P7A) interconnected to 'I
permit supply of emergency feedwater to either one or both of l
the steam generators from either or both of the pumps. The-
motor-driven pump and its associated components are powered. from
the red train, AC engineered safeguards buses. The
i turbine-driven pump.is powered by steam from the steam
l
generators and its associated components and controls are !
supplied by the green train, battery-backed DC buses.
l
The condensate storage tanks (2T41A and 2T41B), with maximum
capacity of 200,000 gallons each, will serve as the normal water
sources for the emergency feedwater system. The on-line CST has
a Technical Specification minimum capacity of 160,000 gallons,
which equates to about a 12-hour supply for natural circulation
cooldown. The backup source of feedwater is the service water
system.
The EFW system automatic operating logic is provided by the i
emergency feed activation system (EFAS). The system will auto
initiate upon loss of main feedwater, low steam generator level,
or low steam generator pressure.
Steam release from the steam generators will be controlled by
, the atmospheric dump valves and/or the mechanical safety relief
I
valves. Controlled operation of the atmospheric dump valves
will be utilized to achieve the desired RCS cooldown rate.
(4) Process Monitoring
The following process conitoring instrumentation is available in
the control room and on the SPDS " Alternate. Shutdown" display:
Source Range Flux
Pressurizer Pressure
Reactor Coolant System B Hot Leg Temperature
Reactor Coolant System D Cold Leg Temperature
B Steam Generator Level
B Steam Generator Pressure
Pressurizer Level
Pressurizer Temperature
The safety parameter display system (SPDS) at ANO is a
computer-based system used for monitoring and display of plant
___ _ _ -
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safety parameters developed in accordance with the requirements {
of NUREG-0737. The SPDS configuration at AND is designed to J
provide redundant isolated data acquisition, processing, and i
display devices. Two redundant display terminals are available
in the control room and also in the TSC for alternate shutdown. )
Software has been developed to display the above parameters as J
well as provide trending data on.an." Alternate Shutdown". display
'
screen.
In addition, the following local indicators are available:
~1
2P7A EFW Pump Discharge Pressure j
Shutdown Cooling Heat Exchanger (2E35A) Inlet Temperature !
I
Shutdown Cooling Heat Exchanger (2E35B) Inlet Temperature
Steam Generator Pressure (at the ADV area)
CST 2T41A Level i
CST 2T41B Level )
!
(5) Support Systems 1
1
The safe shutdown components and systems require the operation l
of several cr.itical support systems in order to properly perform 1
their safe shutdown function. The-following systens must have f
one train operating to support the safe shutdown: j
Service Water System l
l
Diesel Fuel Oil Transfer. System j
Emergency (Engineered Safeguards) AC Power Distribution
System
,
Uninterruptible DC Power Distribution System
l HVAC for all Essential Areas
Radio Communication System
At the time of the ANO-2 inspection, the licensee had not yet
provided an analysis to support the decision not to provide- !
ventilation for either the ANO-1 electrical equipment /switchgear
rooms (Rooms 95-0, 99-N,100-N,104-5, and 110-L on
elevation 368-378 ft. of the auxiliary building) or the ANO-2
electrical equipment /switchgear rooms (Rooms 2091-BB, 2097-X,
2099-W, 2100-Z, 2101-AA, and 2108-S on elevation 368-378 ft. of-
- _ _ _ _ _
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the auxiliary building) protected in accordance with 10 CFR 50,.
Appendix R requirements. These items will remain unresolved
until adequate analysis of the ventilation requirements has been
provided. (313/8714-05; 368/8714-05)
(6) Cold Shutdown i
The RCS temperature'and pressure.will be reduced by natural
circulation cooldown using the atmospheric dump valves and the j
emergency feedwater system.as described ~in the previous l
sections. Once the RCS temperature has been lowered to between
270 F and 275 F, the RCS will be depressurized to below
275 psia, and the shutdown cooling system will be placed'into
operation. The shutdown cooling system will then be used to
reduce the RCS temperature to 200 F and maintain cold shutdown. .l
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b. Alternate Shutdown
.
l
.The licensee has provided alternate shutdown capability for fire-
.
occurring in any of the following areas which results in a functional
loss of control room instrumentation and controls, or. requiring
evacuation of the control room:
2098-L, Cable Spreading Room l
2199-G, Control Room Unit II
2150-C, Core Protection Calculator Room ,
2136-I, HP Office
2137-I, Upper South Electrical Penetration Room
2119-H, Control Room Printer Room (Old Shift Supervisor Office)
i
The Procedure 2203.14, Revision 4, " Alternate Shutdown," will ;
implement the safe shutdown from outside the control room. ;
1
Initial reactivity control will be provided by manually scramming the
reactor jusf prior to evacuating the control room. LThe shift
supervisor will direct the remainder of the shutdown from the TSC j
(3rd floor of the Administration Building) using portable radios to
communicate with shift personnel deployed throughout the plant.
Monitoring of piant parameters is provided on either of the two
redundant safety parameter display system terminals in the TSC.
It should be noted that ANO-1 and -2 share what is essentially a
common control room, separated by only a glass wall. Both unit
control rooms share a common ventilation and air conditioning system. i
for this reason, a significant Unit 2 control room fire requiring
evacuation would in all likelihood require a Unit 1 control room
evacuation and shutdown of Unit 1 in accordance with
Procedure 1203.29, " Remote Shutdown."
The alternate shutdown procedure directs RCS level and pressure
control by locally verifying and manually isolating as required, the
_ -_ - -
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RCS letdown pathways and steam generator. overcooling pathways. .
Operators then verify the coolant charging pump. injection. pathway and-
4
-locally operate the pumps from the switchgear. breakers under
direction of the shift supervisor. i
Although pressurizer heaters are not relied upon for.'saf' e shutdown'
capability, RCS pressure may be more ' easily controlled by local'
manual' operation of the ESF. bus-powered pressurizer proportional
heaters from the local heater SCR Control Panels 2C117 and 2C118.
The proportional heaters may be operated either full on or_ full off
under the direction of the shift supervisor.- The backup pressurizer
heaters may also be controlled similarly from their. load center
supply breakers if offsite power.is available, q
Decay heat removal by natural _ circulation is established by first
verifying the operation of the turbine-driven EFW Pump 2P7A.. Steam.
supply to the turbine and flow path from the CSTs 2T41A and -B to the
"B" steam generator will be verified, and the~ turbine manually l started'
if it did not auto start or had tripped. The EFW flow will'be
locally controlled manually by throttling the 2P7A trip / throttle
valve as directed by the shift supervisor to achieve a stable natural'
circulation cooldown. As time and additional manpower. permit, the
motor-driven EFW Pump 2P7B flow path will be verified. When AC power
is restored, 2P78 may provide feedwater to the steam generators in -
preference to 2P7A to eliminate the need for 2P7A trip / throttle
adjustment.-
Steam release from the steam generators will initially be provided by
the mechanical safety relief valves. As additional manpower becomes
available, the atmospheric dump valves will be manually operated
under the direction'of the shift supervisor. If bein
locally (manual operation of the air operated valves)g operated-
, the operator
will be told to reduce steam header pressure as required'to achieve
the desired cooldown rate. If the isolation valve control circuit is
undamaged, the isolation valves may be electrically operated at the
MCC3 to control RCS cooldown rate.
c. Area Compliance With Appendix R,Section III.G.2
10 CFR 50, Appendix R, Section III.G.2, specifies that where
redundant trains of systems necessary to achieve and maintain hot
shutdown' conditions'are located within the same fire area, one of the
following means of ' ensuring -that one of the redundant trains is free
of' fire damage shall be provided:
(1) separation of cables and equipment and' associated nonsafety
circuits of redundant trains by a fire barrier having a'3-hour.-
rating,
(2) separation of cables and equipment and associated nonsafety
circuits of redundant trains by a horizontal distance of more
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than 20 feet with no intervening combustible or fire hazards.
In addition, fire detectors and an automatic fire suppression !
system shall be installed in the fire area, or ]
(3) enclosure of cable and equipment and associated nonsafety i
circuits of one redundant train in a fire barrier having a i
1-hour rating. In addition, fire detectors and an automatic
fire suppression system shall be installed in the fire area.
(a) Fire Area B, Fire Zone 2091-88, North Electrical Equipment
Room
Fire Zone 2091-BB was selected because it contained
redundant associated circuit components and cabling for the t
pressurizer vent and relief valves high/ low pressure {
interface. The following conditions were found to exisc:
There is less than 20 feet of horizontal separation
between the DC motor starters for 2CV-4698-1, ECCS
vent valve, and 2CV-4740-2, pressurizer LTOP relief
isolation. )
j
Fire detection is provided. l
Automatic fire suppression is not provided.
An exemption from the fixed fire suppression
requirement of Section III.G.3 has been granted by SER
dated March 28, 1983. (An alternate capability has
been provided for the fuel oil transfer pumps in this
zone)
Fire Area B, Fire Zone 2091-BB, north electrical equipment
room, was found to be not in compliance with Appendix R,
Section III.G.2. !
(b) Fire Area JJ, Fire Zone 2109-U, Corridor and Motor Control
Center
Fire Zone 2109-U was selected because it contained
>
redundant associated circuit cabling for the pressurizer
LTOP relief valves high/ low pressure interface, the ;
shutdown cooling suction valves high/ low pressure
interface, and the RCS letdown valves high/ low pressure ;
interface. The following conditions were found to exist:
Less than 20 feet of horizontal separation exists for
the following redundant cables:
- Pressurizer LTOP Relief Isolation 2CV-4731-2,
Power Cable G2861L2A and Control Cable G2861L2C,
_ _ _ _ _ _ - _ _ _
i
l
I
29 1
.and Pressurizer LTOP Relief Isolation 2CV-4730-1,
Power Cable R2851E4A.
l
-
Pressurizer LTOP Relief Isolation 2CV-4740-2, !
Power Cables G2026A,'-B, -C, and -F, and Control
Cable G2D26A3J; and Pressurizer LTOP Relief-
Isolation 2CV-4741-1, Power Cable R2B51K2A.
-
Shutdown Cooling Suction Valve 2CV-5084-1, Power
Cable R2851G28.and Control Cable R2851G2E,~and
Shutdown Cooling Suction Valve 2CV-5086-2, Power
Cables G2B62E5L, -M, and -N.
-
RCS Letdown Flow Valve 2CV-4823-2, Control
'
Cable G2S017C, RCS Letdown to Regenerative Heat
Exchanger Valve 2CV-4820-2, Power Cable G2B61L3B
and Control Cable G2861L3C, and RCS Letdown i
Containment Isolation Valve 2CV-4821-1, Power -!
Cable R2B51M1B and Control Cable R2B51M10.
Fire detection is provided.
Automatic fire suppression is provided.
An exemption from the 20-foot horizontal separation.
requirement of Section III.G.2 has been granted by SER
dated March 28, 1983, for electrical distribution
Panels 2RS3 and 2RS4.
Fire Area JJ, Fire Zono 2109-U, corridor and monitor I
control center, was found to be not in compliance with
Appendix R,Section III.G.2.
(c) Fire Area EE, Fire Zone 2111-T, Lower South Electrical
Penetration Room
Fire Zone 2111-T was selected because it contained
redundant associated circuit cabling for the pressurizer
.
LTOP relief valves high/ low pressure interface and the
shutdown cooling suction valves high/ low pressure
interface. The following conditions were found to exist:
Less than 20 feet of horizontal separation exists for
the following redundant cables: !
-
Pressurizer LTOP Relief Isolation 2CV-4731-2,
Power Cable G2B61L2A and Control Cable G2B61L2C;
and Pressurizer LTOP Relief Isolation 2CV-4730-1,
Power Cable R2851E4A.
l
- _ _ - _ _ - _ _ _ _ .
30
!
- Shutdown Cooling Suction Valve 2CV-5084-1, Power i
Cable R2B51G2B and Control Cable R2B51G2E; and
Shutdown Cooling Suction. Valve 2CV-5086-2, Power
Cables G2B62E5L, -M, and -N.
Fire detection is provided.
Automatic fire suppression is provided.
,
Fire Area EE, Fire Zone 2111-T, lower south electrical
penetration room, was found to be not in compliance with !
Appendix R,Section III.G.2.
Th'e licensee's non-compliance with 10 CFR 50, Appendix R, l
Section III.G.2 is summarized in the following table. . i
These items are considered as unresolved pending resolution 1
between the licensee and NRR. (Seealsoparagraph11.b.(1)) j
l
1
1
l
4
I
.
\
-_-_-_- _
. _ .
. _ _ _ _ - _ _ _ _ _ - _ _ _ _ _ _ _ _ - _ _ _ _ _ - _ _ _ _ _ - _ _ - _ _ _ _ _
l
31
TABLE 9.c
SECTION III.G.2 NON-COMPLIANCE SUMMARY TABLE i
l
Fire Fire Redundant Component Type Cable .Non-
Area Zone Component Description Cable Number Compliance
B 209188 2CV469B-1 ECCS Vent Valve DC Motor Starter <20 ft Sep
2CV4740-2 LTOP Isolation DC Motor Starter
JJ 21090 2CV4731-2 LTOP Isolation Power G2861L2A <20 ft.Sep
l
Control G2861L2C
2CV4730-1 LTOP Isolation Power R2851E4A
JJ 21090 2CV4740-2 LTOP Isolation Power G2026A3A <20 ft Sep
Power G2026A3B l
l
'
Power G2026A3C ,
Power G2D26A3F ;
'
l Control G2026A3J
l
2CV4741-1 LTOP Isolation Power R2851K2A
JJ 21090 2CV5084-1 SDC Suction Power R2851G2B <20 ft Sep
Control R2B51G2E
2CV5086-2 SDC Suction Power G2862E5L
Power G2862E5M i
l Power' G2862E5N
JJ 2109U 2CV4823-2 Letdown Flow Control. G25017C <20 ft Sep
2CV4820-2 Letdown RHX' Power G2B61L3B
Control G2861L3C
2CV4821-1 Letdown Isol. Power R2B51M1B
1
Control R2B51M1C
EE 211fT 2CV4731-2 LTOP Isolation Power . G2861L2A <20 ft Sep
Control G2B61L2C
2CV4730-1 LTOP Isolation Power R2851E4A
EE 2111T 2CV5084-1 SDC Suction Power R2851G2B <20 ft Sep
Control R2B51G2E
2CV5086-2 SDC Suction Power G2B62E5L
Power G2B62E5M
Power G2862E5N
1
!
1
i
' !
i
32
9. Procedures (Unit 2)
1
The procedures reviewed during this inspection include a post-control l
room-fire alternate shutdown procedure, a remote shutdown procedure, l
Technical Specification surveillance procedures, fire alarm corrective i
action procedures, plant operating procedures, an emergency plan
'
implementing procedure, and an emergency operating procedure. j
a. Alternate Shutdown Procedure 2203.14, Revision 4
In the event of a fire in any of the areas which results in.
evacuation of the ANO-2 control room, the licensee may shut the i
reactor down from outside the control room in accordance with the 4
Procedure 2203.14, Revision 4, " Alternate Shutdown."
The minimum operating shift complement for Unit 2 consists.of the .
following personnel: l
- itle Control Room Evacuation Assignment
Shift Supervisor 3rd floor technical support center
Shift Administrative 3rd floor technical support center & plant
Assistant library
Shift SR0/ Control Switchgear, EDG, battery rooms
Room Superviror ;
Reactor Operator EFW pump room, various locations per
No. 1 procedure and shift supervisor's direction
Reactor Operator Various locations in auxiliary / turbine !
No. 2 buildings per procedure and shift
supervisor's direction
Waste Control Fire brigade leader, affected unit 1
Operator
Auxiliary Operator Fire brigade member
The first five people above are required to implement the control
'
room evacuation procedure. The shift supervisor is'in charge of the
normal operating shift and he continues in that role upon entry into
the procedure. He also serves as emergency director of the emergency
' response organization until the TSC is formally activated. The waste
control operator and auxiliary operator together with other onshift
personnel will staff the fire brigade. Available additional
operators will be utilized to support the required operators.
Once the decision to evacuate the control room has been made, the
reactor will be tripped from the control room driving in the control
_-_
q
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!
33 i
j
element assemblies for. initial reactivity control. Several other
immediate actions will be attempted from the main control room prior
to evacuation. If unsuccessful, however, all are covered I
procedurally from outside the control room soon after the evacuation, l
with the exception of MSIV's closure. The licensee agreed to ravise 1
the procedure to include the latter action.
Following the control room evacuation, the operating shift personnel
then proceed to their post-evacuation assignments as. outlined above, J
to implement'their assigned procedure sections, attachments, and i
checklists. The remainder of the alternate safe shutdown will be i
l directed and. monitored by the shift supervisor from the technical .)
support center, 3rd floor of the administration building. ]
l i
b. Procedure Walkdown I
The alternate shutdown procedure (2203.14, Revision 4) walkdown was
initiated at 2:05 p.m. on June 11, 1987, using five people from the ,
licensee's operating staff with the proper training and 1
qualifications to fill the control room operating shift positions.
Since the procedure is symptom oriented, the following initial
conditions were presented:
Fire in control room of sufficient. size to require evacuation ,
I
'
Coincident loss of offsite power !
Both EFW pumps fail to auto start l
Both standby diesel generators failed to auto start I
All system lineups as normally found at full power
Two NRC inspectors accompanied the shift supervisor and the shift
i administrative assistant to the 3rd floor TSC. They remained there
l to observe personnel actions, crew direction and leadership,
communications, use of the SPDS, interface with and initiation with
l
and initiation of emergency plan implementation procedure,.and
training and familiarity with the alternate shutdown procedure. Two
other NRC inspectors each accompanied the shift SR0/ control room
supervisor and the Reactor Operator No. I to observe their actions at
the electrical e,uipment rooms and EDGs, at the EFW pump rooms,
observe lighting at work stations and access routes, evaluate
communications, and evaluate training'and familiarity with the
procedure. Reactor Operator No. 2 was not accompanied by an NRC
inspector during the procedure walkdown, so his actions were
monitored from the TSC as he made his call-ins to the shift
supervisor.
The procedure was halted during Step 12 of Shift Supervisor
Section I, which demonstrated that stable hot standby natural
circulation conditions could be achieved and a controlled rate of ,
cooldown was to be commenced. Emergency feedwater flow to the
"B" steam generator via the turbine-driven EFW Pump P7A was
established and associated valve lineups verified within 23 minutes.
_ _ - _ -
m
!
34
l
a
Both emergency Diesel Generators.2K4A and 2K4B were, manually started
and loaded within 18 minutes of start of event. Motor. control'
centers supplying power to the ECCS vent valve (2CV-4698-1) and ,
pressurit er LTOP relief valves (2CV-4731-2, .2CV-4740-2, 'andi .I
2CV-4741-1)weredeenergizedwithin6' minutes..The'operatingstaff'
'
demonstrated adequate training and familiarity with the' procedure
throughout the walkdown.
The only procedural deficiencies identified;were:
(1) Immediate Action No. 6 in Section I.of the' procedure directs
closure of main steam isolation Valves 2CV-1010-1 and 2CV-1060-2
prior to. control room evacuation.- Credit'is given for only one~
manual action prior to evacuation;t typically ~ this is. the reactor d
i manual trip. . The' procedure should be revised to. direct MSIV i
closure from outside the control room if-Irr. mediate Action No. 6
is unsuccessful. -(All other immediate action steps are covered'
later on.in the procedure.) The licensee' agreed to incorporate-
such a revision.
(2) Step 11 in the Shift Supervisor Followup Actions,Section I
directs the tracking of cumulative Coolant : Charging Pump run
time to determine' the amount of boric acid added to the RCS.
The procedures should provide a table to record and track ~ the
run times and should indicate who is to record this information.
The licensee agreed to incorporate such a revision. q
Another improvement to the alternate shutdown procedure would be to
include a note alerting the shift supervisor that he may also use l
"A" steam generator to remove decay heat during cooldown; however, i
the instrumentation monitoring this component is not protected to the H
requirements of 10 CFR 50, Appendix R. If the "A" steam generator .
instrumentation is unavailable, he must make- a judgement as to - !
whether to supply EFW to the "A" steam generator as directed by the ,,
procedure in Step 9 of Reactor Operator No'. 2 Followup Duties
(Section I). Only one steam generator is required for safe shutdown..
It should be noted that the procedural deficiencies and -improvements 3
indicated for the' Alternate Shutdown Procedure 1203.02, Revision 4 '
t
during the AN0-1 inspection in May 1987 were reviewed and. l
incorporated as applicable into the ANO-2 Alternate Shutdown- i
Procedure 2203.14 Revision 4, prior to the ANO-2' inspection in-
June 1987. These included interface direction between Units 1 and 2 ;
control rooms for a control room / cable spreading room fire, equipment 1
checklists for operators at the alternate shutdown equipment locker,
and a radio check step prior to dispatching operators 'to their
alternate shutdown duty locations.. The licensee should be commended
on his responsiveness in incorporating these improvements. ,
l
- -_ _ _- -_
35
i
c. Fire Alarm Corrective Action Procedures and Fire Preplans
i
Procedure 2203.12K, Revision 13. " Annunciation K12 Corrective Action,"
.
and Procedure 2203.09, Revision 5, " Fire Protection System 4
Annunciation Corrective Action," were reviewed to evaluate the train
of responses followed by the control room operator upon receipt of a.
fire annunciation alarm. The procedures were found to adequately
direct the operators to identify the origin of an alarm, verify its
validity, and respond accordingly. The fire preplans were readily
available in the control room to guide the operators in their
response,
d. Technical Specification Surveillance Procedures
Surveillance test procedures covering a sampling of Technical ,
Specification surveillance requirements for various post-fire safe !
shutdown components and systems were reviewed. These included
procedures for valve lineup verification of several safe shutdown
systems, the motor-driven emergency feedwater Pump 2P7B, and the
turbine-driven emergency feedwater Pump 2P7A. The past two
performances of monthly Technical Specification Surveillance
Procedures 2102.01, Revision 22, " Plant Pre-Heatup and Pre-Critical-
Checklist," Attachment E - Category.E Valve Position Verification; l
,
and 2106.06, Revision 21, " Emergency Feedwater System Operations,". i
l Supplement I for 2P7A and Supplement II for 2P7B, were retrieved for i
'
inspection of test results. This review showed the procedures to be
adequate and the performances reviewed were performed on time and i
documented adequately. j
l
e. 0_perator Training on Safe Shutdown Procedures-- l
In addition to observing the operator's performance during the
walkdown of the alternate shutdown procedure, operatioris technical
support and operator training personnel were interviewed concerning
operator training on Appendix R post-fire safe shutdown procedures
and equipment. The training covered Appendix R equipment, Alternate
Shutdown Procedure 2203.14, and the fire protection program manual
and fire preplans. The program includes classroom instruction,
walkdowns, and hands-on operating experience. An annual
requalification cycle will be maintained in accordance with
10 CFR 55, May 26, 1987, edition.
The licensee is presently revising the Appendix R-related training to
enhance the fire pre-plan instruction and to integrate plant-
simulator exercises into the safe shutdown / alternate shutdown
portions of the course. The mechanical systems specialist observed
the performance of one crew on an unannounced alternate shutdown
exercise on the simulator. The personnel involved demonstrated good
knowledge of the plant and the required procedures and responded
adequately to the scenario presented to them. All six shift crews
( _ _ _ _ _ _ _ _ - -
,
,
'
] ;
I
1
36 3
l
1
,
will have completed an alternate shutdown' exercise on the plant
simulator within the next few veeks. J
l
.A general lesson plan utilized to administer Appendix R was provided-
for inspection. Training records for' operating shift personnel were
also reviewed. These areas reviewed were found to be acequate.
10. Protection for Associated Circuits (Unit 2). ,
1
Common Bus Concern
Spurious Signals Concern-
Common Enclosure Concern
a. Common Bus Concern .
The common bus associated circuit concern is found in circuits,.
either nonsafety-related or safety-related, where there is a common ;
power source with shutdown equipment and the power source is not I
electrically protected from the circuit of concern,
1
The common bus concern is made up of two items: I
Circuit Coordination
High Impedance Fault Analysis
(1) Circuit Coordination
Breaker coordination is audited by reviewing the time current
curves developed during the licensee's bus coordination study.
At ANO, Unit 2, the following circuits were randomly selected 1
for review: ;
Circuit Comment
Bus 2A3 Coordination satisfactory
Bus 2A4 Coordination satisfactory
,
Bus 2B5 Coordination satisfactory
Coordination satisfactory
'
'
Bus 2B6
MCC 2B51 Coordination satisfactory ,
MCC 2B61 Coordination satisfactory '
- MCC 2B53 Coordination satisfactory
l MCC 2B63 Coordination satisfactory
2001 Coordination satisfactory
2002 Coordination satisfactory
2RS1 Coordination satisfactory
2RS2 Coordination satisfactory
2RS3 Coordination satisfactory
2RS4 Coordination satisfactory
2RA1 Coordination satisfactory
2RA2 Coordination satisfactory
4
l
_ _ _ _ ___m.__________ ~__
.__. .
H
37 )
2023 Coordination satisfactory i
2024 Coordination satisfactory j
The licensee's circuit coordination program was found to be
satisfactory, i
To ensure that the existing satisfactory circuit coordination is _
not compromised by future design changes, the licensee has an
established procedure for modification design review, 1
Procedure 216,, dated April 30, 1987,." Guidelines for Evaluation- i
of Safe Shutdawn Capability and Control of Safe Shutdown: 1
Capability Assessment.," which prevides for reviewing ~ 1
modification design far Appendix R concerns'.'
The licensee performs relay testing and maintenance at 10-month
intervals (each refueling outage). Circuit breakers are tested
and maintained at intervals of 60 months. Breaker and relay '
)
maintenance and testing.are currently scheduled manually. The
licensee is .in the precess of converting to' an automated
maintenance schedulirs system. ,
1
Maintenance recordslcr the following randomly selected circuit
breakers or protectiva relays were reviewed to verify that
maintenance and'tcsting are being performed at the specified .)
frequency
Required Completion l
Component. Title Frequency Date
BKR 2A113 Startup Number 3 156 Weeks 07/02/86
Feeder BKR
BKR 2A303 Service Water Pump 78 Weeks 07/07/86 I
2P-4B BKR
BKR 2H13 Startup Number 156 Weeks 06/21/86 i'
Feeder BKR
BKR 28-832 2B85 Supply BKR- 208 Weeks '08/07/86 i
The reviewed records documented compliance with established
maintenance procedures.
Control of fuse replacement is required to ensure maintenance of
coordination for circuits protected by fuses. The licensee's ,
controls for fuse replacement include the following: '
Procedure 1403.85, " Motor Control Center Preventive
Maintenance"
Plant Drawings / Prints
- _ _ _ _ _ - _ - _ _ _
,
!
38 j
i
. i
'
Technical. Manuals j
Job Orders-
!
Materials Controls !
The licensee's fuse replacement controls were found to be
satisfactory.
'
l
1
(2) High Impedance Fault Analysis
The high impedance' fault concern is found in the case where
multiple high. impedance faults exist.as. loads on a safe shutdown j
power supply and cause the loss of the safe shutdown power :
supply prior to' clearing the.high' impedance faults.
l The licensee's analysis .for high impedance faults, EE-87-014, i
( dated January 21, 1987, ANO, Units 1 and 2 position on multiple j
high impedance faults, determined that protection for i
simultaneous high ' impedance faults was~ provided by the
following- {
The 4.16 KV safety buses are equipped with ground fault -
relays.
The 4.16 KV and 480 V distribution systems.are grounded.
High impedance faults would rapidly propagate into low l
impedance faults and cause coordinated circuit breaker'
tripping. ]
I
i
In the event that a required electrical bus is lost due to I
high impedance faults', the licensee's procedures provide i
for manually tripping all breakers on the faulted bus and )
reenergizing required safe shutdown loads. ,
The licensee's analysis and protection for high impedance faults
were found to be satisfactory.
The licensee's protection for the associated. circuit common bus f
concern was found to be satisfactory.
i
b. Spurious Signals i
The spurious signal concern is made up of two items: I
The false motor, control, and instrument readings such as
occurred at the 1975 Brown's Ferry fire. These could be caused
by fire initiated grounds, short or open circuits.
l
!
l
a___-____--_-__-----_-____-_---_--__-__- _ _ - - - _ _ _ _ . _ - - _ _ _ - _ _ - _ _ - _ _ _ _ _ _ _ _ - _ - _ - _ - _ - _ - - -
39
Spurious operation of safety-related or nonsafety-related
components that would adversely affect shutdown capability
(e.g. , RHR/RCS isolation valves).
(1) High/ Low Pressure Interfaces
The licensee has identified the following high/ low pressure
interfaces and methods for controlling the interfaces:
Interface Method of Control Status
l High Point Vents Flow path capacity via 2SV-4636-1,
I. 2SV-4636-3, 2SV-4668-1, 2SV-4668-2,
'2SV-4669-1, and 2SV-4670-2 is less than
the definition of a LOCA.
Pressurizer Vent SER dated May 13, 1983, considered that
Valves 2CV-4698-1 spurious operation of this interface was
l and 2CV-4740-2 unlikely. The basis for this conclusion
l was that.for both valves to open, four
I
circuits would have to be spuriously
completed in two locations.
Documentation (schematics) review and
in plant inspection determined that only
l
'
two shorts, one in the control circuit
of each valve, were required to cause
spurious operation of the high/ low
pressure interface. The' interface is
,
i
susceptible to spurious operation for I
control room and cable spreading room ;
fire-induced short circuits. Addition- -i
l ally, since the power. circuits for these y
valves do not have adequate separation- ~
as described in. paragraph 4.3, high/ low
pressure interface protection is not
provided in the north electrical l
equipment room,
RCS Letdown Flow SER dated May 13, 1983, determined that
Valve 2CV4823-2, this interface was not a concern. The i
RCS Letdown to basis for this determination was that
Regenerative Heat the letdown lines are flow restricted
Exchanger Valve lines protected by reliefs and would
2CV4820-2, and RCS require the failure of at least three
Letdown Containment valves to challenge the relief.
Isolation Valve
2CV4821-1 Documentation (schematics) review and
in plant inspection determined that
single shorts in the control circuit of
each valve were required to cause
,
-- _ - . _ - _ - _ - _ _ - _ . _ _ _ - - _ _ _ _ -
q
.4
L
40
.)
<
spurious operation of the high/ low
pressure interface. The interface is .,
susceptible to spurious operation for a
control room and cable. spreading room
~
fire-induced short. circuits. Addition- l
i ally, since the power and control !
circuits for these valves do not have
'
)
adequate separation as described in i
paragraph 4.3, high/ low pressure I
interface protection is not provided in '
the corridor and motor control center zone
and lower south electrical penetration
room.
LTOP Relief Isolation. -This high/ low pressure interface is
Valves 2CV4731-2.and controlled by tagging 2CV4730-1 circuit
2CV4730-1 breaker'open after shutting the valve I
during plant startup.~ 'This method of
control is effective for control room !
and cable spreading room fires. However,
power and control cables for the valves
do not have adequate separation as
described in paragraph 4.3 in the
corridor and motor control. center zone and
the lower south electrical penetration
room, causing the interface to be
unprotected for shorts and hot shorts in
these areas.
LTOP Relief Isolation This high/ low pressure interface is
Valves 2CV4741-1 and controlled by tagging 2CV4741-1 circuit
2CV4740-2 breaker open after shutting the valve
during plant startup. -This method of I
control is effective for control room
and cable spreading' room fires. However,
power and control cables for the valves
do not have adequate separation as
described in paragraph 4.3 in the
corridor and motor control center zone,
causing the interface to be unprotected ,
for shorts and hot shorts in this area. '
Pending resolution of these items between the licensee and NRR.
This will be considered an unresolved item. (368/8714-06)
(2) Current Transformer Secondaries
'
The licensee's current transformer (CT) Analysis EE-86-002,
dated January 10, 1986, ANO, Unit 1, Possible Appendix R
Concerns Due to Open CT Circuits, also applicable to Unit 2,
determined that the CT saturation characteristics limited
L______________.___ _ _ _ . _ _ _
41 .
potential and energy in CT secondaries such that secondary fires
could not be induced by open CT secondaries.
The licensee's protection for current transformer ypen secondary
concerns was found to be satisfactory.
(3) Isolation of Fire Instigated Spurious Signals
The licensee has provided isolation for fire-instigated spurious
ignals by various methods, including:
administrative controls,
rerouting of cables,
'
wrapping cables,
isolation / transfer switches (redundant fuses used),
"
fuses,
signal isolators, and
manual component operation.
During the inspection, all forms of isolation listed above were
observed.
The licensee's methods of fire instigated spurious signal
isolation were found to be satisfactory.
c. Common Enclosure
The common enclosure associated circuit concern is found when
l redundant circuits are routed'together in a raceway or enclosure and
l they are not electrically protected, or fire can destroy both
circuits due to inadequate fire protection means.
Licensee representatives stated that:
, Redundant safe shutdown cables are never routed in common
I
enclosure.
Nonsafety-related cables routed in common enclosure with
redundant safety-related cables-are never routed between
redundant trains.
All circuits are electrically protected.
During the inspection, the following randomly selected nonsafe
shutdown cables routed in common enclosure with safe shutdown cabies
were verified to be electrically protected:
Component Cable Number Location Protection
2CV8831-1 R2853A4A Raceway EB156 Circuit _ Breaker 2853A4
2CV1074-1 R2853C1C Raceway EB156 Circuit Breaker 2B53C1
_ _ _ - -
-_- - _ - _ _ _ _ _ _ - _. _.
--.
42
2VUCM-68-1 R2853K4A Raceway EB156 Circuit Breaker 2B53K4
2FS-8678-2 G2863C3E Raceway EC275 Control Power Fuse
2RE-8271-2 G2S110L Raceway EC275 Circuit Breaker R2RS1-13
2C184 G25124C Raceway EC275 Control Power Fuse
The licensee's protection for the common enclosure associated circuit
~
concern was found to be satisfactory.
11. Communications (Units 1 and 2)
The licensee has identified three communications systems available for
safe shutdown: por.able hand-held radios, the plant dial telephone
system, and the Gaitronics system. The primary means of communications
during alternate shutdown are the portable hand-held radios available in
the control room and the alternate shutdown radio cabinet. Plug-in
headsets are provided for the radios for ease of use and improved
background noise rejection. During the Alternate Shutdown procedure
walkdown hand-held radios with headsets were used for communications and
were found to be adequate.
The AP&L radio / communications group performs an annual frequency check and
functional check of all the radios. The repeater for the portable. radio
is powered by the security diesel and located on top of the Administration l
Building. A recent security upgrade of the in plant radio system to
provide a coaxial antenna for improved reception throughout the plant
incorporated the requirements for the Appendix R alternate shutdown
equipment locations into its design. The system has been thoroughly
tested to verify its performance in those areas where the portable radios
are required to provide safe shutdown communications.
Additionally, the plant dial telephone system (PAX) and plant announcing ;
system (GAITRONICS) may be available to support safe shutdown
communications. The licensee does not take credit for these systems.
12. Other Electrical Evaluation (Units 1 and 2)
a. Cable Routing
Documentation (cable routing) review and physical in plant inspection '!
were performed on the following:
Component Type Cable
2P36A,B,C Power
2CV-4920-1 Power and Control
2CV-4873-1 Power and Control
2CV-4921-1 Power and Control
2CV-4950-2 Power and Control
2CV-4741-1 . Power and Control
2CV-4740-2 Power and Control
2CV-4731-2 Power and Control ,
. _ - - -
~
)
4
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2CV-4730-1 Power and Control i
2P60A,B Power
'2P4A,B,C Power
2P16A,B Power :
2VEF24A,B,C.D Power- i
2RA1,2' Power ;
I 2021,22,23,24 Power
l SPDS Power i
Control ;
2CV-1026-2 Control '
2CV-1025-1 Control
'2CV-1037-1 Control ;
2CV-1036-2 Control :
2CV-1076-2 Control i
2CV-1075-1 Control 1
2CV-1039-1 Control
2CV-0798-1 Control
2CV-0714-1- Control
2PT-4624-1 Instrumentation. ]
2PT-4624-2 Instrumentation
2TE-4614-1 Instrutnentation
2TE-4714-2A Instrumentation ;
2TE-4611-3B ' Instrumentation -l
2TE-4716 Instrumentation
2LT-4627-1 Instrumentation
2LT-4627-2 Instrumentation
2PT-1141-1 Instrumentation
2PT-1141-2 Instrumentation '
2LT-1179-1 Instrumentation
2LT-1179-2 Instrumentation
2JE-9000-1 Instrumentation i
2JE-9003-2 Instrumentation j
Except for the components discussed in paragraph 8.c, the routing of
the cables for the above components was found to be in compliance
with Appendix R,Section III.G separation requirements. !
The licensee controls / tracks cables using Procedure 216, Revision 1,
dated April 30, 1987, " Guideline for Evaluation of Safe Shutdown
,
Capability and Control of Safe Shutdown Capability Assessment." This
l procedure provides the following forms for maintaining the cable data
j base:
Form Description
216 F2 Used to add new shutdown components.and their related
cables to the data base. '
I
I
216 F3 Use to add new shutdown cables and~ associated cable
data.
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The licensee's control-of cables was found to'be satisfactory.
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b. Review of Unit 1 Cable Routing Open Item
During the May 1987, Unit 1, Appendix R inspection,-cable ,
i separation / routing was found to be an unresolved. item pending 1
l. documentation review and physical in plant inspection. The following
circuits were inspected during the week of the June 1987 Appendix R
inspection for Unit 2 to clear the Unit I cable routing unresolved
item:
Components Type Cables-
CV1407 and CV1408 (BWST) Power and Control j
j
CV1219 and CV1220 (HPSI) Power and Control-
PSV1000 and CV1000 (PZR PORV) Power and Control
CV1228 and CV2618 (SG ATMOS) Power and Control :
l (and associated block valves) j
P4A/B/C (Service Water Pumps) Power and Control l
LT1001 and LT1002 (PZR Level) Instrumentation
i
PT1042 and PT1041 (RCS Press) Instrumentation i
TE1144 and TE1147 (TCS Temp) Instrumentation-
'
LT2620 and LT2624 (SG Level) Instrumentation
NE 501 and NE 502 (Source Range)- Instrumentation i
i
LT4204 and LT4205 (CST Level) Instrumentation i
Redundant Components in Fire Power and Control 4
Area B requiring manual operation
Documentation review and physical in plant inspection for Unit 1
cable separation / routing was completed satisfactorily. This closes
the unresolved item from the May 1987, ANO-1 inspection (see
paragraph 7.d).
c. Modification Review
The licensee's process for controlling the design and installation of ,
modifications was reviewed for proper review and approval, including !
10 CFR 50.49 aspects.
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The control of modification design is governed by Procedure 202,
Revision 10, dated November 7, 1986, " Design Process," which provides
for design review for Appendix R concerns.
The administration of modifications for Appendix R concerns was .found
to be satisfactory. >
13. Unresolved Item
.
.
1
An unresolved item is a matter about which more information is r.equired in
order to determine whether it is.an acceptable item, a violation,, or a
deviation. Three unresolved items are discussed in paragraphs 3, 8.a(5), !
and 10.b(1) of this report. l
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14. Exit Interview
An exit interview was conducted on June 12, 1987, with those personnel
denoted in paragraph 1 of this report. lAt this meeting, the scope of the
inspection and the findings were summarized.
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21 $
0.
AlTACHMENT TO ,
,
NRC-INSPECTION REPORT
50-313/87-14; 50-368/87-14' '
APPENDIX C .l
DOCUMENTS REVIEWED
r ,
A. Letters, Reports, and Procedures j
I
Date Number Rev. Title j
- t
08/30/85 Letter -
J.T. Enos (AP&L) to J.F. Stolz ]
'(NRC) & E.J. Butcher'(NRC), Results
of Reanalysis Against.NRC Clarifi-
cation / Interpretation of Appendix R' )
to 10 CFR 50 Supplemental Informa- l
tion j
02/23/83 Letter -
J.R. Marshall (AP&L) to J.F. Stolz
(NRC) & R.A. Clark (NRC), Hot.to- i
Cold Shutdown Scenario for Loss of i
Offsite Power-Exemption Request l
Details from Appendix R compliance '
submittals
05/11/83 Letter -
J.F. Stolz (NRC) to J.M. Griffen
(AP&L) concerning exemption to
Appendix R to 10 CFR 50, 4
Section III L.I for ANO-1 l
l
l 02/25/83 Memorandum -
L.S. Rubenstein (NRR) to
G.C. Lainas (NRR) .SER Supplement
for Appendix R to 10 CFR 50,
Sections III G. and III L. and ANO,
Units 1 and 2 l
l
11/15/82 Memorandum -
L.S. Rubenstein (NRR) to !
Appendix R to 10 CFR 50,
, Sections III G. and III L. - ANO, j
l Units 1 and 2
03/22/83 Letter -
R.A. Clark (NRC) and J.F. Stolz
(NRC)~to J.M.'Griffen'(AP&L)
concerning Exemptions to certain
requirements of Appendix R to ,
10 CFR 50 for ANO-1 and -2
05/04/87 1203.02 24 Alternate Shutdown
02/15/86 1102.10 27 Plant Shutdown and Cooldown
03/14/86 1203.13 6 Natural Circulation Cooldown ,
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12/16/86 1102.02 36 Plant Startup
FSAR Amend 4 Final Safety Analysis Report -
ANO-1
03/86 FPPM 2 Fire Protection Program Manual
05/21/74 Tech Spec Amend 106 Technical Specifications-
11/06/86 1000,06A 2 Remote Shutdown
'0/18/84 Memo ANO Unit 1 Training Scope of. Licensed'
84 11379 Requalification Cycle 4 8/28/84 to
9/28/84
11/13/84 Memo ANO Unit-1 Training Scope of Licensed
84 12620 Requalification Training Cycle 5
10/02/84 to 11/02/84
06/19/85 Memo ANO Unit 1 Operations Training ANO-
85 08359' Alternate Shutdown Procedure
Training
09/27/85 Memo ANO AND Cycle 4 Requalification
85 12245 Training Alternate Shutdown
10/03/85 Memo AN0 AN0 Cycle 4 Requalification-
85 12399 Training Alternate Shutdown
10/11/85 Memo AN0 AND Cycle 4 Requalification
85 12559 Training Alternate Shutdown !
10/21/85 Memo AN0 AN0 Cycle 4 Requalification
85 12862 Training Alternate Shutdown
10/28/85 Memo AN0 ANO Cycle 4 Requalification
85 13013 Training Alternate Shutdown
11/05/85 Memo ANO AND Cycle 4 Requalification
85 13243 Training Alternate Shutdown
08/31- AA-99999-002 Training Attendance Records:
09/28/84 10 CFR, Appendix R-Related DCPs
10/05/84- AA-21003-004 Training Attendance Records:
01/24/85 AA-21001-008 Alternate Shutdown Procedure
07/26- AA-21003-004 0 Training Attendance Records:
10/31/85 Alternate Shutdown Procedure
l _ _ .- . __- -___ __ - _ - - -
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11/07- AA-21001-008 Training Attendance Records: {
12/02/85 10 CFR 50, Appendix R-Required J
Equipment i
11/08- AA-21002-028 0 Training Attendance Records: Fire
, 12/13/85 Protection Manual
10/31/85 AA-51002-028 2 Plant Systems Training
09/29/83 AA-51001-008 1 Shift Administration l
01/02/87 1203.12K 20 Annunciator K12 Corrective Action
01/13/86 1203.09 Fire Protection System Annuncie. tor
Corrective Action +
12/05/84 1104.02 22 Makeup and Purification System
Operation
03/26/87 1106.06 31 Emergency Feedwater Pump 3peration
05/06/87 87E-0011-01 Calculation: Makeup Pump Room
Temperature with No Cooling Fan
10/15/85 85E-00081-01 Calculation: Decay Heat Vault
Temperature & Rate of Heat Rise
Assessment During Normal Decay Heat
Operation without Cooling
03/31/87 85E-00070-01 Calculation for Cooldown Without ,
Pressurizer Heaters !
l
06/06/86 Memo Appendix R Treatment of HPI !
NEL-066-25 Auxiliary Lube Oil Pumps (P64s)
03/31/87 85E-00071-01 Calculation: Evaluation of Time
Available for Manual Initiation of
EFW following Loss of All Feedwater
01/21/87 EE-87-014 ANO Units 1 & 2 Position On
Multiple High Impedance Faults
01/10/86 EE-86-002 A'N0 Unit 1 Possible Appendix R
Concerns Due to Open CT Circuits
12/05/84 1307,07 1 Testing & Maintenance of Metal Clad
Switchgear & Breakers
05/08/86 1403.81 1 Type K-225 Low Voltage Breakers
05/08/86 1403.82 1 Type K-600 Low Voltage Breakers
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05/08/86 1403.83 1 Type K-1600 Low Voltage Breakers-
05/29/86 1403.150 5. Relay Calibration for WA3 Bus
05/29/86 1403.151 1 Relay Calibration for DG1 Diesel
Generator
11/07/86 202 10 Design Process Procedure
04/30/87 216 1 Guideline for Evaluation of Safe
Shutdown Capability and Control.of 1
Safe Shutdown Capability Assessment
02/27/87 123 1 Forms Control'
)
1
02/02/87 J0#00728878 Fuse F10 in 20108 Replacement "
10/31/84 DCP 84-1061 Control Isolation for EDG Room
Exhaust Fans )
06/15/84 DCP 83-1012 CV-1000 Modification
08/20/82 GCP 82-2079 Install EDG Cross Connect Emergency '
Diesel Fuel Storage Building
06/21/84 DCP 83-1008 Provide Disconnect Switches for "B"
Swing Makeup Pump Appendix R ;
08/18/82 OCAN088204 ANO Units 1 & 2 Additional' l
Information Concerning Reactor
Coolant System High Point' Vents
06/02/87 1107.01 27 Electrical System Operations
08/18/82 OCAN088204 Additional Information Concerning
AP&L Letter Coolant System High Point Vents
05/13/83 05831245 ANO Units 1 and 2 Fire Protection
10/27/83 2307.30 PC-2 Testing'and Maintenance of Metal <
Clad Switchgear and Breakers Type '
AM-7.2-500-6H 1200 & 2000 Amperes
12/12/84 1403.82 PC-1 Type K-600 Low Voltage Breakers
10/27/83 2307.27 PC-1 Testing and Maintenance of Metal
Clad Switchgear & Breakers Type
AM-4.16-250-8H.1200 Amperes
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04/30/86 2403.65 2 Relay Calibration for 2B Load
Centers
02/11/87 2101.02 26 Plant Startup
03/21/86 1403.85 2 Motor Control Center Preventive l
Maintenance
01/21/87 EE-87-014 ANO Units-1 & 2 Position'on
' Multiple High Impedance Faults
04/30/87 216 2 Guidelines for Evaluation of Safe l
Shutdown Capability and Control of 1
Safe Shutdown Capability Assessment ']
01/06/84 0CNA018401 ANO. Units 1 & 2 Fire Protection j
Safety Evaluation Report l
03/22/83 OCNA038328 ANO Units 1 & 2 Fire Protection
Safety Evaluation Report
01/14/87 2203.14 4 Alternate Shutdown ;
06/26/86 2102.10 17 Plant Shutdown and'Cooldown ;
i
01/17/86 2203.13 4 Natural Circulation Cooldown
05/03/85 2203.30 1 Remote Shutdown
SAR Amend 4 Safety Analysis Report - ANO-2
12/11/86 Tech Spec Amend 81 Technical Specifications
08/12/86 AA-22003-004 Training Attendance Report:
Alternate Shutdown Procedure and
Walkthrough
08/17/86 AA-22003-004 Training Attendance Report:
Alternate Shutdown Procedure and '
Wal kthrough
08/26/86 AA-22003-004 Training Attendance Report:
Alternate Shutdown Procedure and
Walkthrough
09/05/86 AA-22003-004 Training Attendance-Report: ;
Alternate Shutdown Procedure and !
Walkthrough
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09/09/86 AA-22003-004 Training Attendance Report:
Alternate Shutdown Procedure and'
Walkthrough
09/16/86 AA-22003-003 Training Attendance Report:
Alternate Shutdown Procedure A0P .
'
2203.14
I
04/29/87 2203.09 5 Fire Protection System Annunciator d
Corrective Action
10/11/85 2102.04 11 Power Operation
03/23/87 2106.06 21 Emergency Feedwater System .]
,
Operations
]
l 10/29/86 AA-62002-008 2 Shift Administration
l 09/20/86 2203.12K 13 Annunciator 2K11 Corrective Action
l
WP870089 ANO Unit 2 Emergency Operating
l Procedure Technical Guidelines-
05/18/87 Memo Ventilation Requirements for
NEL-057-25 Electrical Equipment Room, ANO-1
& -2, Appendix R
09/21/77 Letter D.A. Rueter (AP&L) to D.K. Davis
and'J.F. Stolz (NRC)i ANO, Units 1
& 2, Docket Nos. 50-313 & 50-368,
License No. DPR-51 Fire Protection
08/30/77 Letter D.A. Rueter (AP&L) to J.F. Stolz &
D.K. Davis (NRC): ANO, Units 1 &
l 2, Docket Nos.- 50-313 & 50-368,
1
Fire Hazards Analysis & Miscellan-
eous Fire Protection Submittals
10/15/85 85E-00095-01 Calculation: Room 2091 Temperature
Assessment After Fire, Appendix R
Fire Evaluation
WP870173 4 Basis Document for Alternate
Shutdown Abnormal Operating
Procedure 2203.14, Revision 4, ANO
Unit 2
01/07/87 85E-00071-02 Calculation: Evaluation of Time
Available for Manual Initiation of
EFW following Loss of All_ Feedwater
L-------------------_---_--------_----------._--.------.----- - - - -__---__--__--------z
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.
7
85E-0087-1 2 Safe Shutdown Capability Assessment
(ANO, Unit 2)-
02/11/86 85E-0122 0 Evaluation of'ANO Radio System
Suitability for Alternate Shutdown
l Communications
l B. Drawings - Mechanical
Date Number Rev. Title
01/17/86 M-202 50 Piping & Instrument' Diagram Main
Steam
01/29/87 M-204 Sht Piping & Instrument Diagram
3 of 4 10 Emergency Feedwater
M-204 Piping & Instrument Diagram
Sht 5 0 Emergency Feedwater Storage
01/28/87 M-206 Sht Piping & Instrument Diagram Steam
1 of 2 Peal Generator Secondary System
11/05/86 M-209 35 Piping & Instrument Diagram Cond'r.
Vacuum, Circ. Water & Intake l
Structure Equipment
03/02/87 M-210 44 Piping.& Instrument Diagram Service q
Water j
i
12/10/86 M-217 49 Piping & Instrument Diagram )
Emergency Diesel Generator & Fuel
Oil System
12/04/86 M-230 Sht Piping & Instrument Diagram Reactor
1 of 2 53 Cooling System
12/04/86 M-230 Sht Piping & Instrument Diagram Reactor
2 of 2 53 Cooling System
01/10/87 M-231 Sht Piping & Instrument Diagram Makeup
1 of 2 47 & Purification System
01/06/87 M-231 Sht Piping & Instrument Diagram Makeup.
2 of 2 7 & Purification System
12/13/86 M-232 35 Piping & Instrument Diagram Decay
Heat Removal System
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11/24/86 M-263 Piping &. Instrument Diagram HVAC ;
Sht 3 'O Auxiliary Building Miscellaneous !
Rooms
01/06/87 M-2202'Sht. Piping & Instrument Diagram Main .i
2 of 2 39 Steam 1
01/12/87 M-2204 Sht Piping & Instrument Diagram ,
4 of 4 24 Emergency Feedwater
01/06/87 M-2206 Sht Piping & Instrument Diagram Steam
1 of 2 75 Generator Secondary System
l
l 05/21/87 M-2210 Sht Piping & Instrument Diagram Service l
l
1 of 3 32 Water System
10/18/87 M-2210 Sht Piping & Instrument Diagram Service
2 of 3 Water System
07/07/87 M-2210 Sht Piping & Instrument Diagram Service
3 of 3 36 Water System
01/17/87 M-2212 34 Piping & Instrument Diagram Makeup
Water Demineralization System j
l
03/04/87 M-2230 44 Piping & Instrument Diagram Reactor !
Coolant System
i 01/15/87 M-2231 Sht Piping & Instrument Diagram
1 of 2 54 Chemical & Volume Control System
I
01/15/87 M-2231 41 Piping & Instrument Diagram Safety
Injection System j
01/06/86 M-2232 54 Piping & Instrument Diagram Safety
Injection System )
01/07/87 M-2236 52 Piping & Instrument Diagram
Containment Spray System
01/03/87 M-2263 1 Piping & Instrument Diagram Air
Flow Diagram HVAC Auxiliary
Building - Miscellaneous Rooms
C. Drawings - Electrical
Date Number Rev. Title
12/21/84 2A309-2 0 AN02-4.16 KV Ground Fault Relays
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12/21/84 2A310-1 0 AN02-4.16 KV SWGR Breaker 2A309 ;
12/21/84 2A408-1 0 AN02-4.16 KV SWGR Breaker 2A408 !
12/21/84 2A409-1 0 AN02-4.16 KV SWGR Breaker 2A409
12/21/84 2A409-2 0 AN02-4.16 KV Ground Fault Relays .
12/21/84 28512-1 0 AN02-480 Volt Load Center -
Coordination on Main Breaker 28512
12/21/84 2B612-1 0 AN02-480 Volt Load Center -
Coordination on Main Breaker 2B612
09/27/85 E2006 15 Low Voltage Safety Systems Power j
Supplies Single Line Diagram -l
11/18/84 2B521-2 0 AN02-480 Volt Load Center MCC
Breaker 2B521 J
11/18/84 28621-2 0 AN02-480 Volt Load Center MCC
Breaker 2B621
12/21/84 2001-1 0 AN02-125 VOC Breaker 72-0100
Coordination
12/21/84 2002-2 0 AN02-125 VOC Breaker 72-0211
Coordination
08/05/85 2RS1-1 1 AN02-120 VAC Panel 2RS 1
Coordination
08/05/85 2RS1-2 1 AN02-120 VAC Panel 2RS-1
Coordinatico
08/05/85 2RS2-1 1 AN02-120 VAC Panel 2RS-2
Coordination
08/05/85 2RS2-2 1 AN02-120 VAC Panel 2RS-2
Coordination
08/05/85 2RS3-1 1 AN02-120 VAC Panel 2RS-3
Coordination
08/05/85 2RS3-2 1 AN02-120 VAC Panel 2RS-3
Coordination
08/05/85 2RS4-1 1 AN02-120 VAC Panel 2RS-4
Coordination-
__ _ _- - _
.10 j
08/05/85 2RS4-2 1 AN02-120 VAC Panel 2RS-4
Coordination
12/29/86 E-2566' 4 . Schematic Diagram Pressurizer Vent
Valve
09/04/86 E-2302 Schematic Diagram Pressurizer
Sht 2 Po2 Relief Valves
09/04/86 E-2303 Schematic Diagram Pressurizer
Sh 1 Po2 Relief Valves
12/21/84 2001-0122-1 0 AN02-125 VOC Breaker 72-0122
12/21/84 2001-0133-1 0 AN02-125 VDC Breaker 72-0133
04/23/87 E-2891 31 Conduit & Tray Layout Containment
Penetration Area 24 Partial Plans
05/13/87 E-2867 58 Conduit & Tray Layout Containment
Auxiliary Building Area 24
03/20/85 E-2209 14 Schema' tic Diagram Shutdown Cooling .
Return Header Isolation Valve L
2CV5084-1 & 2CV5086-2
12/10/85 E-2252 6 Schematic Diagram Letdown
Containment Isolation Valve
2CV4823-2
l 09/04/86 E-2233 P1 Schematic Diagram Regenerative Heat
7
Exchanger Inlet 2CV4821-1
10/08/84 A308-1 0 ANO-1, 4.16 KV DG1 Breaker
.J
10/08/83 A309-1 0 ANO-1, 4.16 KV Breaker A309
10/08/84 A408-1 0 ANO-1, 4.16 KV DG2 Breaker 1
10/08/83 A409-1 0 ANO-1, 4.16 KV Breaker A409 .
!
09/07/84 A309-2 0 ANO-1, 4.16 KV Breaker A309 1
1
09/07/84 A409-2 0 AN0-1, 4.16 KV Breaker A409
11/20/84 BS12-1 1 ANO-1, 480 Volt Load Center BS
11/20/84 B612-1 1 ANO-1, 480 Volt Load-Center B6
10/26/84 RS-1 0 ANO-1, 120 VAC RPS & ESF Panel RS1
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10/26/84 RS-2 0 ANO-1, 120 VAC RPS & ESF~ Panel RS2
10/29/84 001-1 0 AND-1, 125 VDC Load Center D01
10/29/84 002-2 0 ANO-1, 125 VOC Load Center 002
'06/19/86 8532-2 0 ANO-1, 480 Volt MCC Breaker B532
Downstream Coordination
11/24/84 B614-2 0 ANO-1, 480 Volt MCC Breaker 8614
Downstream Coordination
11/04/86 E-1 15 Station Single Line Diagram
11/13/86 E-5 12 Single Line Meter & Relay Diagram
4160 Volt System Engineered.
Safeguard
11/03/86 E-8 12 Single Line Meter & Relay Diagram
480 Volt Load Centers Engineered-
Safeguard & Main Supply
'
09/17/85 E-9 22 Single Line Diagram 480 Volt Motor
Control Centers 871, B24, B72, B44
12/12/86 E-15 29 Single Line Diagram 480 Volt Motor
Control Centers B51 & B52
12/11/86 E-16 19 Single Line Diagram 480 Volt Motor
Control Centers 855 & B56
09/29/86 E-17 20 Single Line Meter & Replay Diagram
125 VDC System
01/25/87 E-18 Pul Single Line Diagram 480 Volt Mctor
Control Centers B61 & B62
12/12/86 E-19 .23 Single Line Diagram 480 Volt Motor
Control Centers B53 & B63
07/24/86 E-22 37 ENGINEERED SAFEGUARD & 125 Volt DC
Power Distribution Panels
12/05/86 E-24 3 Single Line Diagram 125 Volt.DC
Motor' Control Center D15 & D25
12/29/83 E-203 6 Schematic Diagram Pressurizer
Proportional Heater Control
04/20/75 M2012-14 12 Electrical Schematic Generator ,
Control Panel J
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12
'10/01/86 E-667 43 Conduit & Tray Layout Auxiliary
Building Area 4 Plan Sh 1
09/23/86 E-667 6 Conduit & Tray Layout Auxiliary
Building Area 4 Plan Sh 2
03/09/87 E-667 9 Conduit & Tray Layout Auxiliary
Building Area 4 Plan Sh 3
12/29/86 E-667 10 Conduit & Tray Layout Auxiliary
Building Area 4 Plan Sh 4
,
03/13/86 E-661 42 Conduit & Tray Layout Auxiliary
'
Building Sh 1
03/05/87 E-258 0 Wiring' Block Diagram Emergency
Feedwater: Initiation & Control
(EFIC) Sh 1A
03/05/87 E-258 0 Wiring Block Diagram Emergency
Feedwater Initiation & Control
(EFIC) Sh 1B
,
09/04/86 E-84 3 Schematic Diagram Typical 480 Volt
Motor Control Centers FVR Starters,.
Interposing Relays Sh 5
01/10/87 E-199 P2
3 Schematic Diagram Reactor Coolant
System MOVs-
l 02/18/86 E-260 0 Wiring Block Diagram Reactor
Nuclear Instrumentation Source
Range Detector Sh 6
02/18/86 E-260 0 Wiring Block Diagram Reactor
Nuclear Instrumentation Source
Range Detector Sh 7
01/12/87 E-262 3 Wiring Block Diagram Plant
Auxiliary Control System Reactor
Coolant Sh 1A
09/03/86 E-331 Po2 Schematic Diagram Miscellaneous
Instrumentation Shs 40 & 41
05/04/82 E-2232 P131 Schematic Diagram Letdown Line Stop~
MOV 2CV4820-2
10/14/82 E-2703 P121 Schematic Diagram Instrumentation ,
Sh 1 Pressurizer Pressure Protection-
1
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10/10/8a E-2703- 13 Schematic Diagram Instrumentation
Sh 2 Pressurizer Pressure Protection
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