ML20133E036
| ML20133E036 | |
| Person / Time | |
|---|---|
| Site: | Mcguire, McGuire  |
| Issue date: | 11/28/1984 |
| From: | Conlon T, Hunt M, Madden P, Miller W, Taylor P, Wiseman G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20133D904 | List: |
| References | |
| 50-369-84-28, 50-370-84-25, NUDOCS 8507220357 | |
| Download: ML20133E036 (33) | |
See also: IR 05000369/1984028
Text
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UNITED STATES
p p MEuq*o NUCLEAR REGULATORY COMMISSION
y* *
, REGION 11
g j 101 MARIETTA STREET.N.W.
- g ATLANTA. GEORGI A 30323
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Report Nos.: 50-369/84-28 and 50-370/84-25
Licensee: Duke Power Company
422 South Church Street
Charlotte, NC 28242
Docket Nos.: 50-369 and 50-370 License Nos.: NPF-9 and NPF-I7
Facility Name: McGuire
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Inspection Conducted: September 24 - 28, 1984
Inspectors:
M. D. unt
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Date Signed
P. M. Madden V
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Date Signed
W k fNaN $y p-20-a+
W. H. Miller Jr.
Q Date Signed
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-P.A.Taylorf Date Signed
Elenum A~ __ //~ A R ~ f f
l G. R. Wiseman Date Signed
Accompanying Personnel: T. E. Conlon, NRC Region II
A. R. Herdt, NRC, Region II
R. Anand, NRC/NRR - ASB
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J. F. Stang, NRC/NRR - CHEB
J. Wilson, NRC/NRR - ASB
A. Coppola, Brookhaven National Laboratory
H. J. Thomas, Brookhaven National Laboratory
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Apprqvedb /Pev v // ,2P- FY
T. E. Conlon, Section Chief Date Signed
Engineering Branch
Division of Reactor Safety
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) SUMMARY
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Scope: This special, announced inspection entailed 350 (70 contractor and 280
NRC) inspector-hours on site in the areas of fire protection, standby shutdown
system (SSS) and related features required to meet 10 CFR 50 Appendix R,
Sections III.G, III.J III.L and III.O.
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G 05000369
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Results: Of the area inspected, six apparent violations were
identified: inadequate or failure to provide fixed suppression system in
accordance with 10 CFR 50, Appendix R, Section III.G.3 for rooms, area, and zones
under consideration paragraph 5.a; failure to provide adequate br eaker/ fuse
protection for equipment required for hot standby paragraph 5.b; failure to
comply with the requirements of 10 CFR 50, Appendix R, Section III.J. -
paragraph 7.a.; failure to provide automatic fire detection for, and fire
barriers to separate, safety-related pumps paragraph 8.a; structural steel fire
barrier supports not provided with fire resistant rating equivalent to the fire
barrier paragraph 9.a; inadequate Appendix R, Section III.G, fire pr;tection
features and separation provided for redundant trains of normal shutdewn systems
and the standby shutdown system - paragraph 9.b. Two deviations were
found: failure to provide battery powered hand lanterns in the control room -
paragraph 7.a.; Failure to provide adequate radio communications between local
control stations and the standby shutdown facility paragraph 8.d.
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REPORT DETAILS
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1. Licensee Employees Contacted
- J. V. Almond, Safety Supervisor
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- T. A. Belk, Engineer Associate
- H. D. Brandes, Design Engineer
- J. M. Bugs, Design Engineer
- K..S. Canady, Manager, Nuclear Engineering Service
- R. Gill, Licensing Engineer
- *A. D. Harrington, Training and Safety Coordinator
- J. R. Hendricks, Principle Design Engineer
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- D. B. Hyde, Associate Engineer
- J. A. Keane, Associate Engineer
, *D. P. Kimball, Associate Engineer.
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- T. A. Ledford, Superintendent, Design Engineering
- T. V. Lyerly, IAE Staff Coordinator
! *W. N. Matthews, Design Engineer
.( *S. H. McInnis, Compliance
M. D. McIntosh, Station Manager
- D. Mendezoff, Licensing Engineer
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- J. A. Oldham, Design Engineer
i *D. J. Rains, Superintendent of Mair.tenance
- W. O. Reeside, Associate Engineer
i *R. W. Revels, Design Engineer
- *N. Rutherford, Licensing Engineer
- B. Travis, Operations Engineer
- G. Vaughn, General Manager, Nuclear Stations
i *L. E. Weaker, Superintendent Station Services
j *C. H. Whitmore, Senior Designer
NRC Resident Inspectors
I *W. T. Orders
! *R. C. Pierson
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- Attended exit interview
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2. Exit Interview
! The inspection scope and findings were summarized on September 28, 1984,
q with those persons indicated in paragraph 1 above. The following inspection
findings were identified to the licensee:
a. Violation Item '(369/84-28-01 and 370/84-25-01), Inadequate or Failure
! to Provide Fixed Suppression Systems in Accordance With 10 CFR 50,
1 Appendix R, Section III.G.3 for Rooms, Areas, or Zones Under
! Consideration paragraph 5.a.
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b. Unresolved Item (369/84-28-02 and 370/84-25-02), Inadequate Fixed Fire
Suppression System Provided for the Cable Spreading Room and Battery
Room paragraph 5.a.(2)(b).
c. Violation Item (369/84-28-03 and 370/84-25-03), Failure to Provide
Adequate Breaker / Fuse Protection for Equipment Required for Hot
Standby paragraph 5.b.(1).
d. Violation Item (369/84-28-04 and 370/84-25-04), Failure to Comply With
the Requirements of 10 CFR 50, Appendix R, Section III.J -
paragraph 7.a.
e. Deviation Item (369/84-28-05 and 370/84-25-05), Failure to Provide
Battery Powered Hand Lanterns in the Control Room paragraph 7.a.
f. Inspector Followup Item (369/84-28-06 and 370/84-25-06), Inadequate
Surveillance Testing Procedures for Emergency Lighting paragraph 7.b.
g. Violation Item (369/84-28-07 and 370/84-25-07), Failure to Provide
Automatic Fire Detection for, and Fire Barriers to Separate
Safety-Related Pumps paragraph 8.a.
h. Unresolved Item (369/84-28-08 and 370/84-25-08), Adequacy of Power
Supplies for Fire Pumps A, B, and C paragraph 8.c.
1. Deviation Item (369/84-28-09 and 370/84-25-09), Failure to Provide
Adequate Radio Communications Between Local Control Stations and
Standby Shutdown Facility paragraph 8.d.
J. Violation Item (369/84-28-10 and 370/84-25-10), Structural Steel Fire
Barrier Supports Not Provided With Fire Resistant Rating Equivalent to
the Fire Barrier paragraph 9.a.
k. Violation Item (369/84-28-11), Inadequate Appendix R, Section III.G,
Fire Protection Features and Separation Provided for Redundant Trains
of Normal Shutdown Systems and the Standby Shutdown System -
paragraph 9.b.
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1. Unresolved Item (369/84-28-12 and 370/84-25-12), NRR Evaluation of
Appendix R, Deviation Request paragraph 9.c.
m. Inspector Followup Item (369/84-28-13 and 370/84-25-13), Amplify and
Clarify Certain Steps of OP/0/A/6100/17 - paragraph 5.q.(2)(a),
n. Unresolved Item (369/84-28-14 and 370/84-25-14) Analysis Effects on
SSF/SSS Operation, Potential Excessive Feedwater to Steam Generators -
paragraph 5.c.(2)(a)(7),
o. Unresolved Item (369/84-28-15 and 370/84-25-15), Correct Procedure l
Deficiencies used to Accomplish 10 CFR 50, Appendix R, Section III.L -
paragraph 5.c.(2)(b).
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3. Licensee Action on Previous Enforcement Matters
This subject was not addressed in the inspection.
4. Unresolved Items
Unresolved ' items are matters about which more information is required to
determine whether they are acceptable or may involve violations or devia-
tions. New unresolved items identified during this inspection are discussed
-in paragraphs 5.a(2)(b), 5.c.(2)(a)(7), 5.c.(2)(b), 8.c. and 9.c.
5. Compliance With 10 CFR 50, Appendix R Sections III.G. and III.L
- An inspection' was conducted to determine if the fire protection features
provided for structures, systems, and components important to safe shutdown
at McGuire were in compliance with 10 CFR 50, Appendix R, Sections III.G.
and III.L. .Since the McGuire Nuclear Station utilizes the dedicated
shutdown system approach, the scope of this inspection determined if the
fire protection features provided were capable of maintaining either the
safe shutdown system or one train of normal plant hot standby systems free
from fire damage, and were capable of limiting potential fire damage to both
trains of redundant normal plant safe shutdown systems in those plant areas
where alternate or dedicated shutdown capabilities are provided.
a. Safe Shutdown Capabilities
In - order to ensure safe shutdown capabilities, where cables or
equipment of redundant trains of systems necessary to achieve and
maintain hot stand-by conditions are located within the same fire area
outside the primary containment,10 CFR 50, Appendix R, Section III.G.2
requires that one train of hot standby systems be maintained free of
fire damage by one of the following means:
Separation of cables and equipment and associated non-safety
circuits of redundant trains by a fire barrier having a 3-hour
rating;
Separation of cables and equipment and - associated non-safety
circuits of redundant trains by a horizontal distance of more than
20 feet with no . intervening combustibles or fire hazards. In
addition, fire det.ectors and an automatic fire suppression system
-shall be installed in the fire area; or,
Enclosure of cable and. equipment and ' associated non-safety
circuits of one redundant' train in a fire barrier.having a.1-hour
rating. In addition, fire detectors 'and 'an automatic fire ,
suppression 1 system shall be installed in the fire area.
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Where the protection of systems whose function is required for hot
! standby _ does not satisfy the above requirements or Section III.G.2,
! alternative or dedicated shutdown capabilities independent of cables,
systems or components in the area, room or zone under consideration
shall be provided in accordance with 10 CFR 50, Appendix R,
Section III.G.3 and III.L. In addition, Section III.G.3 requires that
fire detection and fixed suppression be installed in the area, room or
zone under consideration.
On -the basis of .the above Appendix R criteria, the inspectors made an
audit of cabling and components associated with the dedicated standby
shutdown system (SSS) to determine the adequacy of the separation
afforded to the SSS with respect to plant areas containing both
redundant trains of normal essential hot standby systems (i.e,
auxiliary feedwater system, component cooling water system, nuclear
service water system, chemical volume control system and reactor
coolant system). In addition, the inspectors made an audit of the .
standby shutdown system's ability to achieve and maintain hot standby
l and determined the adequacy of the fire protection features afforded
i for those plant areas which contain both redundant trains or normal
! essential; plant systems required to achieve and maintain safe shutdown
l conditions.
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( (1) Separation of Standby Shutdown System (SSS) from Normal Plant
Shutdown Systems.
A walk down' inspection was.made of the following SSS-cables routes
within the Unit 1 auxiliary building to verify that the SSS cables
were separated from the redundant or compliment device of the
normal essential plant shutdown system in _accordance with the
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requirements of 10 CFR 50, Appendix R, Section III.G.2.
Function Device Cable (s) Device Cable (S)
Pressurizer INCP5151 1CF726 INCLT5170 INC665
Level Inst. INC992
l 1NC990
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Steam Generator 1CFP6090 ICF726 1CFP5530 1CF588
B Level Inst. 1CF773
1CF771
Reactor Coolant INCP5121 1CF726 INCPT5170 INC713
Pressure ~ Inst. INC995
l 1N6993
Changing F1ow 1NVP6420- ICF726 Control- INV657
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Inst.- INV826 -Gauge
- INV827
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Reactor Coolant INC1, Valve INC984 Control Room INC971
System Vent NC272AC Controls (Note *)
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1NC2, Valve Valves NC272AC
NC273AC & NC273AC
Reactor Coolant INC3, Valve INC977 Control Room INC707
System Isolation NC27 Controls (Note *)
Valve NC27
1NC4, Valve INC977 Valve NC29 INC706
NC29
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1N04, Valve IWZ540 Valve ND2AC 1EPE590
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ND2AC
NV Valve IWZ540 a ve 1EPE590
- NOTE
- * Normal shutdown and SSS cables terminate in cabinet SSSFARC.
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Normal shutdown cables extend from this cabinet to normal
system isolation valves. Cables are routed by separate routes
to assure that the cables will not ground or fault in such a
manner to cause spurious valve operations.
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In general, the SSS cables are. routed with Train A cables and are
separated from Train B cables by three hour fire rated barriers.
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A review was conducted of drawing Nos. MC-1919-01.01, MC1920-
01.01, and MC1921-02.01 and " computer. cable routing program data
sheets" to determine the above cable routes within the auxiliary
and reactor buildings. Portions of the SSS cabling and the normal
shutdown cabling within the annulus of the reactor building are in
close proximity to each other. This situation is not in violation
to Appendix R, Section III.G.2,. since the annulus is considered
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part of the containment and is provided with fire detection and
automatic sprinkler protection.
For:a fire within the remainder of the containment, the SSS is-not
required to bring the plant to hot standby condition. 'The n_ormal
shutdown systems would be used and these systems have been
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evaluated by the licensee and NRR and found to be adequate to
bring the plant to 'a . safe shutdown condition following a
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containment fire.
A review of. components and cable route drawing for the SSS within
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Unit 2 indicated that the installation 'within Unit 2 was similar
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to Unit 1 and appears to be provided with sufficient separation
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from the normal shutdown" components' to meet the requirements of
10 CFR 50 Appendix R, Section III.G.
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(2) Fire Protection of Safe Shutdown Capabilities
(a) An inspection was made to determine if the fire protection
features provided for various auxiliary building areas meet
the fire protection requirements of 10 CFR 50 Appendix R,
Section III.G.3.
1) Fire Area 4 - Auxiliary Building Common Area 649
On elevation 716'-0" of the Auxiliary Building a partial
fixed automatic sprinkler system is provided to Common
Area 649 to protect the Nuclear Service Water (RN)
pumps. The area of sprinkler application in Common
Area 649 is provided between column line EE-GG-54 and
column line EE-GG 58. Power cables 1*RN571 and 2*RN559
for RN pumps 1A and 2A are partially routed outside the
area protected by sprinklers in Common Area 649 from
column line GG 56 to HH 54, where they enter the
unprotected electrical duct shaft adjacent to Charging
Pump Room 627 and are routed up through elevation
733'-0" to elevation 750'-0." Power cables 1*RN572 and
2*RN560 for RN pumps 18 and 2B are routed in Common
Area 649 from the pumps to the unprotected electrical
duct shaft near column FF57 within the sprinklered area,
where they enter the shaft and are routed up to
elevation 733'-0."
Even though train "B" RN pump power cables are located
within the sprinklered area and the train "A" RN pump
power cables are partially routed through this area, the
water discharge pattern for the sprinkler heads
installed near the ceiling level over both redundant
trains of RN pumps and their associated cabling appears
to be obstructed by cable trays and piping. In
addition, the sprinkler protection does not fully
protect the train "A" RN pump cabling in Common Area
649. Therefore, based on the sprinkler obstructions and
the lack of adequate fixed suppression coverage for both
redundant trains of normal shutdown systems located in
Auxiliary Building Common Area 649, it cannot be assured
that fire damage to both trains of the nuclear service
water system can be minimized if an exposure fire were
to occur within this plant area.
2) Fire Area 14 - Auxiliary Building Common Area 723
Component Cooling Water (KC) pump suction isolation
valves 1*KC-1A, 1*KC-2B, 1*KC-3A and 1*KC-18B are
located in Common Area 723 on Auxiliary Building
elevation 733'-0". Common Area 723 is partially
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protected by an automatic sprinkler system. The area of
sprinkler application in Common Area 723 is provided
between column line GG-JJ 55 and column line GG-JJ 57.
Valves v*KC-1A,1*KC-2B and 1*KC-188 are located within
the Compohent Cooling Water (KC) pump area which is
protected by the actomatic sprinkler system. The
sprinkler water discharge patterns for those sprinkler
heads which provide partial protection for valves
1*KC-1A,1*KC-2B and 1*KC-188 and their control cables
1*KC501, 1*KC527, 1*KC515, 1*KC541, and 1*KC516 appear
to be obstructed by cable trays and piping. Component
Cooling Water (KC) suction valve 1*KC-3A and its control
cabling 1*KC502 and 1*KC527 and control cables 1*KC501
are located outside the sprinklered area near column
line HH54 and JJ54. Therefore, based on the sprinkler
obstructions and the lack of adequate fixed suppression
coverage for both redundant trains of normal shutdown
systems located in Common Area 723, it cannot be assured
that fire damage to both trains of the Component Cooling
Water System can be minimized if an exposure fire were
to occur within this plant area.
Unit 1 Centrifugal Charging (NV) pump 1A and 1B power
cables 1*NV501 and 1*NV502 are located in Common Area
723 on auxiliary building elevation 733'-0". Common
Area 723 is partially protected by an automatic
sprinkler system. The area of sprinkler application in
Common Area 723 is provided between column line GG-JJ55
and column line GG-JJ57. Cable 1*NV502 is routed up to
elevation 733'-0" from elevation 716'-0" through the
unprotected cable shaft located outside the sprinklered
area near column line HH-54. Once on elevation 733'-0",
cable 1*NV502 stays in the cable shaft and is routed up
to auxiliary building elevation 750'-0" where it
terminates. Cable 1*NV501 is routed up to elevation
733'-0" from elevation 716'-0" through an electrical
floor penetration near column line JJ-55. Cable
1*NV501, once on elevation 733'-0", is routed across
Common Area 723 outside the sprinklered area to column
line JJ-56. Once cable 1*NV501 has reache'd this plant
location, it makes a 90* turn south and is routed
through the sprinklered area. The water discharge
pattern for . the sprinklers installed near the ceiling
level over both redundant trains of Component Cooling
Water pumps and cable 1*NV501 appears to be obstructed
by piping and cable trays.
Therefore, based on the sprinkler obstructions and the
lack 'of adequate fixed suppression coverage for bo.th
redundant trains of normal shutdown systems located in
Common Area 723, it cannot be assured that fire damage
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to both trains of the Unit 1 Chemical Volume Control
System can be minimized if an exposure fire were to
occur within this plant area.
Unit 2 Centrifugal Charging (NV) pump 2A and 2B power
cables 2*NV538 and 2*NV537 are located in auxiliary
building Common Area 723 on elevation 733'-0". Common
Area 723 is partially protected by an automatic
sprinkler system. The area of sprinkler application in
Common Area 723 is provided between column line GG-JJ55
and column line GG-JJ57. Cable 2*NV537 is routed up to
elevation 733'-0" from elevation 716'0" through the
l unprotected cable shaft located outside the sprinklered
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area near column line JJ-58. Once on elevation 733'-0",
cable 2*NV537 stays in the cable shaft and is routed up
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to auxiliary building elevation 750'-0" where it
terminates. Cable 2*NV538 is routed up to elevation
733'-0", from elevation 716'-0" through an electrical
floor penetration near column line HH57. Cable 2*NV538
once on elevation 733'-0", is routed through the
sprinklered portion of common area 723. However, the
water discharge pattern for the sprinkler heads
installed near the ceiling level over cable 2*NV538
appears to be obstructed by piping and cable trays.
- Therefore, based on the sprinkler obstructions and the
l lack of adequate fixed suppression coverage for both
t redundant trains of normal shutdown systems located in
Common Area 723, it cannot be assured that fire damage
to both redund nt trains of the Unit 2 Chemical Volume
Control System can be minimized if an exposure fire were
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to occur within this plant area.
Cabling for both trains of redundant Charging Flow
Isolation Valves is located in Auxiliary Building Common
Area 723 on elevation 733'-0". Common Area 723 is
partially protected by an automatic sprinkler system.
The area of sprinkler application within this area is
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provided between column line GG-JJ57 and GG-JJ55. Cable
1*NV555, which is associated with Charging Flow
Isolation Valve 1*NV244A, is routed up to elevation
733'-0" from elevation 716'-0' through the unprotected
cable shaft located outside the sprinklered area near
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column line HH56. Cables 1*NV575 and 1*NV576 are routed
through the portion of Common Area 723 which is
protected by sprinklers. However, the water discharge
pattern of the sprinkler heads installed near the
ceiling over cables 1*NV575 and 1*NV576 appears to be
obstructed by cable trays and piping. Therefore, based
on the sprinkler obstructions and the lack of fixed
suppression coverage for both redundant trains of normal
shutdown systems located in Common Area 723, it cannot
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! be assured that fire damage to both redundant trains of
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Unit 1 Chemical Volume Control System can be minimized
if an exposure fire were to occur within this plant
area.
3) Fire Area 4 - Mechanical Penetration Room 603
Normal Charging Flow Isolation Valves 1*NV244A and
1*NV245B to the Reactor Coolant System are located in
Mechanical Penetration Room 603 on elevation 716'-0" of
the auxiliary building. This room is not provided with
a fixed suppression system. The cabling associated with
these valves is also located within this room. Cables
1*NV55B and 1*NV579 are routed out of Mechanical
! Penetration Room 603 in two separate directions. These
cables and cable 1*NV576 associated with valve 1*NV245B
and cable 1*NV555 associated with valve 1*NV244A
reconverge in auxiliary building Common Area 649.
Common Area 649 is partially protected by an automatic
sprinkler system. The area of sprinkler application
within this area is provided between column lines
EE-GG 54 and EE-GG58. Cable 1*NV579 and 1*NV576 for
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Charging Flow Isolation Valve 1*NV245B terminate within
cabinet 1ATC3. Termination cabinet 1ATC3 and cable
1*NV576 are located within the sprinklered area.
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However, cable 1*NV579 is only partially located in the
sprinklered area. Cables 1*NV558 and 1*NV555 for
Charging Flow Isolation Valve 1*NV244A terminate within
cabinet 1ATC6. Termination cabinet 1ATC6 and cables
1*NV558 and 1*NV555 are located outside the sprinklered
area. The water discharge pattern for the sprinkler
heads installed near the ceiling over termination
cabinet 1ATC3 and in the area of the cable routes for
cables 1*NV579 and 1*NV576 appears to be obstructed by
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cable trays and piping. Therefore, based on the
sprinkler obstructions and the lack of fixed suppression
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coverage for both redundant trains of normal shutdown
systems located in Common Area 649 and mechanical
penetration room 603, it cannot assured that fire damage
to both redundant trains of the Unit 1 Chemical Volume
Control System can be minimized if an exposure fire were
to occur within either of these plant areas.
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4) Fire Area 14 - Corridor 731
Volume Control Tank Isolation Valves 1*NV141A and
1*NV142B are located in corridor 731 on auxiliary
building elevation 733'-0". Cable 1*NV560 associated
with valve 1*NV141A and cable 1*NV582 associated with
valve 1*NV1428 are routed from the valves down to the
end of Corridor 731, then they take a 90* turn and run
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down the corridor adjacent to the Boric Acid Tank Rooms
and into Common Area 753 where they separate. If an
exposure fire were to affect these valves or their
associated cables, causing these valves to spuriously
operate and go closed, the normal suction flow path to
the charging pumps could be precluded. There are no
fixed suppression capabilities provided in either
Corridor 731 or the corridor adjacent to the Boric Acid
Addition Tank Rooms to protect the Volume Control Tank
Isolation Valves and their associated cabling.
Therefore, based on the lack of fixed suppression in
these plant areas, it cannot be assured that fire damage
to both redundant trains of Unit 1 Chemical Volume
Control System can be minimized.
5) Fire Area 21 - Auxiliary Building Common Area 806
Nuclear _ Service Water Supply Isolation Valve 1*RN86A to
Component Cooling Water Heat Exchanger 1A and Nuclear
Service Water Supply Isolation Valve 1*RN1878 to l
Component Cooling Water Heat Exchanger 18 and their
associated cabling are located in Common Area 806 (fire
area 21) on auxiliary building elevation 750'-0." If an
exposure fire . were to affect these valves or their
associated cables, causing these valves to spuriously
operate and go closed, the normal nuclear service water
supply to the component cooling water heat exchangers
could be precluded. There are no fixed suppression
capabilities provided in Common Area 806 near the area
the' heat exchangers to protect the nuclear service water
supply isolation valves and associated cabling.
Therefore, based on the lack of fixed suppression in the
area of the nuclear service water supply isolation
valves to the component cooling water heat exchangers
and their associated cabling, it cannot be assured that
fire damage to both redundant trains of Unit 1 Component
Cooling Water System can be minimized.
The plant areas identified in items 5.a. 2 a 1)
,. through 5.a.(2)(a)(5), are areas, rooms, or z(on)e(s)u(nder
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' consideration and.if affected by a fire condition would
require the utilization of the dedicated standby
shutdown system to achieve and maintain hot standby
conditions. 10 CFR 50, Appendix R, Section III.G.3,
requires a -fixed suppression to be installed in the
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room, zone or area under consideration. The plant
-areas, zones, or rooms previously identified, are either
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not provided with fixed suppressinn capabilities or the
l fixed suppression system provided does not provide
adequate coverage to protect both redundant trains of
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normal safe shutdown systems. Therefore, if an exposure
fire were to occur within any of these identified areas,
it could not be assured that the fire damage sustained
by both redundant trains of normal shutdown systems
would be -minimal. The failure to meet the fire
protection requirements of 10 CFR 50 Appendix R,
Section III.G.3, as required by the operating license,
is identified as Violation Item (369/84-28-01 and
370/84-25-01) Inadequate or the Failure to Provide Fixed
Suppression Systems in Accordance with 10 CFR 50
Appendix R, Section III.G.3, for Rooms, Areas or Zones
Under Consideration.
(b) An inspection was made to determine if the fixed manual water
. spray systems in the cable spreading rooms and the automatic
sprinkler system provided to protect the cable tray stacks at <
the east and west ends of the battery room were designed and
installed in accordance with Unit I license Condition 2.C.(4)
and Unit 2 License Condition 2.C.(7). These license
conditions require the licensee to maintain in effect and I
fully implement all the provisions of the approved fire
protection plan and the NRC staff's Safety Evaluation Report,
Supplement No. 2.
Safety Evaluation Report, Supplement No. 2, indicates that
the water suppression systems are designed in accordance with
NFPA 13 and 15. In addition, this safety evaluation report
required the cable spreading rooms to be protected by a
manual fixed water spray system with a level of open spray
heads at the ceiling and an additional level of heads below
the lowest cable. trays throughout the- room. However, the
present systems do not appear to be designed in accordance
with NFPA 13 and 15 as the spray nozzles are not distributed
uniformly throughout the cable spreading rooms at the ceiling
level and at the level below the lowest cable trays.
Therefore, it cannot be assured, that if an exposure fire
were to occur in the cable spreading room, that the present
manual fixed water spray system would assist in controlling a
potential exposure fire and minimize fire damage to redundant
trains of cabling.
The Safety Evaluation Report also indicated that the
sprinklers installed in the b;tttery room provide protection
for the cable tray stacks at the east and west ends of the
room from an' exposure fire. The present battery room
sprinkler system design does not comply with the guidance
provided by NFPA-13. The present placement of the ~ sprinkler
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heads under the cable trays is approximately 5 ft. above the.
' floor. Without sprinklers placed at or near the ceiling, it
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cannot be assured that the present sprinkler design would
react in a timely manner to protect the cable tray stacks
against the potential effects of an exposure fire.
These fixed suppression systems inadequacies have been
identified as Unresolved Item (369/84-28-02 and 370/84-
75-02), Inadequate Fixed Fire Suppression System Provided for
the Cable Spreading Room and Battery Room, pending
disposition by NRR. ,
b. Protection of Associated Circuits
The inspection was conducted to verify compliance with the associated
circuit provisions of 10 CFR 50 Appendix R, Sections III.G and III.L.
The emphasis was on the following areas of concern:
Common Bus Concern
Spurious Signal Concern
Common Enclosure Concern
(1) Common Bus Concern
The common bus concern is found in circuits, either non-safety 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.
,
A number of circuits were identified which did not have adequate
circuit breaker or fuse coordination. The licensee indicated that
an ongoing breaker coordination program is in effect. The
inspectors identified the following circuits which did not meet ,
the requirements of Appendix R, Section III.G.2:
1251/. D.C. control power for charging pumps CCPA
or CCPB from panels EVDA or EVDD respectively.
600 VAC power supply for auxiliary feedwater supply
MOVs CA468, CA508, CA54AC and CA58A.
600 VAC Power supply for PORV block valves MOVINC318
and MOVINC358 -
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600 VAC nower supply for RHR isolation valve MOVIN01B I
600 VAC power supply for turbine driven auxiliary
feedwater pump suction valve CA7A
600 VAC power for nuclear service valve RN16B
600 VAC-power for VCT outlet valves NV141A and NV1428
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600.VAC power for component cooling pump suction
valves form RWST NV221A and NV222B
The licensee should perform an analysis to ensure that power to
- hot standby equipment is protected from faults in commonly powered
equipment. Presently, these conditions do net meet the require-
ments of Appendix R, Section III.G.2 and are identified as
Violation Item (369/84-28-03 370/84-25-03), Failure to Provide
Adequate Breaker / Fuse Protection for Equipment Required for Hot
Standby.
(2) Spurious Signal Concern
A review of the licensee's spurious signal analysis was conducted
to determine if the following conditions had been considered:
.
The false motor, control and instrument readings
such as what occurred at the 1975 Browns Ferry
Fire. These could be caused by fire initiated
,
grounds, shorts or open circuits.
Spurious operation of safety-related or non-safetyrelated
components that would adversely affect shutdown capability
(e.g., RHR/RCS Isolation Valves).
The licensee intends to remove power and control voltages from
valves that could affect safe shutdown of the unit should they
operate due to a fire induced spurious signal. These are:
Pressurizer Power Op'erated Relief Valves
J
Power will be removed from the pressurizer pswer
operated relief valves by removable disconnect
cables in the 125 VDC control circuit for valves
INC32B and INC368 and by an interlock relay for
INC36A when the shutdown facility is to be
activated.
Reactor Head Vent Valves
^
The head vent valves NC2'478 and NC275B will be
inhibited from operation by the removal of cable
disconnects in the 125 VDC control circuitry.
The head vent valves NC272,C and NC273A,C will have
control capability trarsferred to the standby shut-
down facility which is electrically isolated from
the control room.
The RCS/RHR boundary valves were the only high I'ow pressure
interface which were identified and analyzed for spurious opera- '
tion. There are , installed, a number of interlocks in series in
the control circuit for the RHR suction valve and none of them are-
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l- located in the same fire area. The valve opening circuit contains
in series a control switch contact, a pressure interlock and valve
limit switch interlock contacts. The limit switch ISW27 is locked
opened and consequently inhibits the ability to apply control
- power to the control circuits for the INDIE and 1ND2A RCS isola-
tion valvec.
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These additional circuit analyses were reviewed by the inspectors:
l RCS Boundary Valves and Centrifugal Pump Charging
The licensee conducted an analysis to determine the
availability of a charging path to maintain reactor
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inventory. It was determined, at the time of the inspection,
,
that the necessary modifications, such as one-hour fire
'
retardant blanket wraps on circuits for the RHR/RCS valves
ND2AC, were installed in the turbine driven auxiliary feed-
water pump room.
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Main Steam Isolation Valves (MSIV) and the Main Steam Power
Operated Relief Valve (PORV)
I
The control circuits for main steam isolation valves ISM 1AB
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ISM 3AB, ISM 3AB, ISM 7AE were analysed. Two solenoid valves
i
are installed in series in the pneumatic control line, the
closure of any one will cause the MSIV to close. The control
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- ircuits for main steam power operated relief valves 15VIAB,
i 1SV7ABC, ISV13AB, ISV19AB were analyzed. Three solenoid
valves are connected-in series in the pneumatic control lines
for the main steam power operated relief valves, the closure
of any one will cause the respective main steam PORV to
close. The McGuire Nuclear Station does not depend on the
main steam PORVs for safe shutdown.
Dedicated instrumentation, electrically independent of control
room, has been provided at the standby shutdown panel to monitor
the following parameters:
Pressurizer Pressure
Pressurizer Level
Standby Makeup Pump Discharge Pressure
. Steam Generator Level
Incore Temperature (T Hot)
The licensee has committed to install RTDs to monitor T Cold i
during the next refueling.
Source range instrumentation was not installed and an exemption
request had been granted.
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The spurious signal concern was not satisfactorily addressed since
some buses feeding safe shutdown equipment did not have adequate -
coordination of circuit breakers and fuses. Some of the safe
shutdown components which were affected by the lack of
coordination were as follows:
i
Centrifugal Charging Pump CCPA
- Centrifugal Charging Pump CCPB
Turbine Driven Auxiliary Feedwater Pump Suction Valve CA7A
PORV Block Valves - MOV INC318 and MOVING 358
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This is another example of Violation Item (369/84-28-03 and
370/84-25-03), Failure to Provide Adequate Breaker / Fuse Protection
for Equipment Required for Hot Shutdown.
'
(3) The Common Enclosure Concern
The common enclosure concern is found when redundant trains are
routed together with a non-safety circuit which crosses from one 1
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raceway or enclosure to another, and the non-safety circuit is not
electrically protected or fire can destroy both redundant trains
due to inadequate fire protection means.
!
The common enclosure concern at McGuire was not a concern since
the standby shutdown system cables were not run in the same trays
with either the redundant trains or their associated non-safety-
related cables.
i
For fires where redundant Trains A and B are to be used instead of
the standby shutdown system, the licensee wrapped one redundant
train. This was done in the Unit 1 Train B switchgear room where
some Train A cabling was wrapped with 3-hour rated fire retardant
blanket.
c. Dedicated Shutdown and Fire Damage Control Capabilities
(1) System Description and Operation
SSER No. 6, dated January 25, 1983, documents NRR review of the
, licensee dedicated shutdown system and its conformance to
10 CFR 50, Appendix P. , Section III.L. The McGuire dedicated
shutdown system is identified as the Standby Shutdown System ,
L(SSS). This system provides an alternate and independent means to l
achieve and maintain the reactor coolant system in hot standby
condition for one or both units. The (SSS) is placed into
operation only 'if a' postulated fire results in the installed
normal and emergency plant systems becoming inoperable. A masonry
structure located adjacent to and outside the plant, houses the
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major equipment and controls for the (SSS). This facility is
designated the Standby Shutdown Facility (SSF) and consist of a
diesel generator, starting batteries and supporting auxiliary
systems. Normal electrical power is supplied via 6.9 kv site
transformer, dedicated emergency power via the diesel generator,
battery bank for 125 VDC, 600 VAC and 125 VAC power distribution
systems, and a control panel for monitoring and controlling
primary and secondary volumes. Reactor coolant system pressure
control is provided by manual control of a bank of pressurizer
heaters and pressurizer level via manual operation of reactor
vessel head vent valves. Makeup water to the reactor coolant
system and sealing water to the reactor coolant pumps (RCP) are
provided by a 26 gpm makeup pump connected to the RCP sealwater
injection line. Each standby makeup pump is located in the unit's
containment building annulus and takes suction from the spent fuel
pool transfer area. Steam generator volume control and decay heat
removal are accomplished by utilizing the normal auxiliary
feedwater system to maintain steam generator water level
requirements. The main steam safety valves are used to control
secondary side pressure and to dump steam to provide decay heat
removal from the reactor coolant system.
The licensee gave a presentation on the general operator actions
and the procedures to be used when the SSF/SSS is placed into
operation for those' cases where the fire renders the control room
and the auxiliary . shutdown panels inoperable. The procedures and
sequence were described as follows:
- AP 1/A/5500/01, Reactor Trip
This procedure provides the instructions to stabilize and
control the plant following the trip of the reactor which
would take place upon detection of a disabling fire.
- OP/0/A/6100/17, Operation of the Standby Shutdown Facility
This procedure describes the use of the SSF/SSS systems,
operational controls, and stations to be manned in order to
maintain the unit or units in a hot standby mode.
- OP/0/A/6100/20, Operational Guidelines Following Fire in
Auxiliary Building er Vital Area
This procedure describes the steps to be taken and plant
systems required to bring the plant to hot shutdown followed
by cold shutdown within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. It is intended by the
licensee to use normal operating and plant shutdown
procedures to accomplish these evolutions.
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- IP/0/A/3090/23, Fire Damage Control Procedure
This procedure establishes methods for the restoration of '
instrumentation, electric motors, system valves, electric
breakers, etc., following a fire so that plant equipment and i
. systems necessary to bring the plant to cold shutdown
conditions are made operable. This procedure is performed in
conjunction with the aforementioned procedures in order to <
expedite efforts to take the plant to cold shutdown.
(2) Review of Operating and Surveillance Procedure
The inspectors reviewed the completed data in PT/A/4209/01C,
Standby Makeup Pump Flow Periodic Test for Units 1 and 2 to verify
that the output capacity for each pump met the 26 gpm minimum
specified in the SSER No. 6. The test results for Unit 1
(completed 9/5/84), indicated a flowrate of 29.0 gpm and for
Unit 2 (completed 8/31/84) a flowrate of 30 gpm which satisfies
those values specified in SSER No. 6.
The inspectors reviewed those plant procedures identified for use
in the case where a fire causes the control room to be abandoned
and the auxiliary shutdown panels are also rendered inoperable.
This review was conducted to verify that information in design and
engineering documents and the information provided in SSER No. 6
have been factored i.nto the appropriate plant operating procedures
for . the SSF/SSS systems and procedure for subsequent cold
shutdown.
In addition to the procedure reviews, walkdown of selected
procedures were conducted to ensure that the instruction provided
-
was complete and usable; that steps identified components _and
equipment correctly and equipment is accessible to plant operators
for operation.
The following plant procedures were reviewed:
-
AP/1/A/5500/01 (Change 1), Reactor Trip
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~0P/1/A/6100/07 (Change 8),. Operation of the Standby Shutdown
Facility
-
OP/0/A/6100/20 (Change 0), Operational Guidelines Following
Fire in Auxiliary Building or Vital Area-
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OP/0/A/6100/13 (Change 0), Operational -Guidelines Following
Fire in Containment or Doghouse
.AP/1/A/5500/17 (Change 0), Loss of Control Room
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The review and walkdown of the procedures resulted in the
following findings and concerns, which apply equally to both
units:
(a) OP/0/A/6100/17, Operation of the Standby Shutdown Facility
1) Provide appropriate steps in the procedure to ensure
that pressurizer spray isolation valves are shut and the
reactor coolant pumps are shutdown. SSER No. 6,
Section C.3.5, describes these components as the means
of terminating pressure decreases in the event, the
spray valves become open due to electrical shorts caused
by a fire.
2) Enclosure 4.1, Section 1,' Step 2.2, requires the normal
power source for SSS system equipment to be switched to
its alternate power source, MCC-SMXG. Appropriate steps
need to be added to the procedure that ensure that
either the diesel generator is supplying power to the
alternate bus or the offsite power supply is available.
3) Enclosure 4.1, Section 1, Step 2.4, Note: Requires
obtaining a master key to gain access through fire
doors. This step needs to be moved to the front of the
procedure as preoperations and made readily available so
as not to delay the detailed procedure steps.
4) Enclosure 4.1, Section 2, Step 2.0, second Note:
Specifies maintaining the spent fuel water level per
OP/1/A/6200/05.
Appropriate precautionary measures need to be added to
the procedure to ensure that the makeup water added to
the spent fuel pool is from the refueling water storage
tank or other suitable source. In addition, the baron
concentration is required to be equal to or greater than
2000 ppm. The standby makeup pump suction is from the
spent fuel pool.
5) Enclosure 4.1, Section 4, Step 2.2. Provide
instructions for converting the incore thermocouple-
digital readout to core temperature.
6) -Enclosure 4.1, Section 2, Step 2.2 Note: This note
needs to be expanded and clarified so.that the operator
knows what to look for when monitoring the standby
makeup pump D/P filter gauge.
The above items collectively are identified as Inspector
Followup Item (369/84-28-13 and 370/84-25-13), Amplify
and Clarify Certain Steps of OP/0/A/6100/17.
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7) Enclosure 4.1, Section 3, Step 2.1, states to verify
that the auxiliary feedwater pumps start when Lo-Lo
level is reached on 2/4 steam generators. Steam
generator levels are maintained by dispatching an
operator to the auxiliary feedwater pump area and
position manual valves to control levels. The inspector
-
expressed a concern as to whether having three auxiliary
feedwater pumps feeding the steam generators, with the
unit in hot standby and natural circulation in progress
would be excessive. The inspector requested that an
analysis be performed to determine if SSF/SSS operations
would be jeopardized regarding maintaining hot standby
conditions, primary, and secondary volume control. The
parameters of specific interest that need to be
addressed are: effects on cooldown rate, amount of
reactor coolant system temperature decrease, the effects
on pressurizer level (amount of shrinkage), overfill of
steam generators, and effects on reactivity shutdown
margin. The time for an operator to be dispatched to
the auxiliary feedwater pump area, and conduct
operations to gain level control need to be considered
in the analysis. These concerns are identified as an
Unresolved Item (369/84-28-14 and 370/84-25-14),
Analysis Effects on SSF/SSS Operation, Potential
Excessive Feeding Steam Generators.
(b) AP/1/A/5500/01, Reactor Trip;
AP/1/A/5500/17, Loss of Control Room; and.
OP/0/A/6100/20, Operational Guidelines Following Fire in
Auxiliary Building or Vital Area
As a result of reviewing and conducting walkdowns of the
above procedures, and holding discussions with licensee
personnel, it. appears that the overall coordinated effort for
a smooth departur_e from the control room to the SSF/SSS to
maintain plant hot standby conditions is fragmented, not well
defined, and may lead to confusion, delays, and possible
errors.
The procedures as presently written has AP/1/A/5500/17, Loss
of Control Room, as the controlling document for establishing
systems and plant conditions for leaving-the controi. room and
then going to the auxiliary shutdown panels for - continued
plant control to eventually cold shutdown. The licensee has
stated, however, that when a fire disables the control room,
it will also make the auxiliary shutdown stations not
available for use. Based on these conditions, the inspectors
have the following findings and concerns:
, _ _ ._ _ . _ _
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1) Review and correct the Initial Conditions specified in
OP/0/A/6100/17. Some of these conditions need to be
made procedure steps in order to establish the initial
conditions since AP/1/A/5500/17 will not be used.
2) Establish the necessary procedure steps to place the
plant and equipment in a stable hot standby which will
permit an orderly departure from the Control Room and
the use of the Standby Shutdown Sy3 ten.
3) Establish che necessary procedure steps to take the
plant from hot standby to cold shutdown. These
procedures will address shutdown from the control room
and shutdown from outside the control room.
4) Conduct final walkdowns of procedures giving particular
attention to procedural adequacy, access to system
components, and verify communications are satisfactory.
These concerns and findings were discussed with the licensee
on October 1, 2, and 3, 1984. Subsequent to these
- discussions, a Confirmation of Action Letter dated October 9,
1984, was issued which identified October 5, and October 19,
1984, as dates for completing corrective actions on
procedures. The inspector identified this area as Unresolved
Item (369/84-28-15 and 370/84-25-15), Correct Procedure
Deficiencies Used to Accomplish 10 CFR 50, Appendix R,
Section III.L.
(c) Personnel Training
- The inspectors- held discussions with training department
instructors to determine that training is being given concerning
the operation and use of the SSF/SSS. It was determined that
-personnel who are receiving training included senior reactor
operator (SRO), ' reactor operator (RO), and nuclear equipment
operators (NE0). The -licensee has scheduled training for the
required personnel. The inspectors reviewed the licensee's lesson
plans, _ training schedules, examination questions, and found these
documents _to be well organized, detailed, and comprehensive.
(d) Fire Damage Control-Procedures
The inspectors reviewed McGuire~ Nuclear Station Fire Damage-
Control Procedure No. IP/0/A/3090/23. The purpose of this
procedure _is to establish a method of making the diesel generators
operable, coritrolling 4.16kv breakers, -installing power and l
control cables to. certain 4.16ky motors, installing instrumenta- '
tion and restoring- valve operability after a' fire in order to
place the plant in cold shutdown status. The procedure identified
the motors, valves'and instrumentation required to bring the plant .
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to cold shutdown that would require restoration if damaged by
fire.
The procedure identifies the types of power cables required for
the identified motors. Included in the procedures are instruc-
tions for preparation of stress cones, instructions and routing
, for the. installation of temporary power and control cables and
clearing any control cables that may be faulted due to fire
damage. .This procedure provides for the manual operation of
4.16kv breakers after disconnecting certain remote control
conductors. Instructions are included for replacing certain
electronic sensing instrumentation (level transmitters, flow
transmitters, pr' essure transmitter, etc.) with direct heading
gauges. While drawings are provided as part of the procedure,
instructions require that the latest revision of the drawings be
used to perform the work.
The inspector examined a large box containing the required number
of stress cone kits needed to restore all the 4.16ky motor power
cables that were identified in the procedure. Included in this
box were the necessary gauges required for instrumentation
restoration along with various lengths of sensing line and
fittings. The licensee had designated nine reels of cables to be
used for the power cable replacement. The box of materials was
tagged to indicate the intended use and stored in the warehouse.
The cable reels.had not yet been located in a designated area.
The inspector examined the designated routes for the replacement
power cables for the centrifugal charging pump motors, the
residual heat removal pump motors and the component cooling pump
motors. The routes were found to be practicable and all areas
were accessible for cable pulling. It appears that the cables
could be installed with off-site power not available to power
electric pulling equipment but would require additional personnel.
,
6. Compliance With 10 CFR 50 Appendix R, Section III.0
The inspector (s) reviewed the as-built documentation / drawing file of the oil
collection system for the reactor coolant pumps.
Potential oil leakage points for each pump have been provided with _ a
Westinghouse designed ar.d.fu'rnished RCP oil containment system consisting of
a upper bracket oil overflow enclosure, lower bearing oil catcher, oil lift
enclosure, and oil ' tooler enclosure. These enclosures are connected to
drain piping which discharges into a separate collectionLtank provided for
each reactor coolant pump. The reactor coolant pump oil collection system
was' originally designed and installed at this facility prior to the issuance
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of Appendix R. In -a letter dated January 9,1981, to the NRC, the licensee
acknowledged their previous commitments to provide the oil . collection
system, but.noted that the RCP oil containment system and its associated
drainage; system would require additional analysis to verify compliance with
_
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the requirements of Appendix R, Section III.O. The analysis was completed
and the oil containment system and related drain piping were seismically
upgraded and modified to function following a design base seismic event.
This upgrade was performed under the licensee's seismic quality assurance
(QA) program as verified by a review of the following records:
a. Seismic Analysis:
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Westinghouse Electric Corporation, Engineering Report Memorandum
5802, dated September 10, 1982
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Duke Power Memorandum, File MC-1435.03, MC-1223.03, MC-1415.00,
MC-1421.00, dated February 17, 1982
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Flow diagrams MC-1553-4.0 and MC-2553-4.0
b. Upgraded Seismic Hanger Inspection:
Design Isometric Date of Final
Hanger Number Drawing QC Inspection
1-MCR-NC-2342 (MCSRD-SPM118/1440 11/16/82
1-MCR-NC-2269 MCSRD-SPM118/144A 02/21/83
1-MCR-NC-2303 MCSRD-SPM118/144C 11/13/82
Since both units were operating at full power during the inspection
period, the inspector (s) were unable to review the reactor coolant pump
oil collection system's installation for conformance to the design
requirements. This review will be made during a ' subsequent NRC
inspection when the units are in a refueling outage. Within the areas
reviewed, no items of noncompliance or deviations were identified.
7. Compi*ance With 10 CFR 50, Appendix R, Section III.J
Section III.J, requires that: " Emergency lighting units with at least an
8-hour battery power supply shall be provided in all areas needed for
operation of safe shutdown equipment and in access and egress routes
thereto."
a. Emergency Lighting System Walkdown
The inspector (s). performed a-walkdown examination of the design and
installation of the 8-hour emergency lighting. units for the facility
based upon the licensee desig, drawings, and the post-fire alternate
shutdown procedures utilizing .the standby shutdcwn system,
OP/0/A/6100/17,.0peration of the Standby Shutdown Facility, and
AP/1/A/5500/17, Loss of Control Room. Also, the battery powered
' lighting units for several plant areas were observed while energized to
determine the lighting beam direction and relative illumination levels.
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4
As a result of the emergency lighting walkdown, it appears that
inadequate emergency Itghting conditions existed in the following plant
areas:
(1) Several lights in the Unit 1 doghouse were mounted behind concrete
! columns, piping, etc., which eliminate their effectiveness to
illuminate access ladders to safe shutdown related SM and SV
valves.
. (2) No 8-hour battery powered lighting units were provided for Units 1
!
and 2, Common Corridor 908 which provide a portion of the
access / egress routes between the main control room and the standby
shutdown facility.
The failure to meet the emergency lighting requirements of 30 CFR 50,
Appendix R,'Section III.J is identified as Violation Item (369/84-28-04
and 370/84-25-04), Failure to Comply with the Requirements of
10 CFR 50, Appendix R, Section III.J.
In addition, the licensee requested exemptions from the emergency
lighting requirements for several plant areas in letters dated
November 18, 1983, and February 20, 1984. These letters requested
i exemption from the requirement to provide 8-hour t:attery powered
emergency -lighting units in the standby shutdown facility and along a
portion of the yard area access route between the standby shutdown
facility and the turbine building. NRR granted the above enmption
based upon the licensee's commitment to place battery powered hand
.
'
lanterns in the control room to be used in emergency situations by the
plant operators. During the plant walkdown, an inspection was made to
determine if the battery powered hand lights were provided. Contrary
to the licensee's commitment the hand lights had not been installed in
the control room. However, after this item was identified, the
licensee took immediate corrective action and provided, for emergency
use, two battery powered hand lanterns within the control room. This
is identified as Deviation Item (369/84-28-05 and 370/84-25-05),
,
Failure.to Provide Battery Powered Hand Lanterns in the Control Room.
b. Evaluation of Emergency Lighting Testing Procedures
4
The inspector (s) reviewed the licensee's periodic opeational testing
procedures (PT/1/B/4350/09, Unit 1 Emergency . Lighting Annual' Test;
.
pT/2/B/4350/09, Unit 2 Emergency - Lighting Annual Test.), and ampleted
tests records for 1983 and 1984 on the emergency lighting system. And
as a result of this review, the following discrepancies were
identified:
(1) The = emergency lighting annual . test procedures- did not address
. verifying whether the light beam is pointed in the correct
direction to illuminate all . proper pieces of equipment, electrical
,
cabinets, enclosures, or access / egress path ways required for
emergency operation.
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(2) The procedures did not provide for immediate corrective actions or
compensatory measures to be taken for deficiencies found during
performance testing of the periodic operational test.
(3) The procedures did not provide the testing frequency and scope of
performance testing in accordance with the manufacturer's
recommendations which:
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Require the test button be depressed for at least one to two
minutes to assure verification of AC to battery DC transfer
capability and adequate electrical power drain from the
battery to verify operation of the battery charging circuit
of .the battery charging circuit, and
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Require a specific testing frequency for a particular
manufacturer's lighting unit. The emergency lighting testing
frequency at McGuire is greater than that recommended by the
manufacturer. From the review of the test records, it
appears that the test period for all emergency battery
powered lighting units ranged from two months to almost seven
months duration for both units. Specific dates that
individual battery powered lighting units were tested are not
recorded making testing frequency of the individual lighting
units indeterminate.
(4) The procedures did not include or reference capacity of the 8-hour
self-contained battery packs nor scheduled battery pack
replacement.
Based upon the deficiencies listed above, the 8-hour emergency lighting
system design, performance testing, and surveillance and maintenance
programs appear to be inadequate for ensuring that such a system is
always operational to enable operators to transfer control to, and
operate the SSF and SSS functions. These findings are identified as
Inspector Followup Item (369/84-28-06 and 370/84-25-06), Inadequate
Surveillance Testing Procedures for Emergency Lighting, and will be
reviewed during a subsequent NRC inspection.
8. Fire Protection and Prevention Program
a. Fire Protection for Safety-Related Pumps
The Unit I license condition 2.C.(4) and Unit 2 license condition
2.C(7) requires the licensee to fully implement and maintain in effect
all provisions of the approved fire protection plan. The McGuire
Nuclear Station Fire Protection Review, September 1982 Revisien,
Section F.11 - " Safety-Related Pumps", indicates that redundant
safety-related pumps are separated by required fire barriers and
automatic detection with alarm and annunciation in the control room.
An inspection was made to determine if this license condition was fully
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plan.
The following safety-related pumps have been identified as not being
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Recycle Evaporator Feed Pumps, Room 620
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Waste Drain Tank Pumps, Room 639
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Boron Injection Recir. Pumps - Unit 2, Room 788
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Boron Injection Recir. Pumps - Unit 1, Room 730
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Fuel Pool Cooling Pumps - Unit 1, Room 816
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Fuel Pool Cooling Pumps - Unit 2, Room 829
The following safety-related pumps have been identified as not being
provided with automatic fire detection capabilities:
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Recycle Evaporator Feed Pumps, Room 620
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Waste Drain Tank Pumps, Room 639
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Fuel Pool Cooling Pumps - Unit 1, Room 816
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_ Fuel Pool Cooling Pumps - Unit 2, Room 829
The failure to fully implement the provisions of the approved fire
protection plan as required by the operating license is identified as
Violation Item (369/84-28-07 and 370/84-25-07), Failure to Provide
Automatic Fire Detection for- and Fire Barriers to Separate
Safety-Related Pumps,
b. Hydrogen Piping Systems
A review was made of the hydrogen gas piping systems to the volume
control tanks and reactor coolant drain -tanks. The hydrogen for the
volume control tanks is supplied from the bulk gas storage located in
the plant yard. The hydrogen to the reactor coolant drain tanks is
supplied from two H2 cylinders also lo'cated in the bulk gas storage
structure outside in the plant yard. The following documentation was
- reviewed to verify that the systems were seismically supported. .
Flow Diagram No.
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- MC-1603-1.0
- MC-1565-1.1
-MC-2565-1.1, and .
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- Duke Power Memorandum W. H. Taylor, Jr. to P. R. Herran, dated
September 27, 1984, McGuire Nuclear Station, W. L. System
Upgrade of Piping to Class F, File MC-1206.02.88 j
The hydrogen gas piping to the volume control tanks inside the i
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auxiliary building .is designed and installed as Duke Class F (ANSI l
B31.1-Seismic loadin'g). Hydrogen: piping to ; the reactor- drain tanks j
, extends from the yard through -the - turbine building, service building, l
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auxiliary building and reactor building through the drain tank. The
piping at the entrance to the auxiliary building to the reactor
building containment isolation boundary is Duke Class F, Class C and
Class B respectively, designed for seismic loading.
As noted in' the above Duke memorandum, the piping (approximately 43
inches in length) from containment isolation valves 1WL39A (Unit 1) and
2WL39A (Unit 2) to the reactor coolant drain tanks has been modified to
seismically qualified Duke Class F. The supports are designed with
seismic loads per calculation MCC-1206.12-05-2014 (Unit 1). and
MCC1206.16-27-3104 (Unit 2). Based upon the above review, the hydrogen
piping within safety-related plant areas appears to meet the general
design requirements.
c. Fire Pumps
During the inspection, a question arose regarding the availability of
the three fire pumps for fire suppression activities in certain
instances. Fire pumps A and B are powered from 6900 VAC non-safety
load centers 2TB and ITD respectively, and fire pump C is powered from
a 44ky substation separate from the plant switchyard.
Appendix R requires the plant to recover from the effects of a fire and
be in cold shutdown in 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> with the loss of offsite power (LOP).
An LOP at McGuire could render all three fire pumps inoperative, in
that ~ none of the fire pumps are fed from the emergency diesel
generators, which would greatly hamper fire suppression efforts.
Assuming that LOP is caused by grid instability, pump C could also be
lost even though it is fed from a dedicated off-site 44kv line which is
separate from the station switchyard.
A review of the Technical Specifications for McGuire Nuclear Station
revealed another, area of concern. Section 3/4.7.10, Fire Suppression
System, requires that at least.two fire suppression pumps be operable
at all times. There is no requirement as to which pumps must be
operable. Therefore, pump C could be inoperative for -an unlimited
amount of time as long as pumps A and B are inservice /available.
Pumps A and B would be lost on the LOP and with pump.C out of service
the plant would have very limited fire suppression capability. Since '
Appendix R requires that the LOP to be considered concurrent with the.
fire event, it is conceivable that no fire pumps would be available to
support fire fighting activities.
This item is being reviewed by the NRC' staff and is identified as
Unresolved Item (369/84-28-08 and 370/84-25-08), Adequacy of Power
Supplies for Fire Pumps A, B and C.
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d. Emergency Communications
Post Fire Alternate Shutdown Procedure OP/0/A/6100/17, Operation of the
Standby Shutdown Facilities, requires that communications must be
established between the various areas of the plant where local control
actions must be taken. This procedure identifies that portable radios
can be utilized _as one method of establishing communications. By
letter dated December 14, 1982, the licensee responded to NRR's
questions concerning the standby shutdown system. The licensee's
response to questions 0 and P states that portable radios will be
available for communications between the standby shutdown facility and
the auxiliary feedwater local control stations.
An inspection was made which evaluated the adequacy of the radio
communications by actually establishing radio communications between
the various local control stations (i .e. , Units 1 and 2 Train "A"
switchgear rooms, Unit 1 doghouse, and Units 1 and 2 Steam Driven
Auxiliary Feedwater Pump Rooms) and the Standby Shutdown Facility.
However, direct radio communications were not feasible between the
standby shutdown facility and the auxiliary feedwater local control
stations and other essential control stations required for the
operation of the standby shutdown system. This item is identified as
Deviation Item (369/84-28-09 and 370/84-25-09), Failure to Provide
Adequate Radio Communications Between Local Control Stations and the
Standby Shutdown Facility.
9. Licensee Identified Items
a. June 1, 1984 Nonconformance Report, Potential Fire Damage to Unit 2
Turbine Driven Auxiliary Feedwater Pump Suction Valves: On May 18,
1984, the licensee discovered that the location and protection of
auxiliary feedwater pump suction valves 2CA-161C and 2CA-1626 were not
in accordance with previous commitments made to the NRC and were in
nonconformance to 10 CFR 50, Appendix R, Section III A. These valves
are required to open automatically to align a long term source of water
supply to the suction of the turbine driven auxiliary feedsater pump
(TDAFP) in case of fire in the adjacent motor driven auxiliary
feedwater pump (MDAFP) room. A fire within the MDAFP room could have
damaged the two MDAFPs and caused damage to the TDAFP suction valve
operators and/or associated cables thus eliminating the capacity for
automatic alignment to the long term water source. However, a
postulated fire in the MDAFP roora of sufficient intensity to- damage
j both MDAFPs and the automatic _ switchover capacity of valves 2CA-161C
' and 2CA-162C was'unlikely due to the light combustible fire loading and
the availability of the early warning automatic fire detection system '
/ and the automatic wet pipe sprinkler system within the room.
Furthermore, the normal suction source to the TDAFP would be available
to provide a water source for a minimum of 41s hours. .This would
probably have allowed adequate time for the fire brigade to extinguish
the fire and' operators to manually realign the valves if necessary.
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This nonconformance was identified by the licensee and promptly
reported to' the NRC on May 18, 1984. Also, as noted above the
deficiency did not present a significant threat to the health and
safety of the public. Therefore, since this discrepancy meets the
- guidelines of. 10 CFR 2, Appendix C, Section IV.A, for licensee
identified problems, no violation is being issued. ,
The licensee's ~ corrective action consisted of the installation of a
one-hour fire. barrier enclosure for valves 2CA-161C and 2CA-162C and
associated cables. This arrangement brought this area up to meet the
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provisions of Appendix R, Section III.G.2, except the structural steel
supports for TDAFP piping and cables to operators for valves 2CA-161C
and 2CA-162C are not protected to provide a fire resistance equivalent
to that of the one-hour fire barriers. This does not meet the
, requirements of'10 CFR 50, Appendix R, Section III.G.2, as required by
License -Section 2, Item C.7.a, and is identified as an example of
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Violation Item (369/84-28-10 and 370/84-25-10), Structural Steel Fire
Barrier Supports Not Provided with Fire Resistant Rating Equivalent to
the Fire Barrier.
b. August 2, 1984 Nonconformance Report, Potential Fire Damage to
Redundant Safe Shutdown Equipment and Cabling in Various Fire Areas of
Units 1 and 2. On July 18, 1984, the licensee discovered a number of
' areas in Units 1 and 2 which were in nonconformance to commitments made
to the NRC and to the requirements of 10 CFR 50, Appendix R,
Section III.G. These discrepancies were promptly reported to the NRC
on July 18, 1984, and with a-followup written report sent to Region II-
on August 2, 1984. The following items were identified:
(1) Two suction valves which are arranged to open automatically to
align a long term source of water supply to the suction of the
Unit 1 TDAFP and associated cabling to the valve operators are
located within a Unit 1_ pipe chase and mechanical penetration
room. Redundant components and the cabling for the normal plant
shutdown systems of centrifugal charging pumps IA and 18 are also
,
located within this same fire area. A postulated fire in this
area could have incapacitated portions of all safe shutdown trains
and plant' shutdown would have been. difficult.to obtain. The room
was provided with automatic fire detectors but a fire suppression
system was not provided, the area is not readily accessible for
manual fire fighting operations, and is in a potentially high
radiation area.
The failure to meet the separation requirements of 10 CFR 50,
Appendix R, Section III.G, for this area as required by the
operating -license is identified as Violation Item (369/84-28-11),
Inadequate Appendix R, Section III.G, Fire Protection Features and
Separation Provided for Redundant Trains of Normal Shutdown
Systems and the Standby Shutdown System.
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The licensee maintained an hourly fire watch patrol for this room
until the TDAFP cabling and valve operators were enclosed in a
three-hour fire rated barrier. This barrier was reviewed by the
inspectors and appeared satisfactory, except the structural steel
supports for valves, piping and cabling to the valve operators
were not provided with a' three-hour fire resistant rating as
required by Appendix R, Section III.G.
(2) Unit 1 Train A associated control circuits and the standby
shutdown system (SSS) cables are both located in Train 8 switch-
gear room. None of these cables were enclosed in a three-hour
fire barrier. A fire detection system is provided for this area
but a fire suppression system is not provided. A postulated fire
within the Unit 1 Train B switchgear room could have damaged
control cables for Train A centrifugal changing and auxiliary
feedwater pumps, control and power cables for the standby makeup
pump, and control and power cables for Train B centrifugal
changing and auxiliary feedwater pumps. This could have prevented
safe plant shutdown.
The failure to meet the separation requirements of 10 CFR 50,
Appendix R, Section III.G, for this area as required by the
operating license is identified as Violation Item (369/84-28-11),
Inadequate Appendix R, Section III.G, Fire Protection Features and
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Separation Provided for Redundant Trains of Normal Shutdown
Systems and the Standby Shutdown System.
The licensee maintained an hourly fire watch for this area until
the normal Train A shutdown component cabling within Train B
switchgear room was enclosed within - a three-hour fire barrier.
This barrier was reviewed by the inspectors and appeared
satisfactory.
(3) Control and cabling for the MDFPs components and the other Train A
and B safe shutdown components including the cer trifugal changing
pumps are. routed through the Units 1 and 2 TDAFP rooms. The TDAFP
rooms are provided with fire detection systems and automatic halon
fire suppression systems. However, the cabling in this area was
not~ enclosed within a one-hour fire barrier as required by '
Appendix R,.Section III.G. A postualted fire within one . of the
TDAFP rooms of sufficient ' intensity to damage the shutdown
components 'to prevent plant shutdown was not probable due to the
low fire loading within the rooms, early . warning fire detection
system, and automatic halon fire suppression system.
This discrepancy was promptly reported to the NRC after being
~ identified by the licensee on July 18,1984, : and with a written
report sent to Region II on August 2, 1984. An hourly fire watch
was initiated and is to be maintained until the. controls and
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cabling associated with the MDAFP suction valves are removed from
the Units 1 and 2 TDAFP rooms and/or the Train B -shutdown cabling
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within - the TDAFP rooms are enclosed within a one-hour fire
barrier. Based on the above, this discrepancy did not present a
significant threat to the health and safety of the public, was
identified by the licensee, and appropriate corrective action was
initiated. Therefore, this discrepancy meets the guidelines of
10 CFR 2, Appendix C, Section IV.A, for licensee identified
problems and no violation is being issued.
c. August 3, 1984 Appendix R Deviation Notice:
During the licensee's ongoing fire protection program review, several
deviations from Appendix R . were identified. These deviations and
technical justifications were forwarded by the licensee to NRC/NRR on
August'3, 1984. These deviations as listed below will remain
outstanding pending NRR evaluation:
(1) Steel Penetrating Fire Barriers
(a) The lh hour fire barriers between redundant nuclear service
water pumps and component cooling water pumps are penetrated
by cable tray hangers.
(b) The 3-hour fire walls separating the TDAFP and MDAFP room are
penetrated by steel pipe supports and restraints.
(2) Reactor Building Wall Penetrations
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(a) Process piping penetrations in reactor building are designed
for pressure boundary integrity .in lieu of fire boundary
penetrations.
(b) Spare sleeves and instrument tubing penetrations are sealed
by steel plate or pipe cap on the auxiliary building side of
4
(c) HVAC duct penetrations do not have fire dampers.
(d) Access into the reactor building from the auxiliary building
is provided by two portals which have not been fire tested.
(3) Fire Boundary Doors With . Security Hardware
Fire doors have been modified to meet security requirements and
-some fire walls have security and other special type doors in lieu
of standard fire dcors.
(4) . Cork Expansion Joints
Cork has .been provided in the structural joints between : cme
structures for seismic considerations. However, this configura-
tion has not been tested for three-hour fire resistance rating.
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.The above items. do not technically meet Appendix R requirements and
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other NRC guidelines. -Therefore, pending NRR evaluation, these items
are identified as Unresolved Item (369/84-28-12 and 370/84-25-12), NRR
Evaluation of Appendix R Deviation Request of August 3,1984, and will
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be-reviewed during a subsequent NRC inspection.
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