ML030650560
| ML030650560 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 09/09/2002 |
| From: | Ernstes M Operator Licensing and Human Performance Branch |
| To: | Keenan J Carolina Power & Light Co |
| References | |
| 50-324/03-301, 50-325/03-301 | |
| Download: ML030650560 (95) | |
See also: IR 05000325/2003301
Text
Final Submittal
(Blue Paper)
Final RO/SRO Written Examination References
BRUNSWICK EXAM
50-2003-301
50-325 & 50-324
FEBRUARY 10 - 13 & 19, 2003
Attachment 5
Page 15 of 29
ATTACHMENT 5 (Cont'd)
FIGURE 1
DRYWELL SPRAY INITIATION LIMIT
I
I
I .P
X~P~N~t>~tk:<I< I
h~LNEA1 MNA,,,
UNSAFE
1-1 I..
7.
r~g
NI
Xzimssc
_ jr-v
IL
0
w
0.
w
I
-j -LJ
L
4
.
I
IF
SAFE
35
- 0
'4
40
~5
55
65
75
50
60
70
DRYWELL PRESSURE (PSIG)
NOTE
DRYWELL AVERAGE AIR TEMPERATURE MAY BE DETERMINED
USING ATTACHMENT 4 OF THE "USER'S GUIDE"
Rev. 40
Page 68 of 139
450
25
20
3
-4
751/2
4tk xtS
400
350 --
300
250
-nfl
LUU, -
150
100
'I
-
4LVffi
I
44>
1/4
r
5
0
15
10
- TZ"..
M kl-$ ý,,
- ll.,,"P,:I,ý-.'ýý,,,ý.)"r"ý',ýýý,ýý.ýlý,ý,-Li,-eýl"".ýý"i'?
pl-
P F-El
i
I
Attachment 5
Page 16 of 29
ATTACHMENT 5 (Cont'd)
FIGURE 2
PRIMARY CONTAINMENT PRESSURE LIMIT-A
100 -
90-
U S F
280
70
D
60
U)
U)
50
50
40
SSAFE
20
10
_
0
0
10
20
30
40
50
60
70
80
PRIMARY CONTAINMENT WATER LEVEL (FEET)
If
using the following instrument:
PCPL-A is:
CAC-PI-1230
70 psig
CAC-Pi-4176
Use the graph
CAC-PR-1257-1
Use the graph
OEOP-01-UG
I
Rev. 40
Page 69 of 139
Attachment 5
Page 17 of 29
ATTACHMENT 5 (Cont'd)
FIGURE 3
UNIT 1 HEAT CAPACITY TEMPERATURE LIMIT
t
LA
0
M
0~
0
w
(
a.
0
0O
300
500
UNSAFE ABOVE
SELECTED LINE
700
220
210 -
200 -
190 -:
180 1
1701
160
150 -
140 -
130
120
110 -
100 --
900 1 1,100
(-)
(-) 0.25 FT
1.25 FT
(-) 2.50 FT
(-) 3.25 FT
(-) 4.25 FT
(-) 5.50 FT
-1,150
0
200
400
600
800
1,000
REACTOR PRESSURE (PSIG)
NOTE
SUPPRESSION POOL WATER TEMPERATURE IS DETERMINED BY:
CAC-TR-4426-1A POINT WTR AVG
OR CAC-TR-4426-2A POINT WTR AVG
OR COMPUTER POINT G050
OR COMPUTER POINT G051
OR CAC-TY-4426-1
OR CAC-TY-4426-2
SELECT GRAPH LINE IMMEDIATELY BELOW SUPPRESSION POOL WATER LEVEL AS THE LIMIT.
I OEOP-01 -UG
I
Rev. 40
Page 70 of 139
-A
SAFE BELOW
SELECTED LINE
N
100
Attachment 5
Page 18 of 29
ATTACHMENT
5 (Cont'd)
FIGURE 3A
UNIT 2 HEAT CAPACITY TEMPERATURE LIMIT
0
0~
IL'
I
0
0
(
z
0
u;
C,
0.25 FT
1.25 FT
2.50 FT
3.25 FT
220
210
200
190
180
170
160
150
140
130
120
110
100
I 30(0 i 500 i 700 1 900 1 1,100
200
400
600
800
1,000
REACTOR PRESSURE (PSIG)
NOTE
SUPPRESSION POOL WATER TEMPERATURE IS DETERMINED BY:
CAC-TR-4426-IA POINT WTR AVG
OR CAC-TR-4426-
2 A POINT WTR AVG
OR COMPUTER POINT G050
OR COMPUTER POINT 0051
OR CAC-TY-442
6 -1
OR CAC-TY-4
4 2 6 -2
SELECT GRAPH LINE IMMEDIATELY BELOW SUPPRESSION POOL WATER LEVEL
AS THE LIMIT.
OEOP-01-UG
Rev. 40
Page 71 of 139
(-)
(-)
(-)
(- )
(-) 4.25 FT
(-) 5.50 FT
-1,150
Attachment 5
Page 19 of 29
ATTACHMENT 5 (Cont'd)
FIGURE 4
UNIT 1 MAXIMUM CORE UNCOVERY TIME LIMIT
20
0
5
10
15
MAXIMUM CORE UNCOVERY TIME - MINUTES
OEOP-01-UG
I
Rev. 40
1
Page 72 of 139
U,
w
z
0
0
w
ce
w
Attachment 5
Page 20 of 29
ATTACHMENT 5 (Cont'd)
FIGURE 4A
UNIT 2 MAXIMUM CORE UNCOVERY TIME LIMIT
0
5
10
15
MAXIMUM CORE UNCOVERY TIME - MINUTES
20
oFoP-ol-UG
Rev. 40
Page 73 of 139
I
U,
z
z
0
D
U)
0
I-C)
4
Attachment 5
Page 21 of 29
ATTACHMENT 5 (Cont'd)
FIGURE 5
CORE SPRAY NPSH LIMIT
F Ld
-~~
~
~
---
-
-
-
-
ii.
0 w
Ix
0~
0J
a
0
0
0.
[L
0n
290
280
270
",)an
2,000
3,000
4,000
5,000
6,000
7,000
CORE SPRAY FLOW (GPM)
NOTE
SUBTRACT 0.5 PSIG FROM INDICATED SUPPRESSION CHAMBER PRESSURE FOR EACH
FOOT OF WATER LEVEL BELOW A SUPPRESSION POOL WATER LEVEL OF -31
INCHES
(-2.6 FEET).
- SUPPRESSION CHAMBER PRESSURE (CAC-PI-1257-2A OR CAC-PI-1257-2B)
OEOP-01 -UG
Rev. 40
Page 74 of 139
-
7
250
240
230
220
210
200
190
180
170
160
0
1,000
- 40 PSIG
- 20 PSIG
- 10 PSIG
- 5 PSIG
- 0 PSIG
!
Attachment 5
Page 22 of 29
ATTACHMENT
5 (Cont'd)
FIGURE 6
0
'a w
M~
'a
0
z
0
a.
M
U)
290
280
270
260
250
240
230
220
210
200
190
180
170
160
0
5,000
10,000
15,000
20,000
RHR PUMP FLOW (GPM)
NOTE
SUBTRACT 0.5 PSIG FROM INDICATED SUPPRESSION CHAMBER PRESSURE FOR EACH
FOOT OF WATER LEVEL BELOW A SUPPRESSION POOL WATER LEVEL OF -31
INCHES
(-2.6 FEET).
- SUPPRESSION CHAMBER PRESSURE (CAC-PI-1257-2A OR CAC-PI-1257-2B)
OEOP-01-UG
I
Rev. 40
Page 75 of 139
Attachment 5
Page 23 of 29
ATTACHMENT 5 (Cont'd)
FIGURE 7
UNIT 1 PRESSURE SUPPRESSION PRESSURE
0
10
20
30
40
SUPPRESSION CHAMBER PRESSURE (PSIG)
OEOP-01-UG
Rev. 40
Page 76 of 139
IL
w
w
0
0
0z)
C0
w
IL
C
U,
+2
+I
0
-I
-2
-3
-4.
-5
-6
-7
-8
Attachment 5
Page 24 of 29
ATTACHMENT 5 (Cont'd)
FIGURE 7A
UNIT 2 PRESSURE SUPPRESSION PRESSURE
U
w
0
z
0
Li,
w
CL
(Li
+2
+1
0
-1
-2
-3
-4
-5
-6
-7
-8
0
10
20
30
40
SUPPRESSION CHAMBER PRESSURE (PSIG)
OEOP-01-UG
Rev. 40
Page 77 of 139
Attachment 5
Page 25 of 29
ATTACHMENT 5 (Cont'd)
FIGURE 8
SRV TAILPIPE LEVEL LIMIT
+6
+5
+4
+3
+2
+1
0
-1
-2
-3
-4
1 100
1 300 1 500 1 700
I 900 11,100
0
200
400
600
800
1,000
REACTOR PRESSURE (PSIG)
IOEOP-01-UG
I
Rev. 40
Page 78 of 139
U
-J
0.
0
0~
zn
0)
w
U)
1,150
Attachment 5
Page 26 of 29
ATTACHMENT
5 (Cont'd)
FIGURE 9
UNIT 1 CORE SPRAY VORTEX LIMIT
+5
+4
+3
+2
+1
0
-1
-2
-3
-4
-5
-6
-7
-8
P
U
-j
-J
0 0
[L
z
0
0)I
w
IL
IL
Do
H Iu-
__
UNSAFE
N
I
0
1,000
2,000
3,000
4,000
5,000
4
III
I -.
6,000
I b,
7,000
CORE SPRAY FLOW (GPM)
I
Rev. 40
Page 79 of 139
a
I
-9
-10
, lzNg
94+/-
--I
+
-
1+
Attachment 5
Page 27 of 29
ATTACHMENT 5 (Cont'd)
FIGURE 10
UNIT 2 CORE SPRAY VORTEX LIMIT
I
I16,0
7,
0
0
1,000
2,000
3,000
4,000
5,000
6,000
7,000
CORE SPRAY FLOW (GPM)
OEOP-01-UG
Rev. 40
Page 80 of 139
w
0
0
0.
IL
C')
+5
+4
+3
+2
+1I
0
-1
-2
-3
-4
-5
-6
-7
-8
-9
-10
Attachment 5
Page 28 of 29
ATTACHMENT 5 (Cont'd)
FIGURE 11
UNIT 1 RHR VORTEX LIMIT
+L
+
+
+
0
5,000
10,000
15,000
20,000
OEOP-01-UG
I
Rev. 40
Page 81 of 139
3
_
4
37~
-SAFE
2
I
_
0
I
2
5
6
7
---
UNSAFE
8
90
lO
u
w
w
0
0
az
0
co
w
a
CD
I
Mj
M
KIIAMR-T
j m
ON
Attachment 5
Page 29 of 29
ATTACHMENT 5 (Cont'd)
FIGURE 12
UNIT 2 RHR VORTEX LIMIT
r
5,000
10,000
15,000
20,000
OEOP-01-UG
Rev. 40
Page 82 of 139
+4
+3
+2
+1
0
-1
-2
-3
-4
-5
-6
-7
-8
-10
SAFE
w
-J
0
z
0
w
a
U)
.
7ý
0
Attachment 10
Page 1 of 4
Attachment 10
Secondary Containment Temperature
And Radiation Limits
Page 129 of 139 1
oEOP-01-UG
Re.4
Rev. 40
Suppression Chamber-to-Drywell Vacuum Breakers
.3.6.1.6
3.6
CONTAINMENT SYSTEMS
3.6.1.6
Suppression Chamber-to-Drywell Vacuum Breakers
APPLICABILITY:
Eight suppression chamber-to-drywell vacuum breakers shall
be OPERABLE for opening.
AND
Ten suppression chamber-to-drywell vacuum breakers shall be
closed, except when performing their intended function.
MODES 1, 2, and 3.
ACTIONS
CONDITION
REQUIRED ACTION
COMPLETION TIME
A. One required
A.1
Restore one vacuum
72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />
bre e to UVrnADlEC
suppression chamber
to-drywell vacuum
breaker inoperable for
opening.
B. One suppression
chamber-to-drywell
vacuum breaker not
closed.
C.
Required Action and
associated Completion
Time not met.
B.1
C.I
breaker
t
o uuts.OLL
status.
Close the open vacuum
breaker.
Be in MODE 3.
4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />
12 hours
AND
C.2
Be in MODE 4.
36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />
Amendment No.
248 I
Brunswick Unit 2
I
3.6-18
Suppression Chamber-to-Drywell Vacuum Breakers
- 3.6.1.6
SURVEILLANCE REQUIREMENTS
SURVEILLANCE
SR 3.6.1.6.2
- - - - - - - - - - NOTE ------------------
Not required to be met for vacuum
breakers that are open during
Surveillances.
Verify
each
vacuum
breaker
is
closed.
Perform a functional test of each
required vacuum breaker.
14 days
AND
Within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />
after any
discharge of
steam to the
suppression
chamber from
any source
31 days
AND
Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />
after any
discharge of
steam to the
suppression
chamber from
any source
Verify the full open setpoint of each
24 months
required vacuum breaker is 2 0.5 psid.
Amendment No. 248 I
Brunswick Unit 2
I
3.6-19
FREQUENCY
Suppression Chamber-to-Drywell Vacuum Breakers
B 3.6.1.6
B 3.6
CONTAINMENT SYSTEMS
B 3.6.1.6
Suppression Chamber-to-Drywell Vacuum Breakers
BASES
BACKGROUND
Brunswick Unit 2
The function of the suppression-chamber-to-drywell
vacuum
breakers is to relieve vacuum in the drywell.
There are
10 internal vacuum breakers located on the vent header of
the vent system between the drywell and the suppression
chamber, which allow flow from the suppression chamber
atmosphere to the drywell when the drywell is at a negative
pressure with respect to the suppression chamber.
Therefore, suppression chamber-to-drywell vacuum breakers
prevent an excessive negative differential pressure across
the suppression chamber-drywell boundary.
Each vacuum
breaker is a self actuating valve, similar to a check valve,
which can be remotely operated for testing purposes.
A negative differential pressure across the drywell wall is
caused by depressurization of the drywell.
Events that
cause this depressurization are cooling cycles, inadvertent
drywell spray actuation, and steam condensation from sprays
or subcooled water reflood of a break in the event of a
primary system rupture.
Cooling cycles result in minor
pressure transients in the drywell that occur slowly and are
normally controlled by heating and ventilation equipment.
Spray actuation or spill of subcooled water out of a break
results in more significant pressure transients and becomes
important in sizing the jnternal vacuum breakers.
In the event of a primary system rupture, steam condensation
within the drywell results in the most severe pressure
Following a primary system rupture, the drywell
atmosphere is purged into the suppression chamber free
airspace, leaving the drywell full of steam.
Subsequent
condensation of the steam can be caused in two possible
ways, namely, Emergency Core Cooling Systems flow from a
recirculation line break, or drywell spray actuation
following a loss of coolant accident (LOCA).
These two
cases determine the maximum depressurization rate of the
drywell.
In addition, the waterleg in the Mark I Vent System
downcomer is controlled by the drywell-to-suppression
chamber differential pressure.
If the drywell pressure is
less than the suppression chamber pressure, there will be an
(continued)
0
3.U A'
Revision No.
18
D J.}.-*
Suppression Chamber-to-Drywell Vacuum Breakers
B 3.6.1.6
BASES
BACKGROUND
(continued)
APPLICABLE
SAFETY ANALYSES
increase in the height of the downcomer waterleg.
This will
result in an increase'in the water clearing inertia in the
event of a postulated LOCA, resulting in an increase in the
peak drywell pressure.
This in turn will result in an
increase in the pool swell dynamic loads.
The internal
vacuum breakers limit the height of the waterleg in the vent
system during normal operation.
Analytical methods and assumptions involving the
suppression chamber-to-drywell vacuum breakers are presented
in Reference 1 as part of the accident response of the
primary containment systems.
Internal (suppression
chamber-to-drywell) and external (reactor building
to-suppression chamber) vacuum breakers are provided as part
of the primary containment to limit the negative
differential pressure across the drywell and suppression
chamber walls that form part of the primary containment
boundary.
The safety analyses assume that the internal vacuum breakers
are closed initially and are fully open at a differential
pressure of 0.5 psid (Ref.
1). Additionally, 3 of the
10 internal vacuum breakers are assumed to fail in a closed
position (Ref.
1).
The results of the analyses show that
the design pressure is not exceeded even under the worst
case accident scenario.
The vacuum breaker opening
differential pressure setpoint and the requirement that 8
of 10 vacuum breakers be OPERABLE (the additional vacuum
breaker is required to meet the single failure criterion)
are a result of the requirement placed on the vacuum
breakers to limit the vent system waterleg height.
The
total cross sectional area of the main vent system between
the drywell and suppression chamber needed to fulfill this
requirement has been established as a minimum of 51.5 times
the total break area.
In turn, the vacuum relief capacity
between the drywell and suppression chamber should be 1/16
of the total main vent cross sectional area, with the valves
set to operate at * 0.5 psid differential pressure.
Design
Basis Accident (DBA) analyses assume the vacuum breakers to
be closed initially and to remain closed and leak tight,
until the suppression pool is at a positive pressure
relative to the drywell.
The suppression chamber-to-drywell vacuum breakers satisfy
Criterion 3 of 10 CFR 50.36(c)(2)(ii) (Ref. 2).
(continued)
Revision No. 18 I
Brunswick Unit 2
B 3.6-44
Suppression Chamber-to-Drywell Vacuum Breakers
B 3.6.1.6
BASES
(continued)
LCO
Only 8 of the 10 vacuum breakers must be OPERABLE for
opening.
All suppression chamber-to-drywell vacuum
breakers, however, are required to be closed (except when
the vacuum breakers are performing their intended design
function).
The vacuum breaker OPERABILITY requirement
provides assurance that the drywell-to-suppression chamber
negative differential pressure remains below the design
value.
The requirement that the vacuum breakers be closed
ensures that there is no excessive bypass leakage should a
LOCA occur.
APPLICABILITY
In MODES 1, 2, and 3, a DBA could result in excessive
negative differential pressure across the drywell wall,
caused by the rapid depressurization of the drywell.
The
event that results in the limiting rapid depressurization of
the drywell is the primary system rupture that purges the
drywell atmosphere and fills the drywell free airspace with
steam.
Subsequent condensation of the steam would result in
depressurization of the drywell.
The limiting pressure and
temperature of the primary system prior to a DBA occur in
MODES 1, 2, and 3.
In MODES 4 and 5, the probability and consequences of these
events are reduced by the pressure and temperature
limitations in these MODES; therefore, maintaining
suppression chamber-to-drywell vacuum breakers OPERABLE is
not required in MODE 4 or 5.
ACTIONS
A.
With one of the required vacuum breakers inoperable for
opening (e.g., the vacuum breaker is not open and may be
stuck closed or not within its opening setpoint limit, so
that it would not function as designed during an event that
depressurized the drywell), the remaining seven OPERABLE
vacuum breakers are capable of providing the vacuum relief
function.
However, overall system reliability is reduced
because a single failure in one of the remaining vacuum
breakers could result in an excessive suppression
chamber-to-drywell differential pressure during a DBA.
Therefore, with one of the eight required vacuum breakers
inoperable, 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is allowed to restore at least one of
the inoperable vacuum breakers to OPERABLE status so that
iontinued8
CAr
Revision No. 18 1
Brunswick Unit 2
[0
>.U. -'t
Suppression Chamber-to-Drywell Vacuum Breakers
B 3.6.1.6
BASES
ACTIONS
A.1
(continued)
plant conditions are consistent with those assumed for the
design basis analysis.
The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Completion Time is
considered acceptable due to the low probability of an event
in which the remaining vacuum breaker capability would not
be adequate.
B.1
With one vacuum breaker not closed, communication between
the drywell and suppression chamber airspace could occur,
and, as a result, there is the potential for primary
containment overpressurization due to this bypass leakage if
a LOCA were to occur.
Therefore, the open vacuum breaker
must be closed.
A short time is allowed to close the vacuum
breaker due to the low probability of an event that would
pressurize primary containment.
If vacuum breaker position
indication is not available, an alternate method of
verifying that the vacuum breakers are closed is to verify
that the differential pressure between the suppression
chamber and drywell is maintained > 0.5 times the initial
differential pressure for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> without nitrogen makeup.
The 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Completion Time is considered adequate to perform I
this test.
C.1 and C.2
If any Required Action and associated Completion Time can
not be met, the plant must be brought to a MODE in which the
LCO does not apply.
To achieve this status, the plant must
be brought to at least MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and to MODE 4
within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
The allowed Completion Times are
reasonable, based on operating experience, to reach the
required plant conditions from full power conditions in an
orderly manner and without challenging plant systems.
SURVEILLANCE
REQUIREMENTS
Each vacuum breaker is verified closed (except when the
vacuum breaker is performing its intended design function)
to ensure that this potential large bypass leakage path is
not present.
This Surveillance is performed by observing
the vacuum breaker position indication or by verifying that
(continued)
Revision No. 23 I
Brunswick Unit 2
B 3.6-46
Suppression Chamber-to-Drywell Vacuum Breakers
B 3.6.1.6
BASES
SURVEILLANCE
REQUIREMENTS
(continued)
the differential pressure between the suppression chamber
and drywell is maintained > 0.5 times the initial
differential pressure for I hour without nitrogen makeup.
The 14 day Frequency is based on engineering judgment, is
considered adequate in view of other indications of vacuum
breaker status available to operations personnel and
procedural controls to ensure the drywell is normally
maintained at a higher pressure than the suppression
chamber, and has been shown to be acceptable through
operating experience.
This verification is also required
within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after any discharge of steam to the
suppression chamber from any source.
A Note is added to this SR which allows suppression chamber
to-drywell vacuum breakers opened in conjunction with the
performance of a Surveillance to not be considered as
failing this SR.
These periods of opening vacuum breakers
are controlled by plant procedures and do not represent
inoperable vacuum breakers.
Each required vacuum breaker must be cycled to ensure that
it opens adequately to perform its design function and
returns to the fully closed position.
This is accomplished
by verifying each required vacuum breaker operates through
at least one complete cycle of full travel.
This SR ensures
that the safety analysis assumptions are valid.
The 31 day
Frequency of this SR was developed, based on Inservice
Testing Program requirements to perform valve testing at
least once every 92 days.
A 31 day Frequency was chosen to
provide additional assurance that the vacuum breakers are
OPERABLE, since they are located in a harsh environment (the
suppression chamber airspace).
In addition, this functional
test is required within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after a discharge of steam
to the suppression chamber from any source.
Verification of the vacuum breaker opening setpoint is
necessary to ensure that the safety analysis assumption
regarding vacuum breaker full open differential pressure of
0.5 psid is valid.
The 24 month Frequency is based on the
(continued)
Revision No. 23 I
Brunswick Unit 2
I
B 3.6-47
Suppression Chamber-to-Drywell Vacuum Breakers
B 3.6.1.6
BASES
SURVEILLANCE
REQUIREMENTS
REFERENCES
(continued)
need to perform this Surveillance under the conditions that
apply during a plant outage and the potential for an
unplanned transient if the Surveillance were performed with
the reactor at power.
The 24 month Frequency has been
demonstrated to be acceptable, based on operating
experience, and is further justified because of other
surveillances performed more frequently that convey the
proper functioning status of each vacuum breaker.
1. UFSAR, Section 6.2.
Revision No. 18 I
Brunswick Unit 2
B 3.6-48
FIGURE 1
Page 1 of 1
Estimated Capability Curves
ATB 4 POLE, 963,000 KVA, 1800 RPM, 24,000 VOLTS
0.90 RE 0.58 SCR, 60 PSIG HYDROGEN PRESSURE, 500 VOLTS EXCITATION
800
600
400
(Ir
n2
0,
0
0
w
-J
200
0
200
400
600
CURVE AB LIMITED BY FIELD HEATING
CURVE BC LIMITED BY ARMATURE HEATING
CURVE CD LIMITED BY ARMATURE CORE END HEATING
Page 25 of 28
Rev. 41
Unit 2
APP UA-13 1-4
Page 1 of 2
GENERATOR AUTO TRIP TO MANUAL
AUTO ACTIONS
1.
If the alarm was caused by an exciter field overcurrent,
the
generator backup lockup is
energized (refer to APP UA-13 1-3,
GEN-XFMR BACKUP L/O UNIT TRIP).
2.
If shift to manual was caused by a loss of control power, the
regulator will shift back to AUTO and reflash the field when
control power is restored.
3.
If
shift to manual was caused by overexcitation and the excitation
has not returned to less than or equal to 100% in
5 seconds, the
generator backup lockup is energized (refer to APP UA-13 1-3,
GEN-XFMR BACKUP L/O UNIT TRIP).
4.
If shift to manual is
due to volts/hertz being excessive, the
following actions will occur:
a.
If generator is tied to grid, no actions result.
b.
If generator is
not tied to grid, the following actions will
occur:
(1)
Use of voltage regulator will be blocked.
(2)
Regulator will run back to no load.
(3)
If
excessive volts/hertz signal is
not cleared in
60 seconds, the exciter field breaker will trip.
CAUSES
1.
Exciter field overcurrent.
2.
Generator field overexcitation.
3.
Excessive volts/hertz in exciter.
4.
Loss of DC control power.
5.
Circuit malfunction.
OBSERVATIONS
1.
GEN-XFMR BACKUP L/O UNIT TRIP (UA-13 1-3) alarm.
2.
GENERATOR FIELD OVEREXCITATION
(UA-13 2-4) alarm.
3.
GENERATOR EXC FIELD OVERCURRENT
(UA-13 3-4) alarm.
4.
Regulator shifts to manual.
Rev. 26
Page 9 of 96
Unit 2
APP UA-13 1-4
Page 2 of 2
ACTIONS
1.
Notify Load Dispatcher of the problem.
2.
If
the voltage regulator mode swaps to manual, place the voltage
regulator selector switch in
MANUAL and perform the following:
a.
If
cause of alarm was momentary, try to determine cause of
alarm and verify system parameters have returned to normal.
b.
When cause for alarm is no longer a concern, return the
voltage regulator selector switch to auto.
3.
If the generator backup lockout is
energized, refer to APP UA-13
1-3,
GEN-XFMR BACKUP L/O UNIT TRIP.
4.
If Circuit Breaker 2 (control power) in
125V DC Distribution Panel
10A is tripped or off, reset and close the breaker.
5.
If Circuit Breaker 2 in
125V DC Distribution Panel 10A trips
again, ensure that a WR/WO is prepared.
DEVICE/SETPOINTS
Voltage regulator control switch
AND
Generator Exciter Field Overcurrent
Relay 76/50
Overexcitation Relay J1K
Volts/Hertz Relay 43T
AUTO
400 amps instantaneous
overcurrent or 180 amps @
60 seconds
105%
Energized
POSSIBLE PLANT EFFECTS
1.
Loss of unit generator.
2.
If
generator trips, possible reactor Scram.
REFERENCES
1.
2.
3.
9527-LL-9351 -
34
APP UA-13 1-3,
GEN XFMR BACKUP L/O UNIT TRIP
GEK-33798 Vol.
II,
Generator Section
Page 10 of 96]
IR
Rev. 26
I
Unit 2
APP-UA-23 6-6
Page 1 of 1
VOLT BALANCE RELAY A OPERATION
AUTO ACTIONS
1.
Transfers excitation to manual.
2.
Prevents generator loss of field relay (40-1) from actuating.
3.
Prevents generator voltage restrained time overcurrent relay
(51V-1)
from actuating.
4.
Prevents generator directional distant relay (21G-1)
from
actuating.
CAUSE
I.
2.
Decreased voltage balance (80% reduction).
Circuit malfunction.
OBSERVATIONS
NONE
ACTIONS
1.
2.
As necessary, adjust generator excitation to maintain voltage.
If
a circuit or equipment malfunction is
suspected, ensure that a
WR/JO is prepared.
DEVICE/SETPOINTS
Generator Voltage Balance
Relay 60-1 Right
80% balanced reduction
POSSIBLE
PLANT EFFECTS
1.
Loss of automatic voltage control.
2.
Loss of some generator protective relaying.
REFERENCES
9527-LL-9361
-
15
I
Rev. 47
[ 2APP-UA-
Page 89 of 92 1
I
-23
Unit 2
APP UA-13 3-1
Page 1 of 1
GEN LOSS OF EXC
AUTO ACTIONS
I.
Energizes the generator primary lockout relays (refer to APP UA-13
1-1, GEN-XFMR PRIMARY L/O UNIT TRIP).
2.
Energizes the generator breaker failure lockout relays if
the
generator failed to trip
on the generator primary lockout relays,
and if
an instantaneous phase or ground overcurrent condition
exists on the breaker.
CAUSES
1.
Loss of generator excitation.
2.
Circuit malfunction.
OBSERVATIONS
1.
GEN-XFMR PRIMARY L/O UNIT TRIP (UA-13 1-1) alarms.
ACTIONS
1.
Refer to APP UA-13 1-1, GEN-XFMR PRIMARY L/O UNIT TRIP.
DEVICE/SETPOINTS
Loss of Field Relay 40
20% restraint
POSSIBLE PLANT EFFECTS
1.
Loss of unit generator.
REFERENCES
1.
9527-LL-9351
-
28
2.
APP UA-13 1-1, GEN-XFMR PRIMARY L/O UNIT TRIP
I
Rev. 26
Page 37 of 96
ATTACHMENT 5
Page 1 of 1
Reactor Pressure vs Saturation Temperature
300
600
900
REACTOR PRESSURE (PSIG)
I OAOP-36.2
Rev. 24
1
Page 177 of 180
0
JI
w
z
0
F
z
-0
0
600
550
500
450
400
350
300
250
200
0
1200
I I I
~ll
l
ll
l l
l l ll
ll
l l ~ l l ll0
I I I
i
l l l l
l l i l
i t i ,
l i l l
it00
II
ATTACHMENT 6
Page 1 of 1
Reactor Cooldown Plot
IL
0
0~
w
0
1
2
3
TIME IN HOURS
OAOP-36.2
Rev. 24
Page 178 of 1801
600
500
400
300
200
100
Unit 2
APP A-05 3-5
Page 1 of 2
REACTOR VESS HI PRESS
AUTOMATIC ACTIONS
NONE
CAUSE
1.
MSIV closure.
2.
MSIV failure (disk/stem separation)
3.
EHC System malfunction.
4.
Pressure setpoint set too high.
5.
Circuit malfunction.
OBSERVATIONS
I.
MSIVs indicating closed.
2.
One of the steam line flow indicators indicating no flow with
associated MSIVs indicating open indicates a disk/stem separation.
.3.
Turbine control valves, stop valves, or bypass valves closing
indicates an EHC System malfunction.
4.
Pressure setpoint set greater than 945 psig.
5.
REACTOR VESSEL 141 PRESS alarm on with reactor pressure less than
1050 psig indicates a defective trip
unit.
ACTIONS
1.
If
a reactor Scram occurs, refer to EOP-01-RSP.
2.
For a disk/stem separation:
a.
Close the MSIVs associated with blocked steam line.
b.
Notify the Reactor Engineer that new core analysis is
needed.
3.
If pressure setpoint is
set too high, reduce reactor pressure to
1030 psig.
4.
If
a circuit malfunction is
suspected, ensure that a WR/JO is
prepared.
DEVICE/SETPOINTS
Pressure Trip Unit B21-PTS-N023A-2
1050 psig
Pressure Trip Unit B21-PTS-N023B-2
1050 psig
Pressure Trip Unit B21-PTS-N023C-2
1050 psig
Pressure Trip Unit B21-PTS-N023D-2
1050 psig
Rev. 44
Page 44 of 93
Unit 2
APP A-05 3-5
Page 2 of 2
POSSIBLE
PLANT EFFECTS
I.
Reactor Scram if
pressure increases to 1060 psig.
REFERENCES
1.
LL-9364 -79
2.
EOP-01-RSP, Reactor Scram Procedure
Rev. 44Pae4of9
Attachment
10
Page 2 of 4
ATTACHMENT 10
SECONDARY CONTAINMENT TEMPERATURE AND RADIATION LIMITS
FIGURE 22
SECONDARY CONTAINMENT AREA TEMPERATURE
TABLE 1
AREA TEMPERATURE LIMITS
PLANT
PLANT
STEAM LEAK
INSTRUMENT
MAX
NORM
MAX SAFE
AUTO
AREA
LOCATION
DETECTION
NUMBER/
OPERATING
OPERATING
GROUP
DESCRIPTION
CHANNEL/LOCATION
WINDOW
VALUE
(0F)
VALUE
ISOL
(NOTE 1)
(4F)
N CORE
N CORE
PANEL XU-3
VA-TI-1603
120
175
N/A
SPRAY
SPRAY ROOM
S CORE
S CORE
PANEL XU-3
VA-TI-1604
120
175
N/A
SPRAY
SPRAY ROOM
RWCU PUMP
B21-XY-5949A
G31-TE-NO1EA
ROOM A
B21-XY-5949B
G31-TE-NO16B
CH.
Al-I
RWCU PUMP
B21-XY-5949A
G31-TE-N016C
ROOM B
B21-XY-5949B
G31-TE-N016D
140
225
CH.
A2-1
B21-XY-5949A
G31-TE-NO16E
ROOM
B21-XY-5949B
G31-TE-N016F
CH.
A3-1
N RHR
B21-XY-5948A
ElI-TE-N009A
N RPC
EQUIP ROOM
CH.
A5-4
190
295
N
PANEL xUU3
VA-TI-1601
S RHR
B21-XY-5948B
Ell-TE-N009B
EQUIP ROOM
CH.
A5-4
PANEL XU-3
VA-TI-1602
S RMR
RCIC EQUIP
B21-XY-5949A
E51-TE-N023A
ROOM
E21-XY-5949B
E51-TE-N023B
165
295
5
CH.
A1-3
5PCI EQUIP
B21-XY-5948A
E41-TE-N030A
ROOM
B21-XY-5948B
E41-TE-N030B
165
CH. A2-I
A21-XY-5949A
E51-TE-NO25A
TUNNEL
B21-XY-5949B
E51-TE-N025B
190
295
5
CH.
A3-3
STEAM
TUNNEL
21-XY-5948A
E51-TE-N025C
HPCI ETM
B21-XY-5948B
E51-TE-N025D
19
54
TUNNEL
CH.
A5-1
20 FT NORTH
B21-XY-5948A
B21-TE-5761A
CH. AI-4
20 FT
20 FT SOUTH
B21-XY-5948B
B21-TE-5763B
140
200
N/A
CH.
AI-4
50 FT MW
B21-XY-5948A
B21-TE-5762A
020NA
50 FT
CH.
A2-41420N/
50 FT SE
B21-XY-5948B
B21-TE-5764B
CH.
A2-4
R-EACTOR
MULTIPLE
WINDOW
ALARM
N/A
3,4,
AND/OR
BLDG
AREAS
PANEL A-02
5-7
SETPOINT
5
RýEACTOR
WINDOW
ALARM
N/A1
BLDGn
PANEL A-06
6-7
SETPOINT
NOTE 1
mAX NORM OPERATING VALUE IS THE ANNUNCIATOR /GROUP
ISOLATION SETPOINT WHERE APPLICABLE
Attachment 10
Page 3 of 4
ATTACHMENT
10 (Cont'd)
FIGURE 23
SECONDARY CONTAINMENT AREA DIFFERENTIAL TEMPERATURE
TABLE 2
AREA DIFFERENTIAL TEMPERATURE LIMITS
PLANT AREA
PLANT
STEAM LEAK
MAX NORM
LOCATION
DETECTION
OPERATING
DESCRIPTION
CHANNEL
VALUE
(4F)
(MOTE
1)
RWCU PUMP
ROOM A
RWCU PUMP
ROOM B
ROOM
N RHR
EQUIP ROOM
S RHR
EQUIP ROOM
EQUIP ROOM
EQUIP ROOM
TUNNEL
]
TUNNEL
REACTOR
MULTIPLE
BLDG
AREAS
B21-XY-5949A
B21-XY-5949B
CH.
A4-1
B21-XY-5949A
B21-XY-5949B
CH.
A5-1
B21-XY-5949A
B21-XY-5949B
CH.
A6-1
B21-XY-5948A
CH.
A6-4
B21-XY-5948B
CH.
A6-4
B21-XY-5949A
B21-XY-5949B
CH.
A2-3
B21-XY-5948A
B21-XY-5948B
CH.
A3-1
B21-XY-5949A
B21-XY-5949B
CH.
A4-3
B21-XY-5948A
B21-XY-5948B
CH.
A6-1
47
50
50
47
47
47
47
ALARM
A-02 6-7
SETPOINT
AU1U
AUTO'
GROUP
ISOL
3
N/A
N/A
N/A
5
4
3,
4,
AND/OR 5
NOTE 1:
MAX NORM OPERATING VALUE IS THE ANNUNCIATOR/GROUP
ISOLATION SETPOINT WýHERE APPLICABLE
OEOP-o1-UG
Rev. 40
Page131 of 39
N RHR
S RHR
STEAM
TUNNEL
Attachment
10
Page 4 of 4
ATTACHMENT 10 (Cont'd)
FIGURE 24
SECONDARY CONTAINMENT AREA RADIATION
TABLE 3
AREA RADIATION LIMITS
PLANT
PLANT LOCATION
MAX NORM
MAX SAFE
AREA
DESCRIPTION
CHANNEL
OPERATING
OPERATING
VALUE (mR/HR)
VALUE
(mR/HR)
N CORE
15
200
7000
SPRAY
ROOM
S CORE
16
200
7000
SPRAY
ROOM
N RHR
N RHR
17
200
7000
ROOM
S RHR
S RHR
18
200
3000
ROOM
HPCI ROOM
N/A
N/A
- 3000
N ACROSS
19
FROM TIP ROOM
RX
DRYWELL
20
BLDG
ENTRANCE
80
2000
20 FT
DECON ROOM
22
ELEV
RAILROAD
23
DOORS
RX BLDG
SAMPLE
24
50 FT
STATION
80
2000
ELEV
RX BLDG
AIR LOCK
25
RX
N OF FUEL
27
so
7000
BLDG
STORAGE POOL
117 FT
BETWEEN RX
28
1000
7000
ELEV
& FUEL POOL
CASK WASH
29
90
7000
AREA
RX BLDG
SPENT FUEL
30
90
3000
80 FT ELEV
COOLING SYSTEM
CONTACT E&RC TO DETERMINE
IF
MAX SAFE OPERATING VALUE IS
EXCEEDED
OEOP-01-UG
Rev. 40
Page 132 of 139
3.6.4.1
3.6
CONTAINMENT SYSTEMS
3.6.4.1
APPLICABILITY:
The secondary containment shall be OPERABLE.
MODES 1, 2, and 3,
During movement of recently irradiated fuel assemblies in
During operations with a potential for draining the reactor
vessel (OPDRVs).
ACTIONS
CONDITION
REQUIRED ACTION
COMPLETION TIME
A.1
Restore secondary
8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />
inoperable in MODE 1,
containment to
2, or 3.
OPERABLE status.
B. Required Action and
B.1
Be in MODE 3.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />
associated Completion
Time of Condition A
AND
not met.
B.2
Be in MODE 4.
36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />
C. -
NOTE------
inoperable during
LCO 3.0.3 is not
movement of recently
applicable.
irradiated fuel
assemblies in the
Suspend movement of
Immediately
or during OPDRVs.
recently irradiated
fuel assemblies in
the secondary
containment.
AND
(continued)
I
Brunswick Unit 2
Amendment No. 244 I
I
I
3.6-29
3.6.4.1
SURVEILLANCE REQUIREMENTS
SURVEILLANCE
FREQUENCY
Verify all secondary containment
24 months
equipment hatches are closed and sealed.
Verify one secondary containment access
24 months
door is closed in each access opening.
Verify each SGT subsystem can maintain
24 months on a
Ž 0.25 inch of vacuum water gauge in the
STAGGERED TEST
secondary containment for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> at a
BASIS
flow rate * 3000 cfm.
Amendment No. 244 I
Brunswick Unit 2
I
3.6-30
B 3.6.4.1
B 3.6
CONTAINMENT SYSTEMS
B 3.6.4.1
BASES
BACKGROUND
APPLICABLE
SAFETY ANALYSES
The function of the secondary containment is to contain and
hold up fission products that may leak from primary
containment following a Design Basis Accident (DBA).
In
conjunction with operation of the Standby Gas Treatment
(SGT)
System and closure of certain valves whose lines
penetrate the secondary containment, the secondary
containment is designed to reduce the activity level of the
fission products prior to relbase to the environment and to
isolate and contain fission products that are released
during certain operations that take place inside primary
containment, when primary containment is not required to be
OPERABLE, or that take place outside primary containment.
The secondary containment is a structure that completely
encloses the primary containment and those components that
may be postulated to contain primary system fluid.
This
structure forms a control volume that serves to hold up the
fission products.
It is possible for the pressure in the
control volume to rise relative to the environmental
pressure.
To prevent ground. level exfiltration while
allowing the secondary containment to be designed as a
conventional structure, the secondary containment requires
support systems to maintain the control volume pressure at
less than the external pressure.
Requirements for these
systems are specified separately in LCO 3.6.4.2, "Secondary
Containment Isolation Dampers (SCIDs)," and LCO 3.6.4.3,
"Standby Gas Treatment (SGT)
System."
There are two principal accidents for which credit is taken
for secondary containment OPERABILITY.
These are a loss of
coolant accident (LOCA)
(Refs.
I and 2) and a fuel handling
accident involving handling recently irradiated fuel (i.e.,
fuel that has occupied part of a critical reactor core
within the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) inside secondary containment.
The secondary containment performs no active function in
response to each of these limiting events; however, its leak
tightness is required to ensure that fission products
entrapped within the secondary containment structure will be
treated by the SGT System prior to discharge to the
environment.
(continued)
Revision No.
21 I
Brunswick Unit 2
B 3.6-69
B 3.6.4.1
BASES
APPLICABLE
Secondary containment satisfies Criterion 3 of
SAFETY ANALYSES
10 CFR 50.36(c)(2)(ii) (Ref. 4).
(continued)
LCO
An OPERABLE secondary containment provides a control volume
into which fission products that leak from primary
containment, or are released from the reactor coolant
pressure boundary components or irradiated fuel assemblies
located in secondary containment, can be processed prior to
release to the environment.
For the secondary containment
to be considered OPERABLE, it must have adequate leak
tightness to ensure that the required vacuum can be
established and maintained, at least one door in each access
to the Reactor Building must be closed, and the sealing
mechanism associated with each penetration (e.g., welds,
bellows, or O-rings) must be OPERABLE.
APPLICABILITY
In MODES 1, 2, and 3, a LOCA could lead to a fission product
release to primary containment that leaks to secondary
containment.
Therefore, secondary containment OPERABILITY
is required during the same operating conditions that
require primary containment OPERABILITY.
In MODES 4 and 5, the probability and consequences of the
LOCA are reduced due to the pressure and temperature
limitations in these MODES.
Therefore, maintaining
secondary containment OPERABLE is not required in MODE 4
or 5 to ensure a control volume, except for other situations
for which significant releases of radioactive material can
be postulated, such as during operations with a potential
for draining the reactor vessel (OPDRVs)
or during movement
of recently irradiated fuel assemblies in the secondary
containment.
Due to radioactive decay, secondary
containment is only required to be OPERABLE during fuel
handling accidents involving handling recently irradiated
fuel (i.e., fuel that has occupied part of a critical
reactor core within the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />).
ACTIONS
A.1
If secondary containment is inoperable, it must be restored
to OPERABLE status within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
The 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> Completion
Time provides a period of time to correct the problem that
is commensurate with the importance of maintaining secondary
(continued)
Revision No. 21 I
Brunswick Unit 2
B 3.6-70
B 3.6.4.1
BASES
ACTIONS
A.1 (continued)
containment during MODES 1, 2, and 3.
This time period also
ensures that the probability of an accident (requiring
secondary containment OPERABILITY) occurring during periods
where secondary containment is inoperable is minimal.
B.1 and B.2
If secondary containment cannot be restored to OPERABLE
status within the required Completion Time, the plant must
be brought to a MODE in which-the LCO does not apply.
To
achieve this status, the plant must be brought to at least
MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and to MODE 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
The
allowed Completion Times are reasonable, based on operating
experience, to reach the required plant conditions from full
power conditions in an orderly manner and without
challenging plant systems.
C.1 and C.2
Movement of irradiated fuel assemblies in the secondary
containment and OPDRVs can be postulated to cause
significant fission product release to the secondary
containment.
In such cases, the secondary containment is
the only barrier to release of fission products to the
environment.
Therefore, movement of recently irradiated
fuel assemblies must be immediately suspended if the
secondary containment is operable.
Suspension of this
activity shall not preclude completing an action that
involves moving a component to a safe position.
Also,
action must be immediately initiated to suspend OPDRVs to
minimize the probability of a vessel draindown and
subsequent potential for fission product release.
Actions
must continue until OPDRVs are suspended.
LCO 3.0.3 is not applicable while in MODE 4 or 5.
However,
since recently irradiated fuel assembly movement can occur
in MODE 1, 2, or 3, Required Action C.1 has been modified by
a Note stating that LCO 3.0.3 is not applicable.
If moving
recently irradiated fuel assemblies while in MODE 4 or 5,
LCO 3.0.3 would not specify any action.
If moving recently
irradiated fuel assemblies while in MODE 1, 2, or 3, the
fuel movement is independent of reactor operations.
Therefore, in either case, inability to suspend movement
(continued)
Revision No. 21 I
B 3.6-71
Brunswick Unit 2
B 3.6.4.1
BASES
ACTIONS
C.1 and C.2
(continued)
of recently irradiated fuel assemblies would not be a
sufficient reason to require a reactor shutdown.
SURVEILLANCE
REQUIREMENTS
Verifying that secondary containment equipment hatches and
one secondary containment access door in each access opening
are closed ensures that the infiltration of outside air of
such magnitude as to prevent maintaining the desired
negative pressure does not occur.
Verifying that all such
openings are closed provides adequate assurance that
exfiltration from the secondary containment will not occur.
In this application, the term "sealed" has no connotation of
leak tightness.
Maintaining secondary containment
OPERABILITY requires verifying one door in each access
opening is closed.
The 24 month Frequency for these SRs has
been shown to be adequate, based on operating experience,
and is considered adequate in view of other indications of
door and hatch status that are available to the operator.
The SGT System exhausts the secondary containment atmosphere
to the environment through appropriate treatment equipment.
To ensure that fission products are treated, SR 3.6.4.1.3
verifies that the SGT System will establish and maintain a
negative pressure in the secondary containment.
This is
confirmed by demonstrating that one SGT subsystem can
maintain Ž 0.25 inches of vacuum water gauge for I hour at a
flow rate * 3000 cfm.
The I hour test period allows
secondary containment to be in thermal equilibrium at steady
state conditions.
Therefore, this test is used to ensure
secondary containment boundary integrity.
Since this SR is
a secondary containment test, it need not be performed with
each SGT subsystem.
The SGT subsystems are tested on a
STAGGERED TEST BASIS, however, to ensure that in addition to
the requirements of LCO 3.6.4.3, either SGT subsystem will
perform this test.
Operating experience has demonstrated
these components will usually pass the Surveillance when
performed at the 24 month Frequency.
Therefore, the
Frequency was concluded to be acceptable from a reliability
standpoint.
(continued)
Revision No. 21 I
I
I
B 3.6-72
Brunswick Unit 2
B 3.6.4.1
BASES
(continued)
REFERENCES
1. NEDC-32466P, Power Uprate Safety Analysis Report for
Brunswick Steam Electric Plant Units 1 and 2,
September 1995.
2.
UFSAR, Section 15.6.4.
3.
Not used.
4.
Revision No.
21 I
I
R 3.6-73
Brunswick Unit 2
3.6.4.1
3.6
CONTAINMENT SYSTEMS
3.6.4.1
APPLICABILITY:
The secondary containment shall be OPERABLE.
MODES 1, 2, and 3,
During movement of recently irradiated fuel assemblies in
During operations with a potential for draining the reactor
vessel (OPDRVs).
ACTIONS
CONDITION
inoperable in MODE 1,
2, or 3.
B.
Required Action and
associated Completion
Time of Condition A
not met.
inoperable during
movement of recently
irradiated fuel
assemblies in the
or during OPDRVs.
REQUIRED ACTION
A.1
Restore secondary
containment to
OPERABLE status.
Be in MODE 3.
B.I
AND
B.2
C. 1
Be in MODE 4.
.
NOTE
LCO 3.0.3 is not
applicable.
Suspend movement of
recently irradiated
fuel assemblies in
the secondary
containment.
AND
COMPLETION TIME
8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />
12 hours
36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />
Immediately
(continued)
Amendment No. 218 I
Brunswick Unit I
3.6-29
I
I
I
3.6.4.1
ACTIONS
CONDITION
REQUIRED ACTION
COMPLETION TIME
c.
(continued)
C.2
Initiate action to
Immediately
suspend OPDRVs.
SURVEILLANCE REQUIREMENTS
SURVEILLANCE
FREQUENCY
Verify all secondary containment
24 months
equipment hatches are closed and sealed.
Verify one secondary containment access
24 months
door is closed in each access opening.
Verify each SGT subsystem can maintain
24 months on a
Ž 0.25 inch of vacuum water gauge in the
STAGGERED TEST
secondary containment for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> at a
BASIS
flow rate s 3000 cfm.
Amendment No. 218 1
Brunswick Unit I
3.6-30
I
B 3.6.4.1
B 3.6
CONTAINMENT SYSTEMS
B 3.6.4.1
BASES
BACKGROUND
The function of the secondary containment is to contain and
hold up fission products that may leak from primary
containment following a Design Basis Accident (DBA).
In
conjunction with operation of the Standby Gas Treatment
(SGT)
System and closure of certain valves whose lines
penetrate the secondary containment, the secondary
containment is designed to reduce the activity level of the
fission products prior to relkase to the environment and to
isolate and contain fission products that are released
during certain operations that take place inside primary
containment, when primary containment is not required to be
OPERABLE, or that take place outside primary containment.
The secondary containment is a structure that completely
encloses the primary containment and those components that
may be postulated to contain primary system fluid.
This
structure forms a control volume that serves to hold up the
fission products.
It is possible for the pressure in the
control volume to rise relative to the environmental
pressure.
To prevent ground. level exfiltration while
allowing the secondary containment to be designed as a
conventional structure, the secondary containment requires
support systems to maintain the control volume pressure at
less than the external pressure.
Requirements for these
systems are specified separately in LCO 3.6.4.2, "Secondary
Containment Isolation Dampers (SCIDs)," and LCO 3.6.4.3,
"Standby Gas Treatment (SGT)
System."
APPLICABLE
SAFETY ANALYSES
There are two principal accidents for which credit is taken
for secondary containment OPERABILITY.
These are a loss of
coolant accident (LOCA)
(Refs.
1 and 2) and a fuel handling
accident involving handling recently irradiated fuel (i.e.,
fuel that has occupied part of a critical reactor core
within the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) inside secondary containment.
The secondary containment performs no active function in
response to each of these limiting events; however, its leak
tightness is required to ensure that fission products
entrapped within the secondary containment structure will be
treated by the SGT System prior to discharge to the
environment.
(continued)
Revision No. 22 I
8 3.6-69
Brunswick Unit I
B 3.6.4.1
BASES
APPLICABLE
Secondary containment satisfies Criterion 3 of
SAFETY ANALYSES
10 CFR 50.36(c)(2)(ii) (Ref.
4).
(continued)
LCO
An OPERABLE secondary containment provides a control volume
into which fission products that leak from primary
containment, or are released from the reactor coolant
pressure boundary components or irradiated fuel assemblies
located in secondary containment, can be processed prior to
release to the environment.
For the secondary containment
to be considered OPERABLE, it must have adequate leak
tightness to ensure that the required vacuum can be
established and maintained, at least one door in each access
to the Reactor Building must be closed, and the sealing
mechanism associated with each penetration (e.g., welds,
bellows or O-rings) must be OPERABLE.
APPLICABILITY
In MODES 1, 2, and 3, a LOCA could lead to a fission product
release to primary containment that leaks to secondary
containment.
Therefore, secondary containment OPERABILITY
is required during the same operating conditions that
require primary containment OPERABILITY.
In MODES 4 and 5, the probability and consequences of the
LOCA are reduced due to the pressure and temperature
limitations in these MODES.
Therefore, maintaining
secondary containment OPERABLE is not required in MODE 4
or 5 to ensure a control volume, except for other situations
for which significant releases of radioactive material can
be postulated, such asduring operations with a potential
for draining the reactor vessel (OPDRVs)
or during movement
of recently irradiated fuel assemblies in the secondary
containment.
Due to radioactive decay, secondary
containment is only required to be OPERABLE during fuel
handling accidents involving handling recently irradiated
fuel (i.e., fuel that has occupied part of a critical
reactor core within the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />).
ACTIONS
A.1
If secondary containment is inoperable, it must be restored
to OPERABLE status within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
The 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> Completion
Time provides a period of time to correct the problem that
is commensurate with the importance of maintaining secondary
(continued)
Revision No. 22 I
Brunswick Unit I
B 3.6-70
B 3.6.4.1
BASES
ACTIONS
A.1
(continued)
containment during MODES 1, 2, and 3.
This time period also
ensures that the probability of an accident (requiring
secondary containment OPERABILITY) occurring during periods
where secondary containment is inoperable is minimal.
B.1 and B.2
If secondary containment cannot be restored to OPERABLE
status within the required Completion Time, the plant must
be brought to a MODE in which- the LCO does not apply.
To
achieve this status, the plant must be brought to at least
MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and to MODE 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
The
allowed Completion Times are reasonable, based on operating
experience, to reach the required plant conditions from full
power conditions in an orderly manner and without
challenging plant systems.
C.1 and C.2
Movement of recently irradiated fuel assemblies in the
secondary containment and OPDRVs can be postulated to cause
significant fission product release to the secondary
containment.
In such cases, the secondary containment is
the only barrier to release of fission products to the
environment.
Therefore, movement of recently irradiated
fuel assemblies must be immediately suspended if the
secondary containment is inoperable.
Suspension of this
activity shall not preclude completing an action that
involves moving a component to a safe position.
Also,
action must be immediately initiated to suspend OPDRVs to
minimize the probability of a vessel draindown and
subsequent potential for fission product release.
Actions
must continue until OPDRVs are suspended.
LCO 3.0.3 is not applicable while in MODE 4 or 5.
However,
since recently irradiated fuel assembly movement can occur
in MODE 1, 2, or 3, Required Action C.1 has been modified by
a Note stating that LCO 3.0.3 is not applicable.
If moving
recently irradiated fuel assemblies while in MODE 4 or 5,
LCO 3.0.3 would rot specify any action.
If moving recently
irradiated fuel assemblies while in MODE 1, 2, or 3, the
fuel movement is independent of reactor operations.
Therefore, in either case, inability to suspend movement
(continued)
Brunswick Unit 1
B 3.6-71
Revision No.
22 I
B 3.6.4.1
BASES
ACTIONS
C.1 and C.2
(continued)
of recently irradiated fuel assemblies would not be a
sufficient reason to require a reactor shutdown.
SURVEILLANCE
REQUIREMENTS
Verifying that secondary containment equipment hatches and
one secondary containment access door in each access opening
are closed ensures that the infiltration of outside air of
such magnitude as to prevent maintaining the desired
negative pressure does not occur.
Verifying that all such
openings are closed provides adequate assurance that
exfiltration from the secondary containment will not occur.
In this application, the term "sealed" has no connotation of
leak tightness.
Maintaining secondary containment
OPERABILITY requires verifying one door in each access
opening is closed.
The 24 month Frequency for these SRs has
been shown to be adequate, based on operating experience,
and is considered adequate in view of other indications of
door and hatch status that are available to the operator.
The SGT System exhausts the secondary containment atmosphere
to the environment through appropriate treatment equipment.
To ensure that fission products are treated, SR 3.6.4.1.3
verifies that the SGT System will establish and maintain a
negative pressure in the secondary containment.
This is
confirmed by demonstrating that one SGT subsystem can
maintain Ž 0.25 inches of vacuum water gauge for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> at a
flow rate * 3000 cfm.
The 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> test period allows
secondary containment to be in thermal equilibrium at steady
state conditions.
Therefore, this test is used to ensure
secondary containment boundary integrity.
Since this SR is
a secondary containment test, it need not be performed with
each SGT subsystem.
The SGT subsystems are tested on a
STAGGERED TEST BASIS, however, to ensure that in addition to
the requirements of LCO 3.6.4.3, either SGT subsystem will
perform this test.
Operating experience has demonstrated
these components will usually pass the Surveillance when
performed at the 24 month Frequency.
Therefore, the
Frequency was concluded to be acceptable from a reliability
standpoint.
(continued)
Revision No.
22 1
Brunswick Unit I
I
I
B 3.6-72
B 3.6.4.1
BASES (continued)
REFERENCES
1. NEDC-32466P, Power Uprate Safety Analysis Report for
Brunswick Steam Electric Plant Units I and 2,
September 1995.
2.
UFSAR, Section 15.6.4.
3.
Not used.
4.
5.
(2) (ii).
6. Regulatory Guide 1.52, Revision 1.
Revision No. 22 1
I
B 3.6-73
Brunswick Unit I
CP&L Nuclear Fuels Mgmt. & Safety Analysis
B2C14 Core Operating Limits Report
Design Calc. No. 2821-0554
Page 9, Revision 0
Table 1
MCPR Limits
(EOC-RPT Not Required)
Steady State, Non-pressurization Transient MCPR Limits
Fuel Type
GEl3
1.29
Al0
1.43
Pressurization Transient MCPR Limits, OLMCPR (100%P):
Turbine Bypass System Operable
Normal and Reduced Feedwater Temperature
Exposure Range:
Exposure Range:
MCPR Option
Fuel Type
BOC to EOFPC-2205 MWd/MT
EOFPC-2205 MWd/MT to EOC
A
GE13
1.39
1.46
A10
1.55
1.62
B
GEl3
1.34
1.38
A10
1.49
1.53
Pressurization Transient MCPR Limits, OLMCPR (100%P):
Turbine Bypass System Inoperable
Normal and Reduced Feedwater Temperature
MCPR Option
Fuel Type
A
GEl3
1.48
A10
1.65
B
GEl3
1.40
Al0
1.56
This Table is referred to by Technical Specifications 3.2.2, 3.4.1 and 3.7.6.
CP&L Nuclear Fuels Mgmt. & Safety Analysis
Design Calc. No. 2B21-0554
B2C14 Core Operating Limits Report
Page 22, Revision 0
Figure 11
GE13 Flow-Dependent MCPR Limit, MCPR(F)
20
25
30
35
40
45
50
55
60
65
70
75
80
85
90
95 100 105 110 115 120
Core Flow (% Rated)
CC.
1.80
1.75
1.70
1.65
1.60
1.55
1.50
1.45
1.40
1.35
1.30
1.25
1.20
1.15
1.10
CP&L Nuclear Fuels Mgmt. & Safety Analysis
Design Calc. No. 2B21-0554
B2C14 Core Operating Limits Report
Page 24, Revision 0
Figure 12
Power - Dependent MCPR Limit, MCPR (P)
3.30
3.20
3.10
3.00
2.90
2.80
2.70
2.60
2.50
2.40
2.30
2.20
2.10
2.00
1.90
1.80
1.70
1.60
1.50
1.40
1.30
1.20
1.10
1.00
i .
-
I
I
I
1
--
Core Flow
Turbine B
-Inoper
Core Flow
Turbine E
Opera
oror
Fovlow
Turbine
'
Inoper
Core Flow
Turbine B
Opera
Rated MCPR Multiplier (Kp)
I
I
I
I
I
S
I
I _____
_____
I
50%
3ypass
able
,> 50%
Bypass
ible
/ <50%I
Bypass
"able
<50%
Iypass
ble
This Figure is Referred To By
Technical Specification 3.2.2, 3.4.1, 3.7.6
20
25
30
35
40
45
50
55
60
65
70
75
80
85
90
95
100
PBYPASS
Power (% Rated)
T 1
1 1Z
Operating Limit MCPR(P) = Kp*Operating Limit MCPR(
For P < 25%:.
No Thermal Limits Monitoring Required
No Limits Specified
For 25% < P < PSYPASS:
Where PBYPASS = 30%
Kp = Maximum of 1.481 or KpLp
For Core Flow * 50% & Turbine Bypass Operable,
Kp~p = [1.90 + 0.02 (30% - P)] /OLMCPR(100)
For Core Flow > 50% & Turbine Bypass Operable
KpLp = [2.20 + 0.02 (30% - P)] / OLMCPR(100)
For Core Flow *50% & Turbine Bypass Inoperable,
Kp~p = [1.96 + 0.072 (30% - P)] / OLMCPR(100)
For Core Flow > 50% & Turbine Bypass Inoperable
KpLp = [2.81 + 0.05 (30% - P)] / OLMCPR(l00)
For 30% < P < 45%:
K, = 1.28 + 0.0134 (45% - P)
For 45% <_ P < 60%:
Kp = 1.15 + 0.00867 (60% - P)
For P 2:60%:
Kp = 1.00 4 0.00375 (100% - P)
'S
I
I
I
I
r
I
J
(ioo)
Primary Containment Air Lock
3.6.1.2
3.6
CONTAINMENT SYSTEMS
3.6.1.2
Primary Containment Air Lock
APPLICABILITY:
The primary containment air lock shall be OPERABLE.
MODES 1, 2, and 3.
ACTIONS
.NOTES -----------------------------------
1. Entry and exit is permissible to perform repairs of the air lock
components.
2.
Enter applicable Conditions and Required Actions of LCO 3.6.1.1, "Primary
Containment," when air lock leakage results in exceeding overall
containment leakage rate acceptance criteria.
..
.
.
..
.
.
.
..
...
.
.
..
.
.--------
CONDITION
A.
One primary
containment air lock
door inoperable.
REQUIRED ACTION
NOTES --------
- - -
1. Required Actions A.1,
A.2, and A.3 are not
applicable if both doors
in the air lock are
inoperable and
Condition C is entered.
2. Entry and exit is
permissible for 7 days
under administrative
controls.
Verify the OPERABLE
door is closed.
A.1_
AND
COMPLETION TIME
2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />
(continued)
Amendment No. 233
Brunswick Unit 2
3.6-3
Primary Containment Air Lock
3.6.1.2
ACTIONS
CONDITION
REQUIRED ACTION
COMPLETION TIME
A. (continued)
A.2
Lock the OPERABLE
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />
door closed.
AND
A.3
NOTE------
Air lock doors in
or areas with limited
access due to
inerting iay be
verified locked
closed by
administrative means.
Verify the OPERABLE
Once per 31 days
door is locked
closed.
NOTES--------
air lock interlock
1. Required Acfions B.1,
mechanism inoperable.
B.2, and B.3 are not
applicable if both doors
in the air lock are
inoperable and
Condition C is entered.
2.
Entry into and exit from
permissible under the
control of a dedicated
individual.
B.1
Verify an OPERABLE
2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />
door is closed.
AND
(continued)
Amendment No. 233
Brunswick Unit 2
3.6-4
I
CAROLINA POWER & LIGHT COMPANY
Information
BRUNSWICK NUCLEAR PLANT
Use
PLANT OPERATING MANUAL
VOLUME I
BOOK 2
ADMINISTRATIVE INSTRUCTION
UNIT
0
1111111II111111
II1i11
U111111111lll111111li11111i
0AI-1 07
INSTRUCTIONS FOR WORKING IN HOT
ENVIRONMENTS
REVISION
10
EFFECTIVE DATE
02/15/99
Signature and Date on File
Industrial Hygiene and Safety
Representative
Siqnature and Date on File
Page 1 of 20
Sponsor
Approval
Outages and Scheduling Manager
Date
Date
I OAI-107
Rev. 10
REVISION SUMMARY
Removal of reference to LPU and change in signature authority.
LIST OF EFFECTIVE PAGES
Page(s)
Revision
1-20
10
OAI-107
Rev. 10
Page 2of20
TABLE OF CONTENTS
SECTION
PAGE
1.0
PURPOSE ............................................................................................................
4
2.0
REFERENCES .....................................................................................................
4
3.0
DEFINITIO NS ........................................................................................................
5
4.0
RESPO NSIBILITIES .............................................................................................
7
5.0
INSTRUCTIO NS ..................................................................................................
9
5.1
Precautions .................................................................................................
10
5.2
Heat Illness Prevention and First Aid ..........................................................
11
5.3
Use of Recom m ended Action Tim es ..........................................................
13
5.4
Heat Stress Evaluation ................................................................................
14
5.5
Use of Ice Vests ..............................................................................................
15
5.6
Use of Supplied Air Hood/Helm ents ..............................................................
15
5.7
Designated Drinking Areas .........................................................................
15
ATTACHMENTS
1
W ork Rate Guidelines .........................................................................................
17
2
Recommended Action Times
...................................
18
3
Cool Vest Flow Path ........................................................................................
19
4
Heat Stress Evaluation Form ..........................................................................
20
IOAI-107
I
Rev. 10
Page 3 of 20
1.0
PURPOSE
The purpose of this procedure is to provide guidance to all employees for
preventing heat-induced occupational illnesses or injuries, thus, enhancing
employee safety and increasing productivity.
Heat related fatigue can lead to decreased job performance as well as
contributing to work place accidents and illness. Productivity and worker safety
can be enhanced through the management of heat stress.
This program is based EPRI Report NP-4453 "Heat Stress-Management
Program for Nuclear Power Plants". The EPRI report outlines a three Step
method for managing heat stress. These steps are: environmental assessment
by trained evaluators, control methods, and training.
2.0
REFERENCES
2.1
EPRI NP 4453, Heat Stress Management Program for Nuclear Power Plants
2.2
NIOSH Publication 72-10269, Criteria for a Recommended Standard:
Occupational Exposure to Hot Environments
2.3
NIOSH, The Industrial Environment - Its Evaluation and Control, Chapters
30, 31, and 38
2.4
The American Industrial Hygiene Association, Heating and Cooling for Man
in Industry
2.5
E. Kamon and C. Ryan, Effective Heat Strain Index Using Pocket Computer,
AIHA Journal, August 1981
2.6
American Red Cross, Advanced First Aid and Emergency Care, Second
Edition
2.7
E&RC-01 36, Setup and Use of Airline Respiratory Protection Devices
2.8
E&RC-0229, Control & Use of HEPA Vacuum Cleaners and Mobile Air
Filtration Units
I OAI-107
I
Rev. 10
Page 4 of 20
3.0
DEFINITIONS
3.1
Heat Stress
The physiological stress which occurs when the body's temperature rises
above normal. This occurs when the body produces or gains more heat
than it is capable of losing. It is caused by any combination of air
temperature, thermal radiation, humidity, air flow, restrictive clothing, and
physical work load which may result in elevated core body temperature and
subsequent illness.
3.2
Action Time
An estimate of the length of time workers may be exposed in hot
environments and not suffer heat stress disorders, used for planning
purposes. The length of Action Times is not absolute because of worker
variability in response to heat. The times reflect an approximate 20F rise in
body temperature.
3.3
Protective Clothing (POs)
Items worn to prevent radioactive contamination.
3.4
Wet Suit
Full body impermeable plastic suit worn to prevent radioactive skin
contamination.
3.5
Chemical Suit
Full body impermeable neoprene or Tyvek coveralls worn to prevent
chemical skin contamination.
3.6
Personal Cooling Device
Equipment such as ice vests or vortex cooling units placed on a person to
minimize heat gain and/or increase heat loss.
3.7
Supplied Air Hood/Helmet
Air-supplied hood respirator which delivers respirator air over the head and
body.
Page 5 of 20 1
OAI-107
Rev. 10
3.0
DEFINITIONS
3.8
Wet Bulb Globe Thermometer - used to establish the work area
Temperature Index Heat Stress that allows for the effects of Humidity, and
Radiant Heat, that modify dry bulb temperatures.
3.9
Acclimation
The gradual process of improved heat tolerance after continuous exposure
to heat. Acclimation consists of reduced heart rate, increased sweat
production, production of less salty sweat, and lower body temperature.
3.10
Dry Bulb Temperature
The temperature as measured by a standard thermometer without respect
to humidity or radiant heat.
3.11
Globe Temperature
Temperature resulting from radiant heat sources, measured with a black
globe thermometer.
3.12
High Heat Stress Job/Work
Any job in which the calculated Action Time is less than 30 minutes.
3.13
Metabolic Heat Load
Heat generated from physical work (muscle contraction).
3.14
Moderate Heat Stress Job/Work
Any job/work in which the calculated Action Time is greater than 30 minutes
but less than 240 minutes.
3.15
Relative Humidity
The amount of moisture in the air compared to the amount of moisture the
air can hold for a given temperature.
OAl-107
Rev. 10
1
Page6of 20
3.0
DEFINITIONS
3.16
Recovery Period
Recovery time allocated to workers who have performed work in hot
environments. Recovery shall not take place in a hot environment. Water
should be available for consumption in the recovery area.
3.17
Self-Determination
Allowing for worker discretion to exit High Heat Stress Work Areas when
he/she feels the onset of heat stress symptoms.
3.18
Time Keeper
A person responsible to monitor action times.
3.19
Wet-Bulb Temperature (natural)
The temperature of the air when it is subjected to evaporative cooling.
3.20
High Temperature Work Level
Any work area > 95°F.
3.21
Designated Drinking Areas (DDA)
Specific areas designated within the Radiation Control Area to allow
ingestion of liquids as part of the Heat Stress Program.
4.0
RESPONSIBILITIES
4.1
General Manager - Brunswick Plant
The General Managers - Brunswick Plant are responsible for the
implementation of this procedure to ensure that personnel who perform work
in high temperature environments follow the guidance of this procedure.
OAI-10 7
I
Rev. 10
1
Page 7 of 20
4.0
RESPONSIBILITIES
4.2
Managers
4.2.1
Managers will ensure that the supervisors reporting to them utilize
this procedure and follow its guidance when planning work in hot
environments.
4.2.2
Managers shall ensure that training or instruction on heat stress
mitigation is arranged for and conducted for employees prior to initial
work in high temperature environments.
4.3
Supervisor
4.3.1
The supervisor or person in charge of the job is responsible for
following the guidance in this procedure when planning a job that is to
be performed within a hot environment, and ensure that heat stress
mitigation has been considered during job planning.
4.3.2
Safety of employees shall be the responsibility of supervision
whenever employees must enter or work in a hot environment or may
be subject to heat stress causing conditions.
4.3.3
Shall ensure that heat stress caused illnesses are recorded on the
SAF-CPL-009 form as appropriate.
4.4
Individuals
Each individual is responsible for complying with:
4.4.1
The requirements of this procedure.
4.4.2
Written and/or oral instructions given by supervision on mitigating
heat stress.
4.4.3
Instructions given by the supervisor on the use of body cooling
devices.
4.4.4
Being attentive to symptoms of heat stress while working in hot
environments, and stopping work and notifying their supervisor if they
feel ill due to heat stress.
4.4.5
Each individual is responsible for being prepared to work in a hot
environment; rested and have no medical problems that would be
affected by heat related work.
OAI-107
I
Rev. 10
1
Page8of2o0
4.0
RESPONSIBILITIES
4.5
Heat Stress Evaluators
Individuals trained in the use of this procedure and the WBGT thermometer
for the purpose of evaluating the potential for heat stress during jobs
completed on site.
4.6
Industrial Hygiene/Safety Representative
Shall provide technical assistance on plant heat stress issues.
4.7
Training Department
Is responsible for teaching heat stress in initial GET and in the annual
retraining.
5.0
INSTRUCTIONS
5.1
Precautions
CAUTION
Workers should never work alone in high heat stress areas.
5.1.1
If any individual begins to feel symptoms of heat illness, he/she shall
immediately exit the area, de-suit, notify the job supervisor, rest in a
cool area, and drink plenty of fluids. Seek medical help if necessary,
by calling the Control Room (extension 4444).
OAI-107
Rev. 10
Page 9 of 20
5.1
Precautions
5.1.2
All jobs in high temperature environments should address heat stress
prevention controls in the planning stages.
1 .
In situations where individuals know that their work schedule for the
next day will involve entering a heat stress area, they should drink
plenty of liquids in the 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to reporting to work.
2.
Action Times, recovery times, personnel rotation, and the use of body
cooling devices should be addressed.
3.
Whenever possible, engineering controls should be used to
eliminate/reduce the exposure (i.e., isolation of the heat source,
introduction of cooled air, circulation of present air, reduced humidity,
etc.). The impact of these engineering controls should be reviewed
with the Environmental and Radiation Control (E&RC) unit for jobs in
radiologically controlled areas. When introducing air circulation and
handling devices, follow the guidelines provided in E&RC-0229.
5.1.3
Individuals who work in hot environments may become dehydrated
due to sweating. Water should be replaced at rest breaks to prevent
heat-related illness. Employees should be encouraged to drink small
amounts often, regardless of thirst. Salt tablets are not
recommended. Liquids designed to replace these salts (i.e.,
Gatorade or water) are recommended as replacement fluids.
5.1.4
Individuals who work in high temperature environments must
periodically rest in a cooler area to shed body heat. Duration of
breaks, extent of clothing removal, and rest area should be
determined by the job supervisor, using the guidance in Section 5.2.
Certain employees may require varying rest periods with some
requiring less time than shown in Section 5.2.
5.1.5
Individuals who will be working in high temperature environments for
the first time will be more susceptible to heat illness than those
accustomed to hot work. After working in hot environments for
several days, their bodies may adjust to heat exposure and they may
tolerate longer heat exposures at higher work rates (acclimatization).
OAI-107
I
Rev. 10
Page l0of 20
5.1
Precautions
5.1.6
Individuals vary greatly in their tolerance to heat exposure. Factors
which may affect heat tolerances may include:
- Age
- Weight
- Sex
- Physical fitness
- General health
- Colds, viruses, and infection
- Some medications
- Consumption of alcoholic beverages
5.2
Heat Illness Prevention and First Aid
NOTE:
The following first aid actions are recommendations only. If any
individual begins to feel symptoms of heat illness, the worker should immediately
exit the area, notify the supervisor, and seek first aid/medical attention.
5.2.1
Workers should be encouraged to drink one pint of water/fluid per
hour of scheduled work prior to entering high heat areas.
5.2.2
Workers shall be encouraged to drink water/fluid after high
temperature work to maintain fluid balance.
5.2.3
Where feasible, high temperature work shall be scheduled to
minimize thermal stress in the work area. This includes scheduling
work at times where the WGBT and or metabolic heat load are lower
and or anti-C requirements are less restrictive.
QAI-107
Rev. 10
Page 11 of 20
5.2
Heat Illness Prevention and First Aid
5.2.4
Prior to re-entering a High Temperature Work area, workers shall
have an adequate recovery period to dissipate excess heat and
replace water. Recovery shall take place in a cool location (less than
80 degrees F) where drinking water is available.
The length of the recovery period depends on the length of exposure
and the maximum stay time of the job. Recovery periods of up to one
hour may be necessary for jobs which approach or exceed the
planned stay times. The following formula shall be used as a general
guide for determining the minimum length of recovery period. Actual
recovery time can be modified only by agreement of both worker and
supervisor. This shall be documented on Attachment 4, Heat Stress
Evaluation Form.
NOTE:
All times used should be in minutes
REC=
AET x 60
REC ---------
Recovery Time
AET
Actual Exposure Time to the Hot
Environment
MST ---------
Actual Stay Time or Action Time from
Attachment 2
5.2.5
Heat Cramps - are muscle spasms due to a loss of salt through
sweating. The legs, arms, and abdominal muscles are the most
commonly affected muscle groups. Cramps can also result from
drinking large amounts of water without electrolytes. Heat cramps
may be a sign of approaching heat exhaustion.
First Aid - Rest in cool area, drink water or liquids containing
electrolytes and eat food high in salt content.
5.2.6
Heat Exhaustion - is dehydration caused by prolonged heavy
sweating. There is insufficient flow of blood to the brain (blood is
shunted to the skin to lose heat). Symptoms include dry mouth,
excessive thirst, loss of coordination, headache, dizziness, fatigue,
pale and shaky look, and cool clammy skin. This condition may
develop into heat stroke.
First Aid - Rest in a cool area, lie down, elevate feet, apply cool wet
clothes, fan with air, and drink liquids.
OAI-107
Rev. 10
Page 12 of 20
5.2
Heat Illness Prevention and First Aid
5.2.7
Heat Stroke - is a serious medical emergency caused by a failure of
the body's cooling mechanisms. Symptoms include hot, dry skin
(sweating stops), extremely high body temperatures, chills,
convulsions, and unconsciousness.
First Aid: Immediate, rapid cooling of the body is necessary. Use
safety showers, move air over the body with a fan or by fanning, or
cover the body with a wet sheet. Call the Control Room (extension
4444) to seek immediate medical attention.
5.3
Use of Recommended Action Times
5.3.1
A fundamental rule of heat stress management is that
self-determination by the worker should take precedence over other
factors. Attachment 2, Recommended Action Times, may be
modified and even exceeded only by agreement of supervisor and
worker. Attachment 4, Heat Stress Evaluation Form, shall document
this change, however a worker must leave a high temperature work
area if he/she feels the onset of heat stress symptoms.
5.3.2
By using the recommended Action Time as a general guideline, and
assessing the physical condition of his workers, the job supervisor
can determine how long his workers may be able to work before rest
breaks are given. Workers must, and have the right to, exit the hot
environment prior to the time limit if they feel that they cannot
continue.
5.3.3
Work should be planned so that an adequate number of workers are
prepared to work in the high temperature environment. The
supervisor should also consider whether there are enough workers to
complete the task if some workers cannot last the recommended
time.
5.3.4
A worker may extend his Action Time long enough to bring the task to
a satisfactory and safe stopping point if he/she feels fully capable of
staying longer and has supervision approval. In no case should this
extended time be more than 25% above the recommended time limit
stated in Attachment 4.
OAI-107
Rev. 10
Page 13of 20
5.4
Heat Stress Evaluation
5.4.1
5.4.2
5.4.3
The heat stress evaluation process involves assessing the variables
that affect heat stress, including WBGT measurements, metabolic
work load, and clothing type. These factors are converted to
recommended action times for planning purposes. A Heat Stress
Evaluation Form (Attachment 4), or other record containing the same
information should be used for heat stress job planning.
All potential High Heat Stress Work shall be identified.
For initial evaluations, the WBGT shall be measured using a WBGT
Meter. Measurements shall be representative of the work area
thermal load. Succeeding evaluations may be based solely on dry
bulb temperature when a correlation between dry bulb temperature
and WBGT is established.
NOTE:
Care must be used when conducting WBGT readings so as to not
create an ALARA concern for the rare case of a worker being assigned both a
dose and a heat stress action time; the most limiting shall be used.
5.4.4
The type of work clothing required for the job shall be determined.
The categories include: street clothing, single cotton blend or
paper/Tyvek coveralls, double cotton blend coveralls, and single
cotton blend coveralls with impervious plastic (rain suit) or Tyvek
outer suit.
5.4.5
The metabolic heat load shall also be assessed using Attachment 1
as a guide.
5.4.6
The Job Action Time is determined from Attachment 2 using the
WBGT reading, clothing type, and metabolic load. Action Times are
used for job planning. Action Times stated in Attachment 2 are not
absolute because of the great variability in worker response to heat
stress. Many healthy/acclimatized workers could exceed the Action
Times without suffering any adverse effects. However, a few workers
could experience heat stress symptoms prior to reaching the
maximum Action Time.
5.4.7
A re-evaluation is necessary, if there are changes in the WBGT
(+/-30F), the metabolic work load category or the required clothing
type during the course of the operation.
OAI-1 07
Rev. 10
Page 14 of 20
5.5
Use of Ice Vests 5.5.1
By using Attachment 2, the job supervisor can determine if ice vests
would be beneficial for the job.
1.
Proper handling of ice vests and ice packs is essential to maintaining
an adequate supply in good condition. The flow path (Attachment 3)
for the use of the vests must be followed by all workers to ensure
availability.
2.
Job supervisors shall ensure their workers are trained in the use of
ice vests prior to using the vests in hot environments.
3.
The ice vest should be worn so that the vest fits snugly. A shirt
should be worn under the vest to prevent frost burn. Under garment
shirts are available upon request.
4.
As much as practical, the ice vest should not be donned until just
prior to entering the hot environment.
5.
The ice vest will provide cooling only while the ice is melting. Once
the ice has melted, body temperature will increase quickly. Workers
should monitor their condition and exit the work area as soon as the
ice vest has lost its cooling effectiveness.
5.6
Use of Supplied Air Hood/Helmets
Supplied air hood/helmets are used mainly as respirators and their uses are
authorized only by the E&RC Unit.
5.7
Designated Drinking Areas
NOTE:
DDA's are only a part of the total Heat Stress Program. Control of the
Areas to ensure that sanitary conditions exist and radiological controls are
followed will dictate that the number and locations are limited.
5.7.1
RC Supervision will evaluate the request for DDA's authorize their
placement and determine survey requirements.
OAI-107
I
Rev. 10
Page 15 of 20
5.7
Designated Drinking Areas
5.7.2
Areas will be bounded off and posted similar to the following:
Designated Drinking Area
1.
Whole Body Frisk Required Prior to Entry
2.
Workers Shall Drink Only Within the DDA
3.
"NO ANTICONTAMINATION CLOTHING ALLOWED"
5.7.3
Personnel SHALL NOT ENTER the DDA dressed in, or with
anticontamination clothing.
5.7.4
Personnel SHALL PERFORM a WHOLE BODY FRISK PRIOR TO
ENTRY and DRINKING in a DDA.
1
Rev. 10
Page 16 of 20
ATTACHMENT 1
Page 1 of 1
Work Rate Guidelines
CATEGORY
TYPE OF ACTIVITY
EXAMPLES
- sitting with moderate arm and
trunk movement
- sitting with moderate arm and
leg
- standing, light work at machine
or bench
- standing, light work with some
walking and minimal climbing
- inspections and surveys with
minimal climbing
- supervising or monitoring
areas or equipment
- bench work
standing with moderate work
walking with moderate lifting
or pushing
- painting
floor cleaning
MODERATE
walking with occasional
ladder or stair climbing
- insulation removal or
installation
- fitting and welding light
pieces
- surveys and inspections with
moderate climbing
- walking with frequent stair
- scaffold erection
and ladder climbing
- rigging
HEAVY
- transporting equipment by hand
- heavy lifting, pushing, or
- manual decontaminating
pulling
- shoveling
- mopping
OAI-107
Rev. 10
1
Page 17 of 20
LIGHT
ATTACHMENT 2
Page 1 of 1
Recommended Action Times*
Single Cotton Blend or Paper
Double Cotton Blend or Paper
Single Cotton Blend Plus Impervious
Coveralls
Coveralls
Garment
LIGHT
MODERATE
HEAVY
LIGHT
MODERATE
HEAVY
LIGHT
MODERATE
HEAVY
SUPPLIED
(F°)
WORK
WORK
WORK
WORK
WORK
WORK
WORK
WORK
WORK
AIR HOOD/
HELMETS
(HOURS)
75-78.9
NIL
NL
150m
NL
180m
90m
190m
65m
40m
79-82.9
NL
145m
80m
240m
80m
50m
130m
45m
30m
4
83-86.9
225m
75m
45m
165m
55m
35m
90m
35m
20m
4
87-90.9
150m
50m
35m
105m
40m
25m
55m
30m
15m
4
91-93.9
105m
40m
25m
80m
35m
20m
45m
25m
15m
3
94-97.9
75m
35m
15m
50m
25m
15m
35m
20m
3 98-100.9
50m
25m
45m
20m
25m
15m
PPCR
3
101-104.9
35m
20m
30m
15m
20m
PCOR
PC
2 1/2
105-108.9
25m
15m
25m
15m
2 1/2
109-111.9
20m
2Om
21/2
112-115,9
15m
POR
15m
2
NOTE:
Ice vests will provide cooling for 45 to 120 minutes depending on conditions. Work time limits should be
determined by the user based on when his/her ice vest no longer provides cooling.
m = minutes
h = hours
NL = No Limit
PCR = Personal Cooling Recommended
- Without Personal Cooling Equipment
OAI-107
Rev. 10
Page 18 of 20
ATTACHMENT 3
Page 1 of 1
Cool Vest Flow Path
Vest in Bin
Acquired By User
- May reuse vest with
fresh ice packs
Used in
Cont. Area
"I
Leaves Area
4I
Whole Body Frisk
with Vest on
Vest Clean*
I
Remove and
frisk ice pack
Vest Cont.
I
Put vest and
ice in a
Yellow Baa
Ice Pack
Cont.
I
Put ice
pack in
Yellow Bag
Used in
Clean Area
I
Leaves Area*
Returns Ice Pack
to Freezer
4I
Monitors Vest
With SAM or other
frisking devices
I
Places vest in Vest
Laundry Barrel or
Other Designated
Location
I
Return Cont.
ice packs and
vests to
Personnel
Decon
or other
designated
location
Monitors vest
with SAM or
other frisking devices
4I
Place vest in Vest
Laundry Barrel
or other
designated location
I
Vests laundered
and frisked
Vests put in bins
OAI-107
Rev. 10
Page 19of 20
Ice Pack
Clean
4,
Return ice
pack
to freezer
ATTACHMENT 4
Page 1 of 1
Heat Stress Evaluation Form
(Section 5.4)
Task(s):
Supervisor:
Job Date:
Job Location:
Number of Workers:
Est. Person-Hours:
Plant Status (for job planning use):
5.2.4 Modified Action Time/Recover Period (AS REQUIED)
Signatures
(Worker)
(Supervisor)
(Time Keeper)
5.4.4 CLOTHING TYPE
(Circle)
single coveralls
(cotton blend)
or
paper coveralls
double coveralls
(cotton blend)
or
paper coveralls
impervious
outer &
cotton blend
inner
5.4.5 METABOLIC
HEAT LOAD
(Circle)
5.4.3 Dry Bulb =
F__
5.4.6 ACTION TIME =
CONTROL METHODS:
low
Wet Bulb =
0F
moderate
Globe Temp =
F
high
WBGT =
'F
minutes
5.2.4 Recovery Period =
minutes
Signature (Evaluator):
Date:
Signature (Job Supervisor):
Date:
OAI-107
Rev. 10
Page 20 of 20
Attachment 6
Page 1 of 19
Attachment 6
Reactor Water Level Caution
(Caution 1)
OEOP-01-UG
Rev. 40
1
Page 83 of 139
Attachment 6
Page 2 of 19
ATTACHMENT 6
REACTOR WATER LEVEL CAUTION
(Caution 1)
A reactor water level instrument may be used to determine reactor water level
only when the conditions for use as listed in Table 1 are satisfied for that
instrument.
TABLE 1
CONDITIONS FOR USE OF REACTOR WATER LEVEL INSTRUMENTS
NOTE
Reference leg area drywell temperature is
determined using Figure 13,
ERFIS,
or Instructional Aid based on Figure 13.
NOTE
If
the temperature near any instrument run is
in the UNSAFE region of the
REACTOR SATURATION LIMIT (Figure 14),
the instrument may be unreliable due to
boiling in
the run.
NOTE
Immediate reference leg boiling is
not expected to occur for short duration
excursions into the unsafe region due to heating of the drywell.
The thermal
time constant associated with the mass of metal and water in the reference leg
will prohibit immediate boiling of the reference leg.
Reference leg boiling
is an obvious phenomenon.
Large scale oscillations of all
water level
instruments associated with the reference leg that is boiling will occur.
This occurrence will be obvious and readily observable by the operator.
Additionally,
if
the operator is
not certain whether boiling has occurred, he
can refer to plant history as provided on water level recorders or ERFIS.
Reference leg boiling is indicated by level oscillations without corresponding
pressure oscillations.
Instrument
Conditions for Use
Narrow Range Level Instruments
Unit 1 Only:
The indicated level is
C32-LI-R606A,
B,
C (NO04A,
B, C)
in the SAFE region of Figure 15.
C32-LPR-R608
(NO04A,
B)
Indicating Range 150-210 Inches
Unit 2 Only:
The indicated level is
Cold Reference Leg
in the SAFE region of Figure 15A.
Shutdown Range Level Instruments
The indicated level is
in the SAFE
B21-LI-R605A,
B (N027A,
B)
region of Figure 16.
Indicating Range 150-550 Inches
Cold Reference Leg
To determine reactor water level at
the Main Steam Line Flood Level
(MSL),
see Figure 21.
NOTE
Figure 21 has two curves:
The upper
curve is for reference leg area
drywell temperature equal to or
greater than 200'F.
The lower curve
is for reference leg area drywell
temperature less than 2000 F.
OEOP-01-UG
Rev. 40
Page 84 of 139
Attachment 6
Page 3 of 19
ATTACHMENT 6 (Cont'd)
TABLE 1 (Cont'd)
Wide Range Level Instruments
B21-LI-R604A,
B (N026A,
B)
C32-PR-R609 (N026B)
Indicating Range 0-210 Inches
Cold Reference Leg
1.
Temperature on the Reactor
Building 50' below 140 0 F
(B21-XY-5948A A2-4,
B21-XY-5948B A2-4,
ERFIS
Computer Point B21TAI02,
B21TAl03)
AND
2.
IF the reference leg area
drywell temperature is
in the
UNSAFE region of the Reactor
Saturation Limit (Figure 14),
THEN the indicated level is
greater than 20 inches
IF the reference leg area
drywell temperature is
in the
SAFE region of the Reactor
Saturation Limit (Figure 14),
THEN the indicated level is
greater than 10 inches.
QEOP-01-UG
I
Rev. 40
Page 85 of 139
Instrument
I
Conditions for Use
Attachment 6
Page 4 of 19
ATTACHMENT 6 (Cont'd)
TABLE 1 (Cont'd)
Fuel Zone Level Instruments
B21-LI-R610
(N036)
B21-LR-R615
(N037)
Indicating Range -150
-
+150 Inches
Cold Reference Leg
OEOP-01-UG
I
Rev. 40
1
Page 86 of 139
Instrument
I
Conditions for Use
1.
IF the reference leg area
drywell temperature is less than
4400P, THEN the indicated level
is greater than -150
inches
IF the reference leg area
drywell temperature is greater
than or equal to 4400F,
THEN the
indicated level is greater than
-130 inches.
AND
2.
Reactor Recirculation Pumps are
shutdown.
NOTE
To determine reactor water level at
TAF,
see Unit 1 Only:
Figure 17 and
Unit 2 Only:
Figure 17A
To determine reactor water level at
the minimum steam cooling level
(LL-4),
see Unit 1 Only:
Figure 18
and Unit 2 Only:
Figure 18A
To determine reactor water level at
the minimum zero injection level
(LL-5),
see Unit I Only:
Figure 19
and Unit 2 Only:
Figure 19A
To determine reactor water level at
90 inches, see Figure 20.
Continued on next page.
Attachment 6
Page 5 of 19
ATTACHMENT 6 (Cont'd)
TABLE 1 (Cont'd)
OEOP-01 -UG
Rev. 40
Page 87 of 139
Instrument
I
Conditions for Use
NOTE
Each figure has two curves:
The upper curve for reference leg
area drywell temperature greater than
200 0 F.
The lower curve for reference
leg area drywell temperature less
than or equal to 2000 F.
If
containment conditions are such that
reference leg area temperatures could
not be controlled and maintained less
than the 2000 F requirement, then the
upper lines on the graph should be
utilized.
NOTE
These level instruments are valid for
indication with RHR LPCI flow.
Attachment 6
Page 6 of 19
ATTACHMENT 6 (Cont'd)
FIGURE 13
LEVEL INSTRUMENT REFERENCE LEG AREA
DRYWELL TEMPERATURE CALCULATIONS
1.
For all
Level Instruments EXCEPT B21-LI-R605 A,
B,
(N027 A,
B); the
reference leg area drywell temperature is
the highest of the following
points:
Recorder
CAC-TR-4426-1B Point 1258-1
CAC-TR-4426-IB Point 1258-3
CAC-TR-4426-2B Point 1258-2
CAC-TR-4426-2B Point 1258-4
Microprocessor
CAC-TY-4426-1 Point 5801
CAC-TY-4426-1 Point 5803
CAC-TY-4426-2 Point 5802
CAC-TY-4426-2 Point 5804
I
Rev. 40
1
Page88of 139
Attachment 6
Page 7 of 19
ATTACHMENT
6 (Cont'd)
FIGURE 13 (Cont'd)
LEVEL INSTRUMENT REFERENCE LEG AREA
DRYWELL TEMPERATURE CALCULATIONS
2.
For Level Instruments B21-LI-R605A,
B (N027A,
B),
the reference leg area
drywell temperature is
the highest of the following points:
Recorder
CAC-TR-4426-1A Point 1258-22 __
CAC-TR-4426-IB Point 1258-3
CAC-TR-4426-2A Point 1258-23 __
CAC-TR-4426-2A Point 1258-24 __
CAC-TR-4426-2B Point 1258-2
CAC-TR-4426-2B Point 1258-4
Microprocessor
CAC-TY-4426-1 Point 5822 __
CAC-TY-4426-1 Point 5803 __
CAC-TY-4426-2 Point 5823 __
CAC-TY-4426-2 Point 5824 __
CAC-TY-4426-2 Point 5802
CAC-TY-4426-2 Point 5804
I OEOP-01-UG
I
Rev. 40
Page 89 of 139
Attachment 6
Page 8 of 19
ATTACHMENT 6 (Cont'd)
FIGURE 14
REACTOR SATURATION LIMIT
UNSAFE
300
Ile
t.
zjz~rtz 7K
tA-Vb--41-
600
550
500
450
400
350
300
250
200
500
4
400
SAFE
700
600
900
800
VI
'N1
1,100
1,000
REACTOR PRESSURE (PSIG)
I OEOP-01-UG
Rev. 40
Page 90 of 139
w
I-.
w
,L
w
-J
Li
a
a
wU
H
q1
100
0
ii
-I
200
1,200
- !!/-~1,1
..
. . ..
_
R
.4
i
Ill
Attachment 6
Page 9 of 19
ATTACHMENT 6 (Cont'd)
FIGURE 15
UNIT 1 NARROW RANGE LEVEL
INSTRUMENT (NO04A, B,
C)
CAUTION
350
400
450
REFERENCE LEG AREA DRYWELL TEMP (OF)
OEOP-01-UG
Rev. 40
Page 91 of 139
170
165
160
155
,-J
w
w
iw
z
150
300
Attachment 6
Page 10 of 19
ATTACHMENT 6 (Cont'd)
FIGURE 15A
UNIT 2 NARROW RANGE LEVEL
INSTRUMENT
(NO04A,
B,
C)
CAUTION
350
400
450
REFERENCE LEG AREA DRYWELL TEMP (OF)
OEOP-01-UG
Rev. 40
Page 92 of 139
170
2
w
,_.1
zj
z
165
160
155
150
300
Attachment 6
Page 11 of 19
ATTACHMENT
6 (Cont'd)
FIGURE 16
SHUTDOWN RANGE LEVEL
INSTRUMENT (N027A,
B)
CAUTION
REFERENCE LEG AREA DRYWELL TEMP (OF)
OEOP-01-UG
I
Rev. 40
1
Page 93 of 139
300
250
200
z
-J w
w
-J
w
z
150
Attachment 6
Page 12 of 19
ATTACHMENT 6 (Cont'd)
FIGURE 17
UNIT 1 REACTOR WATER LEVEL AT TAP
0
-10
-20
-30
-40
-50
4FHH
It -ABOVE
- -- ---- - -
-60
_
BELOW
- 70
TA F
- 80
-90
REF LEG
/TEMP
ABOVE
200F
REF LEG
TEMP
BELOW OR
EQUAL TO
200°F
- -IUU
100
iiiIII
300
60
200
l!IlL
I
.
.
.
500
700
900
1,100
400
600
800
1,000
1,150
REACTOR PRESSURE (PSIG)
NOTE
WHEN REACTOR PRESSURE IS LESS THAN
60 PSIG,
USE INDICATED LEVEL.
TAF IS -7.5 INCHES.
OEOP-01-UG
I
Rev. 40
Page 94 of 139
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= I
Attachment 6
Page 13 of 19
ATTACHMENT
6 (Cont'd)
FIGURE 17A
UNIT 2 REACTOR WATER LEVEL AT TAF
0
10
-20
-30
-40
-50
-60
-70
-80
-90
-100
60
200
400
600
800
1,000
REACTOR PRESSURE (PSIG)
NOTE
WHEN REACTOR PRESSURE IS LESS THAN
60 PSIG, USE INDICATED LEVEL.
TAF IS -7.5
INCHES.
OEOP-01 -UG
I
Rev. 40
1
Page 95 of 139
w
X
0.
z
-J
w
w
-Z
Attachment 6
Page 14 of 19
ATTACHMENT
6 (Cont'd)
FIGURE 18
UNIT 1 REACTOR WATER LEVEL AT LL-4
(MINIMUM STEAM COOLING LEVEL)
0
-10
-20
-30
-40
-50
-60
-70
-80
-90
-100
ABOVE
.. .. L L- 4
--
BELOW
LL-L4
R LI Ii
4~N$W...............I .I
k00
30
0111
300
500
7
700
9
900
I
1,150
1,100
60
200
400
600
800
1,000
REACTOR PRESSURE (PSIG)
NOTE
WHEN REACTOR PRESSURE IS LESS THAN
60 PSTG,
USE INDICATED LEVEL.
LL-4 IS
-32.5
INCHES.
Rev. 40
Page 96 of 139
L) z
-LJ
REF LEG
TEMP
ABOVE
2007F
SREF E
TEMP
BELOW OR
EQUAL TO
200F
Attachment 6
Page 15 of 19
ATTACHMENT 6 (Cont'd)
FIGURE 18A
UNIT 2 REACTOR WATER LEVEL AT LL-4
(MINIMUM STEAM COOLING LEVEL)
0
-10
-20
-30
-40
-50
-60
-70
-80
-90
-100
IREF
LEG
TEMP
ABOVE
200°F
REF LEG
TEMP
BELOW OR
EQUAL TO
200°F
-1,150
boo 1 300 I 500 I 700 1 900 11,100
60
200
400
600
800
1,000
REACTOR PRESSURE (PSIG)
NOTE
WHEN REACTOR PRESSURE
IS LESS THAN
60 PSIG,
USE INDICATED LEVEL.
LL-4 IS
-32.5 INCHES.
OEOP-01-UG
Rev. 40
Page 97 of 139
w
z
w
-J
Cl
aj
z
Attachment 6
Page 16 of 19
ATTACHMENT
6 (Cont'd)
FIGURE 19
UNIT 1 REACTOR WATER LEVEL AT LL-5
(MINIMUM ZERO INJECTION LEVEL)
0
-10
-20
-30
-40
-50
-60
-70
-80
-90
-100
60
200
400
600
800
1,000
REACTOR PRESSURE (PSIG)
NOTE
WHEN REACTOR PRESSURE IS LESS THAN
60 PSIG, USE INDICATED LEVEL.
LL-5 IS -47.5 INCHES.
I OEOP-o0-UG
I
Rev. 40
Page 98 of 139
w
z
w
Lu
z
Attachment 6
Page 17 of 19
ATTACHMENT 6 (Cont'd)
FIGURE 19A
UNIT 2 REACTOR WATER LEVEL AT LL-5
(MINIMUM ZERO INJECTION LEVEL)
TfhTh
IT1 H
II
I
ABOVE
LL-5
0
-10
-20
-30
-40
-50
-60
-70
-80
-90
-100
REFLEG
TEMP
ýABOVE
20WF
BELOW
qLL-5
.
I l l ll
300
500
700
900
II I L
1,100
- REF LEG
TEMP
BELOW OR
EQUAL TO
200*F
1,150
60
200
400
600
800
1,000
REACTOR PRESSURE (PSIG)
NOTE
WHEN REACTOR PRESSURE IS
LESS THAN
60 PSIG,
USE INDICATED LEVEL.
LL-5 IS
-47.5 INCHES.
I OEOP-o1-UG
I
Rev. 40
1
Page 99 of 139
w
r.
z
w
...I
w
I I-
z
hUll WhLLJI Wil WI Will WI WILLUI H
100
II
I
Attachment 6
Page 18 of 19
ATTACHMENT 6 (Cont'd)
FIGURE 20
REACTOR WATER LEVEL AT 90 INCHES
100
90
80
100
300
500
0
200
400
700
90(
600
800
3 11,10o
1,000
w
z
-J
w
w
-J
REACTOR PRESSURE (PSIG)
OEOP-01-UG
Rev. 40
Page 100 of 139
-_- ---- ---
A B O V E-
---
90 INCHES ----
BE.LOW
--
90 INCHES
70
60
50
40
30
20
10
0
-10
REF LEG
TEMP
ABOVE OR
EQUAL TO
200'F
REF LEG
TEMP
BELOW
200°F
-1,150
0
r[
l
Attachment 6
Page 19 of 19
ATTACHMENT
6 (Cont'd)
FIGURE 21
REACTOR WATER LEVEL AT MSL
(MAIN STEAM LINE FLOOD LEVEL)
300
REF LEG
/TEMP
ABOVE OR
EQUAL TO
200°F
REF LEG
TEMP
BELOW
200°F
1,150
60
200
400
600
800
1,000
REACTOR PRESSURE (PSIG)
NOTE
WHEN REACTOR PRESSURE IS
LESS THAN
60 PSIG,
USE INDICATED LEVEL.
MSL IS
+250 INCHES.
I
Rev. 40
Page 101 of 139
250
z
-j
w
'U
25
z
200