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| number = ML100550590 | | number = ML100550590 | ||
| issue date = 02/02/2010 | | issue date = 02/02/2010 | ||
| title = | | title = Developmental Revision B - Technical Specifications Bases B 3.6 - Containment Systems | ||
| author name = | | author name = | ||
| author affiliation = Tennessee Valley Authority | | author affiliation = Tennessee Valley Authority |
Revision as of 18:37, 30 January 2019
ML100550590 | |
Person / Time | |
---|---|
Site: | Watts Bar |
Issue date: | 02/02/2010 |
From: | Tennessee Valley Authority |
To: | Office of Nuclear Reactor Regulation |
References | |
Download: ML100550590 (6) | |
Text
Containment B 3.6.1 (continued)
Watts Bar - Unit 2 B 3.6-1 (developmental)
A B 3.6 CONTAINMENT SYSTEMS B 3.6.1 Containment
BASES BACKGROUND The containment is a free standing steel pressure vessel surrounded by a reinforced concrete shield building. The containment vessel, including all
its penetrations, is a low leakage steel shell designed to contain the
radioactive material that may be released from the reactor core following
a Design Basis Accident (DBA). Additionally, the containment and shield
building provide shielding from the fission products that may be present in
the containment atmosphere following accident conditions.
The containment vessel is a vertical cylindrical steel pressure vessel with
hemispherical dome and a concrete base mat with steel membrane. It is completely enclosed by a reinforced concrete shield building. An annular space exists between the walls and domes of the steel containment
vessel and the concrete shield building to provide for the collection, mixing, holdup, and controlled release of containment out leakage. Ice
condenser containments utilize an outer concrete building for shielding
and an inner steel containment for leak tightness.
Containment piping penetration assemblies provide for the passage of
process, service, sampling, and instrumentation pipelines into the
containment vessel while maintaining containment integrity. The shield
building provides shielding and allows controlled filtered release of the
annulus atmosphere under accident conditions, as well as environmental
missile protection for the containment vessel and Nuclear Steam Supply
System.
The inner steel containment and its penetrations establish the leakage
limiting boundary of the containment. Maintaining the containment
OPERABLE limits the leakage of fission product radioactivity from the
containment to the environment. SR 3.6.1.1 leakage rate requirements
comply with 10 CFR 50, Appendix J, Option B (Ref. 1), as modified by
approved exemptions.
Containment B 3.6.1 BASES (continued)
Watts Bar - Unit 2 B 3.6-2 (developmental)
A BACKGROUND (continued)
The isolation devices for the penetrations in the containment boundary are a part of the containment leak tight barrier. To maintain this leak tight
barrier:
- a. All penetrations required to be closed during accident conditions are either: 1. capable of being closed by an OPERABLE automatic containment isolation system, or
- 2. closed by manual valves, blind flanges, or de-activated automatic valves secured in their closed positions, except as
provided in LCO 3.6.3, "Containment Isolation Valves."
- c. All equipment hatches are closed.
APPLICABLE
SAFETY ANALYSES The safety design basis for the containment is that the containment must
withstand the pressures and temperatures of the limiting DBA without
exceeding the design leakage rates.
The DBAs that result in a challenge to containment OPERABILITY from
high pressures and temperatures are a loss of coolant accident (LOCA),
a steam line break (SLB), and a rod ejection accident (REA) (Ref. 2). In
addition, release of significant fission product radioactivity within
containment can occur from a LOCA or REA. In the DBA analyses, it is assumed that the containment is OPERABLE such that, for the DBAs involving release of fission product radioactivity, release to the
environment is controlled by the rate of containment leakage. The
containment was designed with an allowable leakage rate of 0.25% of
containment air weight per day (Ref. 3). This leakage rate, used in the
evaluation of offsite doses resulting from accidents, is defined in
10 CFR 50, Appendix J, Option B (Ref. 1), as L a: the maximum allowable containment leakage rate at the calculated peak containment internal
pressure (P a) related to the design basis LOCA. The allowable leakage rate represented by L a forms the basis for the acceptance criteria imposed on all containment leakage rate testing. L a is assumed to be 0.25% per day in the safety analysis at P a = 15.0 psig which bounds the calculated peak containment internal pressure resulting from the limiting
design basis LOCA (Ref. 3).
Containment B 3.6.1 BASES (continued)
Watts Bar - Unit 2 B 3.6-3 (developmental)
A APPLICABLE SAFETY ANALYSES (continued)
Satisfactory leakage rate test results are a requirement for the
establishment of containment OPERABILITY.
The containment satisfies Criterion 3 of the NRC Policy Statement.
LCO Containment OPERABILITY is maintained by limiting leakage to 1.0 L a , except prior to the first start up after performing a required Containment
Leakage Rate Testing Program leakage test. At this time, applicable
leakage limits must be met.
Compliance with this LCO will ensure a containment configuration, including equipment hatches, that is structurally sound and that will limit
leakage to those leakage rates assumed in the safety analysis.
Individual leakage rates specified for the containment air lock (LCO 3.6.2), purge valves with resilient seals, and Shield Building
containment bypass leakage (LCO 3.6.3) are not specifically part of the acceptance criteria of 10 CFR 50, Appendix J, Option B. Therefore, leakage rates exceeding these individual limits only result in the
containment being inoperable when the leakage results in exceeding the
acceptance criteria of Appendix J, Option B.
APPLICABILITY In MODES 1, 2, 3, and 4, a DBA could cause a release of radioactive material into containment. In MODES 5 and 6, the probability and
consequences of these events are reduced due to the pressure and
temperature limitations of these MODES. Therefore, containment is not
required to be OPERABLE in MODE 5 to prevent leakage of radioactive
material from containment. The requirements for containment during
MODE 6 are addressed in LCO 3.9.4, "Containment Penetrations."
Containment B 3.6.1 BASES (continued)
(continued)
Watts Bar - Unit 2 B 3.6-4 (developmental)
A ACTIONS A.1 In the event containment is inoperable, containment must be restored to
OPERABLE status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. The 1-hour Completion Time provides
a period of time to correct the problem commensurate with the
importance of maintaining containment OPERABLE during MODES 1, 2, 3, and 4. This time period also ensures that the probability of an accident (requiring containment OPERABILITY) occurring during periods when
containment is inoperable is minimal.
B.1 and B.2
If 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 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to MODE 5 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 SR 3.6.1.1
Maintaining the containment OPERABLE requires compliance with the
visual examinations and leakage rate test requirements of the
Containment Leakage Rate Testing Program. Failure to meet air lock, Shield Building containment bypass leakage path, and purge valve with
resilient seal leakage limits specified in LCO 3.6.2 and LCO 3.6.3 does
not invalidate the acceptability of these overall leakage determinations
unless their contribution to overall Type A, B, and C leakage causes that
to exceed limits. As left leakage prior to the first startup after performing
a required leakage test is required to be < 0.6 L a for combined Type B and C leakage and 0.75 L a for overall Type A leakage. At all other times between required leakage rate tests, the acceptance criteria is
based on an overall Type A leakage limit of 1.0 L a. At 1.0 L a the offsite dose consequences are bounded by the assumptions of the safety
analysis.
SR Frequencies are as required by the Containment Leakage Rate
Testing Program. These periodic testing requirements verify that the
containment leakage rate does not exceed the leakage rate assumed in
the safety analysis.
Containment B 3.6.1 BASES (continued)
Watts Bar - Unit 2 B 3.6-5 (developmental)
A REFERENCES
- 1. Title 10, Code of Federal Regulations, Part 50, Appendix J, Option B, "Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors - Performance-Based
Requirements." 2. Watts Bar FSAR, Section 15.0, "Accident Analysis." 3. Watts Bar FSAR, Section 6.2, "Containment Systems." 4. Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program," September 1995.
Containment Air Locks B 3.6.2 (continued)
Watts Bar - Unit 2 B 3.6-6 (developmental)
A B 3.6 CONTAINMENT SYSTEMS B 3.6.2 Containment Air Locks
BASES BACKGROUND Containment air locks form part of the containment pressure boundary and provide a means for personnel access during all MODES of
operation.
Each air lock is nominally a right circular cylinder, 8 ft 7 inches in
diameter, with a door at each end. The doors are interlocked to prevent
simultaneous opening. During periods when containment is not required
to be OPERABLE, the door interlock mechanism may be disabled, allowing both doors of an air lock to remain open for extended periods
when frequent containment entry is necessary. Each air lock door has been designed and tested to certify its ability to withstand a pressure in excess of the maximum expected pressure following a Design Basis
Accident (DBA) in containment. As such, closure of a single door
supports containment OPERABILITY. Each of the doors contains double
gasketed seals and local leakage rate testing capability to ensure
pressure integrity. To effect a leak tight seal, the air lock design uses
pressure seated doors (i.e., an increase in containment internal pressure
results in increased sealing force on each door).
Each personnel air lock is provided with limit switches on both doors that
provide control room indication of door position.
The containment air locks form part of the containment pressure
boundary. As such, air lock integrity and leak tightness is essential for
maintaining the containment leakage rate within limit in the event of a
DBA. Not maintaining air lock integrity or leak tightness may result in a
leakage rate in excess of that assumed in the plant safety analyses.
Containment Air Locks B 3.6.2 BASES (continued)
(continued)
Watts Bar - Unit 2 B 3.6-7 (developmental)
A APPLICABLE SAFETY ANALYSES The DBAs that result in a significant release of radioactive material within
containment are a loss of coolant accident and a rod ejection accident (Ref. 2). In the analysis of each of these accidents, it is assumed that
containment is OPERABLE such that release of fission products to the
environment is controlled by the rate of containment leakage. The
containment was designed with an allowable leakage rate (L a) of 0.25%
of containment air weight per day (Ref. 2), at the calculated peak
containment pressure of 15.0 psig. This allowable leakage rate forms the
basis for the acceptance criteria imposed on the SRs associated with the
air locks.
The containment air locks satisfy Criterion 3 of the NRC Policy
Statement.
LCO Each containment air lock forms part of the containment pressure boundary. As part of containment pressure boundary, the air lock safety
function is related to control of the containment leakage rate resulting
from a DBA. Thus, each air lock's structural integrity and leak tightness
are essential to the successful mitigation of such an event.
Each air lock is required to be OPERABLE. For the air lock to be
considered OPERABLE, the air lock interlock mechanism must be
OPERABLE, the air lock must be in compliance with the Type B air lock
leakage test, and both air lock doors must be OPERABLE. The interlock allows only one air lock door of an air lock to be opened at one time. This provision ensures that a gross breach of containment does not exist when
containment is required to be OPERABLE. Closure of a single door in
each air lock is sufficient to provide a leak tight barrier following
postulated events. Nevertheless, both doors are kept closed when the air
lock is not being used for normal entry into and exit from containment.
APPLICABILITY In MODES 1, 2, 3, and 4, a DBA could cause a release of radioactive material to containment. In MODES 5 and 6, the probability and
consequences of these events are reduced due to the pressure and
temperature limitations of these MODES. Therefore, the containment air
locks are not required in MODE 5 to prevent leakage of radioactive
material from containment. The requirements for the containment air
locks during MODE 6 are addressed in LCO 3.9.4, "Containment
Containment Air Locks B 3.6.2 BASES (continued)
(continued)
Watts Bar - Unit 2 B 3.6-8 (developmental)
A ACTIONS The ACTIONS are modified by a Note that allows entry and exit to perform repairs on the affected air lock component. If the outer door is inoperable, then it may be easily accessed for most repairs. It is
preferred that the air lock be accessed from inside containment by
entering through the other OPERABLE air lock. However, if this is not
practicable, or if repairs on either door must be performed from the barrel
side of the door, then it is permissible to enter the air lock through the
OPERABLE door which means there is a short time during which the
containment boundary is not intact (during access through the
OPERABLE door). The ability to open the OPERABLE door, even if it
means the containment boundary is temporarily not intact, is acceptable
due to the low probability of an event that could pressurize the
containment during the short time in which the OPERABLE door is
expected to be open. After each entry and exit, the OPERABLE door
must be immediately closed.
A second Note has been added to provide clarification that, for this LCO, separate Condition entry is allowed for each air lock. This is acceptable, since the Required Actions for each Condition provide appropriate
compensatory actions for each inoperable air lock. Complying with the
Required Actions may allow for continued operation, and a subsequent
inoperable air lock is governed by subsequent Condition entry and
application of associated Required Actions.
In the event the air lock leakage results in exceeding the overall
containment leakage rate, Note 3 directs entry into the applicable
Conditions and Required Actions of LCO 3.6.1, "Containment."
A.1, A.2, and A.3
With one air lock door in one or more containment air locks inoperable, the OPERABLE door must be verified closed (Required Action A.1) in
each affected containment air lock. This ensures that a leak tight
containment barrier is maintained by the use of an OPERABLE air lock
door. This action must be completed within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. This specified time
period is consistent with the ACTIONS of LCO 3.6.1, which requires
containment be restored to OPERABLE status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
In addition, the affected air lock penetration must be isolated by locking closed the OPERABLE air lock door within the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Completion Time.
The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Completion Time is reasonable for locking the OPERABLE
air lock door, considering the OPERABLE door of the affected air lock is
being maintained closed.
Containment Air Locks B 3.6.2 BASES (continued)
Watts Bar - Unit 2 B 3.6-9 (developmental)
A ACTIONS A.1, A.2, and A.3 (continued)
Required Action A.3 verifies that an air lock with an inoperable door has
been isolated by the use of a locked and closed OPERABLE air lock
door. This ensures that an acceptable containment leakage boundary is
maintained. The Completion Time of once per 31 days is based on
engineering judgment and is considered adequate in view of the low
likelihood of a locked door being mispositioned and other administrative
controls. Required Action A.3 is modified by a Note that applies to air
lock doors located in high radiation areas and allows these doors to be
verified locked closed by use of administrative means. Allowing verification by administrative means is considered acceptable, since access to these areas is typically restricted. Therefore, the probability of
misalignment of the door, once it has been verified to be in the proper
position, is small.
The Required Actions have been modified by two Notes. Note 1 ensures
that only the Required Actions and associated Completion Times of
Condition C are required if both doors in the same air lock are inoperable.
With both doors in the same air lock inoperable, an OPERABLE door is
not available to be closed. Required Actions C.1 and C.2 are the
appropriate remedial actions. The exception of Note 1 does not affect
tracking the Completion Time from the initial entry into Condition A; only
the requirement to comply with the Required Actions. Note 2 allows use
of the air lock for entry and exit for 7 days under administrative controls if both air locks have an inoperable door.
This 7 day restriction begins when the second air lock is discovered
inoperable. Containment entry may be required on a periodic basis to
perform Technical Specifications (TS) Surveillances and Required
Actions, as well as other activities on equipment inside containment that
are required by TS or activities on equipment that support TS-required
equipment. This Note is not intended to preclude performing other activities (i.e., non-TS-required activities) if the containment is entered, using the inoperable air lock, to perform an allowed activity listed above.
This allowance is acceptable due to the low probability of an event that
could pressurize the containment during the short time that the
OPERABLE door is expected to be open.
Containment Air Locks B 3.6.2 BASES (continued)
Watts Bar - Unit 2 B 3.6-10 (developmental)
A ACTIONS (continued)
B.1, B.2, and B.3 With an air lock interlock mechanism inoperable in one or more air locks, the Required Actions and associated Completion Times are consistent
with those specified in Condition A.
The Required Actions have been modified by two Notes. Note 1 ensures
that only the Required Actions and associated Completion Times of
Condition C are required if both doors in the same air lock are inoperable.
With both doors in the same air lock inoperable, an OPERABLE door is
not available to be closed. Required Actions C.1 and C.2 are the appropriate remedial actions. Note 2 allows entry into and exit from containment under the control of a dedicated individual stationed at the
air lock to ensure that only one door is opened at a time (i.e., the
individual performs the function of the interlock).
Required Action B.3 is modified by a Note that applies to air lock doors
located in high radiation areas and allows these doors to be verified
locked closed by use of administrativ e means. Allowing verification by administrative means is considered acceptable, since access to these
areas is typically restricted. Therefore, the probability of misalignment of
the door, once it has been verified to be in the proper position, is small.
C.1, C.2, and C.3
With one or more air locks inoperable for reasons other than those
described in Condition A or B, Required Action C.1 requires action to be
initiated immediately to evaluate previous combined leakage rates using
current air lock test results. An evaluation is acceptable, since it is overly
conservative to immediately declare the containment inoperable if both
doors in an air lock have failed a seal test or if the overall air lock leakage
is not within limits. In many instances (e.g., only one seal per door has
failed), containment remains OPERABLE, yet only 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> (per LCO 3.6.1)
would be provided to restore the air lock door to OPERABLE status prior to requiring a plant shutdown. In addition, even with both doors failing the seal test, the overall containment leakage rate can still be within limits.
Required Action C.2 requires that one door in the affected containment
air lock must be verified to be closed within the 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Completion Time.
This specified time period is consistent with the ACTIONS of LCO 3.6.1, which requires that containment be restored to OPERABLE status within
1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
Containment Air Locks B 3.6.2 BASES (continued)
Watts Bar - Unit 2 B 3.6-11 (developmental)
A ACTIONS C.1, C.2, and C.3 (continued)
Additionally, the affected air lock must be restored to OPERABLE status
within the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Completion Time. The specified time period is
considered reasonable for restoring an inoperable air lock to OPERABLE
status, assuming that at least one door is maintained closed in each
affected air lock.
D.1 and D.2
If the inoperable containment air lock 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 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to MODE 5
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 SR 3.6.2.1
Maintaining containment air locks OPERABLE requires compliance with
the leakage rate test requirements of the Containment Leakage Rate
Testing Program. This SR reflects the leakage rate testing requirements
with regard to air lock leakage (Type B leakage tests). The acceptance
criteria were established during initial air lock and containment
OPERABILITY testing. The periodic testing requirements verify that the
air lock leakage does not exceed the allowed fraction of the overall
containment leakage rate. The Frequency is required by the
Containment Leakage Rate Testing Program.
The SR has been modified by two Notes. Note 1 states that an
inoperable air lock door does not invalidate the previous successful
performance of the overall air lock leakage test. This is considered
reasonable since either air lock door is capable of providing a fission
product barrier in the event of a DBA. Note 2 requires the results of the
air lock leakage tests to be evaluated against the acceptance criteria of
the Containment Leakage Rate Testing Program, 5.7.2.19. This ensures
that air lock leakage is properly accounted for in determining the
combined Type B and C containment leakage rate.
Containment Air Locks B 3.6.2 BASES Watts Bar - Unit 2 B 3.6-12 (developmental)
A SURVEILLANCE REQUIREMENTS (continued)
The air lock interlock is designed to prevent simultaneous opening of both
doors in a single air lock. Since both the inner and outer doors of an air
lock are designed to withstand the maximum expected post accident
containment pressure, closure of either door will support containment
OPERABILITY. Thus, the door interlock feature supports containment
OPERABILITY while the air lock is being used for personnel transit in and
out of the containment. Periodic testing of this interlock demonstrates
that the interlock will function as designed and that simultaneous opening
of the inner and outer doors will not inadvertently occur.
Due to the purely mechanical nature of this interlock, and given that the
interlock mechanism is only challenged when the containment air lock
door is opened, this test is only required to be performed upon entering or
exiting a containment air lock but is not required more frequently than
every 184 days. The 184 day Frequency is based on engineering
judgment and is considered adequate in view of other indications of door
status available to operations personnel and because the interlock is only
disabled in MODES 5 and 6.
REFERENCES
- 1. Title 10, Code of Federal Regulations, Part 50, Appendix J, Option B, "Primary Reactor Containment Leakage Testing for
Water-Cooled Power Reactors - Performance-Based
Requirements." 2. Watts Bar FSAR, Section 15.0, "Accident Analysis."
Containment Isolation Valves B 3.6.3 (continued)
Watts Bar - Unit 2 B 3.6-13 (developmental)
A B 3.6 CONTAINMENT SYSTEMS B 3.6.3 Containment Isolation Valves
BASES BACKGROUND The containment isolation valves form part of the containment pressure boundary and provide a means for fluid penetrations not serving accident
consequence limiting systems to be provided with two isolation barriers
that are closed on a containment isolation signal or which are normally
closed. These isolation devices are either passive or active (automatic).
Manual valves, de-activated automatic valves secured in their closed
position (including check valves with flow through the valve secured),
blind flanges, and closed systems are considered passive devices. Check
valves, or other automatic valves designed to close without operator
action following an accident, are considered active devices. Two barriers
in series are provided for each penetration so that no single credible
failure or malfunction of an active component can result in a loss of
isolation or leakage that exceeds limits assumed in the safety analyses.
One of these barriers may be a closed system. These barriers (typically
containment isolation valves) make up the Containment Isolation System.
Automatic isolation signals are produced during accident conditions.
Containment Phase "A" isolation occurs upon receipt of a safety injection
signal. The Phase "A" isolation signal isolates non-essential process
lines in order to minimize leakage of fission product radioactivity.
Containment Phase "B" isolation occurs upon receipt of a containment
pressure - High High signal and isolates the remaining process lines, except systems required for accident mitigation. In addition to the
isolation signals listed above, the purge and exhaust valves receive an
isolation signal on a containment high radiation condition. As a result, the
containment isolation valves (and blind flanges) help ensure that the
containment atmosphere will be isolated from the environment in the
event of a release of fission product radioactivity to the containment
atmosphere as a result of a Design Basis Accident (DBA).
The OPERABILITY requirements for containment isolation valves help
ensure that containment is isolated within the time limits assumed in the
safety analyses. Therefore, the OPERABILITY requirements provide
assurance that the containment function assumed in the safety analyses
will be maintained.
Containment Isolation Valves B 3.6.3 BASES (continued)
Watts Bar - Unit 2 B 3.6-14 (developmental)
A BACKGROUND (continued)
Reactor Building Purge Ventilation System The Reactor Building Purge Ventilation system operates to supply outside
air into the containment for ventilation and cooling or heating, to equalize
internal and external pressures and to reduce the concentration of noble
gases within containment prior to and during personnel access. The
supply and exhaust lines each contain two isolation valves. Because of
their large size and their exposure to higher containment pressure during
accident conditions, the 24 inch containment lower compartment purge
isolation valves are physically restricted to 50 degrees open.
Since the valves used in the Reactor Building Purge Ventilation System
are designed to meet the requirements for automatic containment
isolation valves, these valves may be opened as needed in MODES 1, 2, 3 and 4.
APPLICABLE
SAFETY ANALYSES The containment isolation valve LCO was derived from the assumptions
related to minimizing the loss of reactor coolant inventory and
establishing the containment boundary during major accidents. As part of
the containment boundary, containment isolation valve OPERABILITY
supports leak tightness of the containment. Therefore, the safety
analyses of any event requiring isolation of containment is applicable to
this LCO.
The DBAs that result in a significant release of radioactive material within
containment are a loss of coolant accident (LOCA) and a rod ejection
accident (Ref. 1). In the analyses for each of these accidents, it is
assumed that containment isolation valves are either closed or function to
close within the required isolation time following event initiation. This
ensures that potential paths to the environment through containment
isolation valves (including containment purge valves) are minimized.
The DBA analysis assumes that, within 60 seconds after the accident, isolation of the containment is complete and leakage terminated except
for the design leakage rate (L a) and for valves in the Essential Raw Cooling Water (ERCW) system and Co mponent Cooling System (CSS).
These valves are in liquid cont aining systems and have been evaluated
to have no impact on the DBA analysis. The containment isolation total
response time of 60 seconds includes signal delay, diesel generator
startup (for loss of offsite power), and containment isolation valve stroke
times.
Containment Isolation Valves B 3.6.3 BASES (continued)
Watts Bar - Unit 2 B 3.6-15 (developmental)
A APPLICABLE SAFETY ANALYSES (continued)
The single failure criterion required to be imposed in the conduct of plant
safety analyses was considered in the original design of the containment
purge valves. Two valves in series on each purge line provide assurance
that both the supply and exhaust lines could be isolated even if a single
failure occurred. The inboard and outboard isolation valves on each line
are provided with redundant control and power trains, pneumatically
operated to open, and spring-loaded to close upon power loss or air
failure. This arrangement was designed to preclude common mode
failures from disabling both valves on a purge line.
The containment isolation valves satisfy Criterion 3 of the NRC Policy
Statement.
LCO Containment isolation valves form a part of the containment boundary.
The containment isolation valves' safety function is related to minimizing
the loss of reactor coolant inventory and establishing the containment
boundary during a DBA.
The automatic power operated isolation valves are required to have
isolation times within limits and to actuate on an automatic isolation
signal. The 24 inch containment lower compartment purge valves must
have blocks installed to prevent full opening. Blocked purge valves also
actuate on an automatic signal. The valves covered by this LCO are
listed along with their associated stroke times in the FSAR (Ref. 2).
The normally closed containment isolation valves are considered
OPERABLE when manual valves are closed, automatic valves are
de-activated and secured in their closed position, blind flanges are in
place, and closed systems are intact. These passive isolation
valves/devices are those listed in Reference 2.
Purge valves with resilient seals and shield building bypass valves meet
additional leakage rate requirements. The other containment isolation
valve leakage rates are addressed by LCO 3.6.1, "Containment," as
Type C testing.
This LCO provides assurance that the containment isolation valves will
perform their designed safety functions to minimize the loss of reactor
coolant inventory and establish the containment boundary during
accidents.
Containment Isolation Valves B 3.6.3 BASES (continued)
(continued)
Watts Bar - Unit 2 B 3.6-16 (developmental)
A APPLICABILITY In MODES 1, 2, 3, and 4, a DBA could cause a release of radioactive material to containment. In MODES 5 and 6, the probability and consequences of these events are reduced due to the pressure and
temperature limitations of these MODES. Therefore, the containment
isolation valves are not required to be OPERABLE in MODE 5. The
requirements for containment isolation valves during MODE 6 are
addressed in LCO 3.9.4, "Containment Penetrations."
ACTIONS The ACTIONS are modified by a Note allowing penetration flow paths, to be unisolated intermittently under administrative controls. These
administrative controls consist of stationing a dedicated operator (licensed or unlicensed) at the valve controls, who is in continuous
communication with the control room. In this way, the penetration can be
rapidly isolated when a need for containment isolation is indicated. For
valve controls located in the control room, an operator (other than the
Shift Operations Supervisor (SOS), ASOS, or the Operator at the
Controls) may monitor containment isolation signal status rather than be
stationed at the valve controls. Other secondary responsibilities which do
not prevent adequate monitoring of containment isolation signal status
may be performed by the operator provided his/her primary responsibility
is rapid isolation of the penetration when needed for containment
isolation. Use of the Unit Control Room Operator (CRO) to perform this
function should be limited to those situations where no other operator is
available.
A second Note has been added to provide clarification that, for this LCO, separate Condition entry is allowed for each penetration flow path. This
is acceptable, since the Required Actions for each Condition provide
appropriate compensatory actions for each inoperable containment
isolation valve. Complying with the Required Actions may allow for
continued operation, and subsequent inoperable containment isolation
valves are governed by subsequent Condition entry and application of
associated Required Actions.
The ACTIONS are further modified by third Note, which ensures
appropriate remedial actions are taken, if necessary, if the affected
systems are rendered inoperable by an inoperable containment isolation
valve.
In the event the isolation valve leakage results in exceeding the overall
containment leakage rate, Note 4 directs entry into the applicable
Conditions and Required Actions of LCO 3.6.1.
Containment Isolation Valves B 3.6.3 BASES (continued)
Watts Bar - Unit 2 B 3.6-17 (developmental)
B ACTIONS (continued)
A.1 and A.2 In the event one containment isolation valve in one or more penetration
flow paths is inoperable except for purge valve or shield building bypass
leakage not within limit, the affected penetration flow path must be
isolated. The method of isolation must include the use of at least one
isolation barrier that cannot be adversely affected by a single active
failure. Isolation barriers that meet this criterion are a closed and
de-activated automatic containment isolation valve, a closed manual
valve, a blind flange, and a check valve with flow through the valve
secured. For a penetration flow path isolated in accordance with
Required Action A.1, the device used to isolate the penetration should be
the closest available one to containment. Required Action A.1 must be
completed within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Completion Time is reasonable, considering the time required to isolate the penetration and the relative
importance of supporting containment OPERABILITY during MODES 1, 2, 3, and 4.
For affected penetration flow paths that cannot be restored to
OPERABLE status within the 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Completion Time and that have been
isolated in accordance with Required Action A.1, the affected penetration
flow paths must be verified to be isolated on a periodic basis. This is
necessary to ensure that containment penetrations required to be isolated
following an accident and no longer capable of being automatically
isolated will be in the isolation position should an event occur. This
Required Action does not require any testing or device manipulation.
Rather, it involves verification that those isolation devices outside containment and capable of being mispositioned are in the correct
position. The Completion Time of "Once per 31 days for isolation devices
outside containment" is appropriate considering the fact that the devices
are operated under administrative controls and the probability of their
misalignment is low. For the isolation devices inside containment, the
time period specified as "Prior to entering MODE 4 from MODE 5 if not
performed within the previous 92 days" is based on engineering judgment
and is considered reasonable in view of the inaccessibility of the isolation
devices and other administrative controls that will ensure that isolation
device misalignment is an unlikely possibility.
Condition A has been modified by a Note indicating that this Condition is
only applicable to those penetration flow paths with two containment
isolation valves. For penetration flow paths with only one containment
isolation valve and a closed system, Condition C provides the appropriate
actions.
Containment Isolation Valves B 3.6.3 BASES (continued)
Watts Bar - Unit 2 B 3.6-18 (developmental)
B ACTIONS A.1 and A.2 (continued)
Required Action A.2 is modified by two Notes. Note 1 applies to isolation devices located in high radiation areas and allows these devices to be
verified closed by use of administrative means. Allowing verification by
administrative means is considered acceptable, since access to these
areas is typically restricted. Note 2 applies to isolation devices that are locked, sealed, or otherwise secured in position and allows these devices to be verified closed by use of administrative means. Allowing verification by administrative means is considered acceptable, since the function of locking, sealing, or securing components is to ensure that these devices are not inadvertently repositioned. Therefore, the probability of misalignment of these devices, once they have been verified to be in the
proper position, is small.
B.1 With two containment isolation valves in one or more penetration flow
paths inoperable, the affected penetration flow path must be isolated
within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. The method of isolation must include the use of at least
one isolation barrier that cannot be adversely affected by a single active
failure. Isolation barriers that meet this criterion are a closed and
de-activated automatic valve, a closed manual valve, and a blind flange.
The 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Completion Time is consistent with the ACTIONS of
LCO 3.6.1. In the event the affected penetration is isolated in accordance
with Required Action B.1, the affected penetration must be verified to be
isolated on a periodic basis per Required Action A.2, which remains in
effect. This periodic verification is necessary to assure leak tightness of
containment and that penetrations requiring isolation following an
accident are isolated. The Completion Time of once per 31 days for
verifying each affected penetration flow path is isolated is appropriate
considering the fact that the valves are operated under administrative
control and the probability of their misalignment is low. Condition B is
modified by a Note indicating this Condition is only applicable to
penetration flow paths with two containment isolation valves. Condition A
of this LCO addresses the condition of one containment isolation valve
inoperable in this type of penetration flow path.
Containment Isolation Valves B 3.6.3 BASES (continued)
Watts Bar - Unit 2 B 3.6-19 (developmental)
B ACTIONS (continued)
C.1 and C.2 With one or more penetration flow paths with one containment isolation
valve inoperable, the inoperable valve flow path must be restored to
OPERABLE status or the affected penetration flow path must be isolated.
The method of isolation must include the use of at least one isolation
barrier that cannot be adversely affected by a single active failure.
Isolation barriers that meet this criterion are a closed and de-activated
automatic valve, a closed manual valve, and a blind flange. A check
valve may not be used to isolate the affected penetration flow path.
Required Action C.1 must be completed within the 4-hour Completion
Time. The specified time period is reasonable considering the relative
stability of the closed system (hence, reliability) to act as a penetration
isolation boundary and the relative importance of maintaining
containment integrity during MODES 1, 2, 3, and 4. In the event the
affected penetration flow path is isolated in accordance with Required
Action C.1, the affected penetration flow path must be verified to be
isolated on a periodic basis. This periodic verification is necessary to
assure leak tightness of containment and that containment penetrations
requiring isolation following an accident are isolated. The Completion
Time of once per 31 days for verifying that each affected penetration flow
path is isolated is appropriate because the valves are operated under
administrative controls and the probability of their misalignment is low.
Condition C is modified by a Note indicating that this Condition is only
applicable to those penetration flow paths with only one containment
isolation valve and a closed system. This Note is necessary since this
Condition is written to specifically address those penetration flow paths in
a closed system. Required Action C.2 is modified by two Notes. Note 1 applies to valves and blind flanges located in high radiation areas and
allows these devices to be verified closed by use of administrative
means. Allowing verification by adm inistrative means is considered acceptable, since access to these areas is typically restricted. Note 2 applies to isolation devices that are locked, sealed, or otherwise secured in position and allows these devices to be verified closed by use of administrative means. Allowing verifi cation by administrative means is considered acceptable, since the function of locking, sealing, or securing components is to ensure that these devices are not inadvertently repositioned. Therefore, the probability of misalignment of these valves, once they have been verified to be in the proper position, is small.
Containment Isolation Valves B 3.6.3 BASES (continued)
Watts Bar - Unit 2 B 3.6-20 (developmental)
B ACTIONS D.1 With the shield building bypass leakage rate not within limit, the
assumptions of the safety analyses are not met. Therefore, the leakage
must be restored to within limit within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. Restoration can be
accomplished by isolating the penetration(s) that caused the limit to be
exceeded by use of one closed and de-activated automatic valve, closed
manual valve, or blind flange. When a penetration is isolated the leakage
rate for the isolated penetration is assumed to be the actual pathway
leakage through the isolation device. If two isolation devices are used to
isolate the penetration, the leakage rate is assumed to be the lesser
actual pathway leakage of the two devices. The 4-hour Completion Time
is reasonable considering the time required to restore the leakage by
isolating the penetration(s) and the relative importance of shield building
bypass leakage to the overall containment function.
E.1, E.2, and E.3
In the event one or more containment purge valves in one or more
penetration flow paths are not within the purge valve leakage limits, purge
valve leakage must be restored to within limits, or the affected penetration
flow path must be isolated. The method of isolation must be by the use of
at least one isolation barrier that cannot be adversely affected by a single
active failure. Isolation barriers that meet this criterion are a closed and
de-activated automatic valve, closed manual valve, or blind flange. A
purge valve with resilient seals utilized to satisfy Required Action E.1
must have been demonstrated to meet the leakage requirements of
SR 3.6.3.5. The specified Completion Time is reasonable, considering
that one containment purge valve remains closed so that a gross breach
of containment does not exist.
In accordance with Required Action E.2, this penetration flow path must
be verified to be isolated on a periodic basis. The periodic verification is
necessary to ensure that containment penetrations required to be isolated
following an accident, which are no longer capable of being automatically
isolated, will be in the isolation position should an event occur. This
Required Action does not require any testing or valve manipulation.
Rather, it involves verification that those isolation devices outside containment potentially capable of being mispositioned are in the correct
position. For the isolation devices inside containment, the time period
specified as "Prior to entering MODE 4 from MODE 5 if not performed
within the previous Containment Isolation Valves B 3.6.3 BASES (continued)
Watts Bar - Unit 2 B 3.6-21 (developmental)
B ACTIONS E.1, E.2, and E.3 (continued) 92 days" is based on engineering judgment and is considered reasonable
in view of the inaccessibility of the isolation devices and other
administrative controls that will ensure that isolation device misalignment
is an unlikely possibility.
For the containment purge valve with resilient seal that is isolated in
accordance with Required Action E.1, SR 3.6.3.5 must be performed at
least once every 92 days. This assures that degradation of the resilient
seal is detected and confirms that the leakage rate of the containment
purge valve does not increase during the time the penetration is isolated.
The normal Frequency for SR 3.6.3.5, 184 days, is based on an NRC
initiative, Generic Issue B-20 (Ref. 3). Since more reliance is placed on a
single valve while in this Condition, it is prudent to perform the SR more
often. Therefore, a Frequency of once per 92 days was chosen and has
been shown to be acceptable based on operating experience.
Required Action E.2 is modified by two Notes. Note 1 applies to isolation devices located in high radiation areas and allows these devices to be verified closed by use of administrative means. Allowing verification by administrative means is considered acceptable, since access to these areas is typically restricted. Note 2 applies to isolation devices that are locked, sealed, or otherwise secured in position and allows these devices to be verified closed by use of administrative means. Allowing verification by administrative means is considered acceptable, since the function of locking, sealing, or securing components is to ensure that these devices are not inadvertently repositioned.
F.1 and F.2
If the Required Actions and associated Completion Times are not 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 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to MODE 5 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.
Containment Isolation Valves B 3.6.3 BASES (continued)
Watts Bar - Unit 2 B 3.6-22 (developmental)
B SURVEILLANCE REQUIREMENTS SR 3.6.3.1
This SR ensures that the purge valves are closed as required or, if open, open for an allowable reason. If a purge valve is open in violation of this
SR, the valve is considered inoperable. If the inoperable valve is not
otherwise known to have excessive leakage when closed, it is not
considered to have leakage outside of limits. The SR is not required to
be met when the purge valves are open for the reasons stated. The
valves may be opened for pressure control, ALARA or air quality
considerations for personnel entry, or for Surveillances that require the
valves to be open. All purge valves are capable of closing in the
environment following a LOCA. Therefore, these valves are allowed to be
open for limited periods of time. The 31-day Frequency is consistent with
other containment isolation valve requirements discussed in SR 3.6.3.2.
This SR requires verification that each containment isolation manual
valve and blind flange located outside containment, the containment
annulus, and the Main Steam Valve Vault Rooms, and not locked, sealed, or otherwise secured and required to be closed during accident conditions is closed. The SR helps to ensure that post accident leakage of
radioactive fluids or gases outside of the containment boundary is within
design limits. This SR does not require any testing or valve manipulation.
Rather, it involves verification that those containment isolation valves in areas where the valves are capable of being mispositioned are in the
correct position. Since verification of valve position for these valves is
relatively easy, the 31 day Frequency is based on engineering judgment
and was chosen to provide added assurance of the correct positions.
The SR specifies that containment isolation valves that are open under
administrative controls are not required to meet the SR during the time
the valves are open. This SR does not apply to valves that are locked, sealed, or otherwise secured in the closed position, since these were verified to be in the correct position upon locking, sealing, or securing.
The Note applies to valves and blind flanges located in high radiation
areas and allows these devices to be verified closed by use of
administrative means. Allowing verifi cation by administrative means is considered acceptable, since access to these areas is typically restricted
for ALARA reasons. Therefore, the probability of misalignment of these
containment isolation valves, once they have been verified to be in the
proper position, is small.
Containment Isolation Valves B 3.6.3 BASES (continued)
Watts Bar - Unit 2 B 3.6-23 (developmental)
B SURVEILLANCE REQUIREMENTS SR 3.6.3.3
This SR requires verification that each containment isolation manual
valve and blind flange located inside containment, the containment
annulus, and the Main Steam Valve Vault Rooms, and not locked, sealed, or otherwise secured and required to be closed during accident conditions is closed. The SR helps to ensure that post accident leakage of
radioactive fluids or gases outside of the containment boundary is within
design limits. For these containment isolation valves, the Frequency of "Prior to entering MODE 4 from MODE 5 if not performed within the
previous 92 days" is appropriate since these containment isolation valves
are operated under administrative controls (e.g., locked valve program)
and may be verified by administrative means, because the probability of
their misalignment is low. The SR specifies that containment isolation
valves that are open under administrative controls are not required to
meet the SR during the time they are open. This SR does not apply to valves that are locked, sealed, or otherwise secured in the closed position, since these were verified to be in the correct position upon locking, sealing, or securing.
The Note allows valves and blind flanges located in high radiation areas
to be verified closed by use of administrative means. Allowing verification
by administrative means is considered acceptable, since access to these
areas is typically restricted for ALARA reasons. Therefore, the probability
of misalignment of these containment isolation valves, once they have
been verified to be in their proper position, is small.
Verifying that the isolation time of each power operated and automatic
containment isolation valve is within limits is required to demonstrate
OPERABILITY. The isolation time test ensures the valve will isolate in a
time period less than or equal to that assumed in the safety analyses.
The isolation time and Frequency of this SR are in accordance with the
Inservice Testing Program or 92 days.
Containment Isolation Valves B 3.6.3 BASES (continued)
Watts Bar - Unit 2 B 3.6-24 (developmental)
A SURVEILLANCE REQUIREMENTS (continued)
For containment purge valves with resilient seals, additional leakage rate
testing beyond the test requirements of 10 CFR 50, Appendix J, Option B (Ref. 4), is required to ensure OPERABILITY.
Operating experience has demonstrated that this type of seal has the
potential to degrade in a shorter time period than do other seal types.
Based on this observation and the importance of maintaining this
penetration leak tight (due to the direct path between containment and the
environment), a Frequency of 184 days was established as part of the
NRC resolution of Generic Issue B-20, "Containment Leakage Due to
Seal Deterioration" (Ref. 3).
Additionally, this SR must be performed within 92 days after opening the
valve. The 92-day Frequency was chosen recognizing that cycling the
valve could introduce additional seal degradation (beyond that occurring
to a valve that has not been opened). Thus, decreasing the interval (from
184 days) is a prudent measure after a valve has been opened.
SR 3.6.3.6 Automatic containment isolation valves close on a containment isolation
signal to prevent leakage of radioactive material from containment
following a DBA. This SR ensures that each automatic containment
isolation valve will actuate to its isolation position on a containment
isolation signal. This Surveillance is not required for valves that are
locked, sealed, or otherwise secured in the required position under
administrative control. The 18-month Frequency is based on the 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.
Operating experience has shown that these components usually pass this
Surveillance when performed at the 18-month Frequency. Therefore, the
Frequency was concluded to be acceptable from a reliability standpoint.
Containment Isolation Valves B 3.6.3 BASES Watts Bar - Unit 2 B 3.6-25 (developmental)
A SURVEILLANCE REQUIREMENTS SR 3.6.3.7 Verifying that each 24 inch containment lower compartment purge valve
is blocked to restrict opening to 50 is required to ensure that the valves can close under DBA conditions within the times assumed in the analyses
of References 1 and 2. If a LOCA occurs, the purge valves must close to
maintain containment leakage within the values assumed in the accident
analysis. At other times when purge valves are required to be capable of
closing (e.g., during movement of irradiated fuel assemblies),
pressurization concerns are not present, thus the purge valves can be
fully open. The 18-month Frequency is appropriate because the blocking
devices are typically removed only during a refueling outage.
This SR ensures that the combined leakage rate of all Shield Building
bypass leakage paths is less than or equal to the specified leakage rate.
This provides assurance that the assumptions in the safety analysis are
met. The as-left bypass leakage rate prior to the first startup after
performing a leakage test, requires calculation using maximum pathway
leakage (leakage through the worse of the two isolation valves). If the
penetration is isolated by use of one closed and de-activated automatic
valve, closed manual valve, or blind flange, then the leakage rate of the
isolated bypass leakage path is assumed to be the actual pathway
leakage through the isolation device. If both isolation valves in the
penetration are closed, the actual leakage rate is the lesser leakage rate
of the two valves. At all other times, the leakage rate will be calculated
using minimum pathway leakage.
The frequency is required by the Containment Leakage Rate Testing
Program. This SR simply imposes additional acceptance criteria.
Although not a part of L a , the Shield Building Bypass leakage path combined leakage rate is determined using the 10 CFR 50, Appendix J, Option B, Type B and C leakage rates for the applicable barriers.
Containment Isolation Valves B 3.6.3 BASES Watts Bar - Unit 2 B 3.6-26 (developmental)
A REFERENCES 1. Watts Bar FSAR, Section 15.0, "Accident Analysis." 2. Watts Bar FSAR, Section 6.2.4.2, "Containment Isolation System Design," and Table 6.2.4-1, "Containment Penetrations and Barriers." 3. Generic Issue B-20, "Containment Leakage Due to Seal Deterioration." 4. Title 10, Code of Federal Regulations, Part 50 Appendix J, Option B, "Primary Reactor Containment Leakage Testing for
Water-Cooled Power Reactors - Performance - Based
Requirements."
Containment Pressure B 3.6.4 (continued)
Watts Bar - Unit 2 B 3.6-27 (developmental)
B B 3.6 CONTAINMENT SYSTEMS B 3.6.4 Containment Pressure
BASES BACKGROUND The containment pressure is limited during normal operation to preserve the initial conditions assumed in the accident analyses for a loss of
coolant accident (LOCA) or steam line break (SLB). These limits also prevent the containment pressure from exceeding the containment design
negative pressure differential (-2.0 psid) with respect to the shield building
annulus atmosphere in the event of inadvertent actuation of the
Containment Spray System or Air Return Fans.
Containment pressure is a process variable that is monitored and
controlled. The containment pressure limits are derived from the input
conditions used in the containment functional analyses and the
containment structure external pressure analysis. Should operation occur
outside these limits coincident with a Design Basis Accident (DBA), post
accident containment pressures could exceed calculated values.
APPLICABLE
SAFETY ANALYSES Containment internal pressure is an initial condition used in the DBA
analyses to establish the maximum peak containment internal pressure.
The limiting DBAs considered, relative to containment pressure, are the
LOCA and SLB, which are analyzed using computer pressure transients.
The worst case LOCA generates larger mass and energy release than
the worst case SLB. Thus, the LOCA event bounds the SLB event from
the containment peak pressure standpoint (Ref. 1).
The initial pressure condition used in the containment analysis was
15.0 psia. This resulted in a maximum peak pressure from a LOCA of
10.23 psig. The containment analysis (Ref. 1) shows that the maximum allowable internal containment pressure, P a (15.0 psig), bounds the calculated results from the limiting LOCA. The maximum containment
pressure resulting from the worst case LOCA, does not exceed the
containment design pressure, 13.5 psig.
Containment Pressure B 3.6.4 BASES (continued)
Watts Bar - Unit 2 B 3.6-28 (developmental)
A APPLICABLE SAFETY ANALYSES (continued)
The containment was also designed for an external pressure load
equivalent to 2.0 psig. The inadvertent actuation of the Containment
Spray System was analyzed to determine the resulting reduction in
containment pressure. The initial pressure condition used in this analysis
was -0.1 psig. This resulted in a minimum pressure inside containment of
1.4 psig, which is less than the design load.
For certain aspects of transient accident analyses, maximizing the
calculated containment pressure is not conservative. In particular, the
cooling effectiveness of the Emergency Core Cooling System during the
core reflood phase of a LOCA analysis increases with increasing
containment backpressure. Therefore, for the reflood phase, the
containment backpressure is calculated in a manner designed to
conservatively minimize, rather than maximize, the containment pressure
response in accordance with 10 CFR 50, Appendix K (Ref. 2).
Containment pressure satisfies Criterion 2 of the NRC Policy Statement.
LCO Maintaining containment pressure at less than or equal to the LCO upper pressure limit ensures that, in the event of a DBA, the resultant peak
containment accident pressure will remain below the containment design
pressure. Maintaining containment pressure at greater than or equal to
the LCO lower pressure limit ensures that the containment will not exceed
the design negative differential pressure following the inadvertent
actuation of the Containment Spray System or Air Return Fans.
APPLICABILITY In MODES 1, 2, 3, and 4, a DBA could cause a release of radioactive material to containment. Since maintaining containment pressure within
limits is essential to ensure initial conditions assumed in the accident
analyses are maintained, the LCO is applicable in MODES 1, 2, 3 and 4.
In MODES 5 and 6, the probability and consequences of these events are
reduced due to the pressure and temperature limitations of these
MODES. Therefore, maintaining containment pressure within the limits of
the LCO is not required in MODES 5 or 6.
Containment Pressure B 3.6.4 BASES (continued)
Watts Bar - Unit 2 B 3.6-29 (developmental)
B ACTIONS A.1 When containment pressure is not within the limits of the LCO, it must be
restored to within these limits within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. The Required Action is
necessary to return operation to within the bounds of the containment
analysis. The 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Completion Time is consistent with the ACTIONS of
LCO 3.6.1, "Containment," which requires that containment be restored to
OPERABLE status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
B.1 and B.2
If containment pressure cannot be restored to within limits 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 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to MODE 5 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 SR 3.6.4.1
Verifying that containment pressure is within limits ( -0.1 and +0.3 psid relative to the annulus, value does not account for instrument error) ensures that plant operation remains within the limits assumed in the
containment analysis. The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Frequency of this SR was developed
based on operating experience related to trending of containment
pressure variations during the applicable MODES. Furthermore, the 12
hour Frequency is considered adequate in view of other indications
available in the control room, including alarms, to alert the operator to an
abnormal containment pressure condition.
REFERENCES 1. Watts Bar FSAR, Section 6.2.1, "Containment Functional Design." 2. Title 10, Code of Federal Regulations, Part 50, Appendix K, "ECCS Evaluation Models."
Containment Air Temperature B 3.6.5 (continued)
Watts Bar - Unit 2 B 3.6-30 (developmental)
A B 3.6 CONTAINMENT SYSTEMS B 3.6.5 Containment Air Temperature
BASES BACKGROUND The containment structure serves to contain radioactive material that may be released from the reactor core following a Design Basis Accident (DBA). The containment average air temperature is limited, during
normal operation, to preserve the initial conditions assumed in the
accident analyses for a loss of coolant accident (LOCA) or steam line
break (SLB).
The containment average air temperature limit is derived from the input
conditions used in the containment functional analyses and the
containment structure external pressure analyses. This LCO ensures that
initial conditions assumed in the analysis of containment response to a
DBA are not violated during plant operations. The total amount of energy
removed from containment by the Containment Spray and Cooling
systems during post accident conditions is dependent upon the energy
released to the containment due to the event, as well as the initial
containment temperature and pressure.
APPLICABLE
SAFETY ANALYSES Containment average air temperature is an initial condition used in the
DBA analyses that establishes the containment environmental
qualification operating envelope for both pressure and temperature. The
limit for containment average air temperature ensures that operation is
maintained within the assumptions used in the DBA analyses for
containment (Ref. 1).
The limiting DBAs considered relative to containment OPERABILITY are
the LOCA and SLB. The DBA LOCA and SLB are analyzed using
computer codes designed to predict the resultant containment pressure
transients. No two DBAs are assumed to occur simultaneously or
consecutively. The postulated DBAs are analyzed with regard to
Engineered Safety Feature (ESF) systems, assuming the loss of one ESF
bus, which is the worst case single active failure, resulting in one train
each of Containment Spray System, Residual Heat Removal System, and
Air Return System being rendered inoperable.
Containment Air Temperature B 3.6.5 BASES (continued)
Watts Bar - Unit 2 B 3.6-31 (developmental)
B APPLICABLE SAFETY ANALYSES (continued)
The limiting DBA for the maximum peak containment air temperature is
an SLB. For the upper compartment, the initial containment average air
temperature assumed in this design basis analyses (Ref. 2) is 85 F. For the lower compartment, the initial average containment air temperature
assumed in this design basis analyses is 120 F. These temperatures result in a maximum containment air temperature.
The higher temperature limits are also considered in the depressurization
analyses to ensure that the minimum pressure limit is maintained
following an inadvertent actuation of the Containment Spray System for
both containment compartments.
The containment pressure transient is sensitive to the initial air mass in
containment and, therefore, to the initial containment air temperature.
The limiting DBA for establishing the maximum peak containment internal
pressure is a LOCA. The lower temperature limits, 85 F for the upper compartment and 100 F for the lower compartment, are used in this analysis to ensure that, in the event of an accident, the maximum
containment internal pressure will not be exceeded in either containment
compartment.
Containment average air temperature satisfies Criterion 2 of the NRC
Policy Statement.
LCO During a DBA, with an initial containment average air temperature within the LCO temperature limits, the resultant peak accident temperature is
maintained below the containment design temperature. As a result, the
ability of containment to perform its design function is ensured. In
MODES 2, 3 and 4, containment air temperature may be as low as 60 F (value does not account for instrument error) because the resultant
calculated peak containment accident pressure would not exceed the
design pressure due to a lesser amount of energy released from the pipe
break in these MODES (Ref. 3).
APPLICABILITY In MODES 1, 2, 3, and 4, a DBA could cause a release of radioactive material to containment. In MODES 5 and 6, the probability and
consequences of these events are reduced due to the pressure and
temperature limitations of these MODES. Therefore, maintaining
containment average air temperature within the limit is not required in
MODE 5 or 6.
Containment Air Temperature B 3.6.5 BASES (continued)
(continued)
Watts Bar - Unit 2 B 3.6-32 (developmental)
B ACTIONS A.1 When containment average air temperature in the upper or lower
compartment is not within the limit of the LCO, the average air
temperature in the affected compartment must be restored to within limits
within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. This Required Action is necessary to return operation to
within the bounds of the containment analysis. The 8-hour Completion
Time is acceptable considering the sens itivity of the analysis to variations in this parameter and provides sufficient time to correct minor problems.
B.1 and B.2
If the containment average air temperature cannot be restored to within
its limits 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 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to MODE 5
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 SR 3.6.5.1 and SR 3.6.5.2
LCO 3.6.5 specifies that the containment average air temperature shall
be the following values which do not account for instrument error:
- a. 85 F and 110 F for the containment upper compartment, and
- b. 100 F and 120 F for the containment lower compartment.
Verifying that containment average air temperature is within the LCO
limits ensures that containment operation remains within the limits
assumed for the containment analyses. In order to determine the
containment average air temperature, a weighted average is calculated
using measurements taken at locations within the containment selected to
provide a representative sample of the overall containment atmosphere.
The 24-hour Frequency of these SRs is considered acceptable based on
observed slow rates of temperature increase within containment as a
result of environmental heat sources (due to the large volume of
containment). Furthermore, the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Frequency is considered
adequate in view of other indications available in the control room, including alarms, to alert the operator to an abnormal containment
temperature condition.
Containment Air Temperature B 3.6.5 BASES Watts Bar - Unit 2 B 3.6-33 (developmental)
B REFERENCES 1. Watts Bar FSAR, Section 6.2, "Containment Systems." 2. Watts Bar System Description N3-30RB-4002R5, "Reactor Building Ventilation System." 3. Westinghouse Letter WAT-D-10698, dated November 23, 1999.
Containment Spray System B 3.6.6 (continued)
Watts Bar - Unit 2 B 3.6-34 (developmental)
A B 3.6 CONTAINMENT SYSTEMS B 3.6.6 Containment Spray System
BASES BACKGROUND The Containment Spray System provides containment atmosphere cooling to limit post accident pressure and temperature in containment to
less than the design values. Reduction of containment pressure helps
reduce the release of fission product radioactivity from containment to the
environment, in the event of a Design Basis Accident (DBA). The
Containment Spray System is designed to meet the requirements of
10 CFR 50, Appendix A, GDC 38, "Containment Heat Removal," GDC 39, "Inspection of Containment Heat Removal Systems," and GDC 40, "Testing of Containment Heat Removal Systems," (Ref. 1), or other
documents that were appropriate at the time of licensing (identified on a
plant specific basis).
The Containment Spray System consists of two separate trains of equal
capacity, each capable of meeting the system design basis spray
coverage. Each train includes a containment spray pump, one
containment spray heat exchanger, a spray header, nozzles, valves, and
piping. Each train is powered from a separate Engineered Safety Feature (ESF) bus. The refueling water storage tank (RWST) supplies borated
water to the Containment Spray System during the injection phase of
operation. In the recirculation mode of operation, containment spray
pump suction is transferred from the RWST to the containment
recirculation sump(s).
The diversion of a portion of the recirculation flow from each train of the
Residual Heat Removal (RHR) System to additional redundant spray
headers completes the Containment Spray System heat removal
capability. Each RHR train is capable of supplying spray coverage, if
required, to supplement the Containment Spray System.
The Containment Spray System and RHR System provide a spray of
subcooled borated water into the upper region of containment to limit the
containment pressure and temperature during a DBA. In the recirculation
mode of operation, heat is removed from the containment sump water by
the Containment Spray System and RHR heat exchangers. Each train of
the Containment Spray System, supplemented by a train of RHR spray, provides adequate spray coverage to meet the system design
requirements for containment heat removal.
Containment Spray System B 3.6.6 BASES (continued)
Watts Bar - Unit 2 B 3.6-35 (developmental)
A BACKGROUND (continued)
The Containment Spray System is ac tuated either automatically by a containment High-High pressure signal or manually. An automatic actuation starts the two containment spray pumps, opens the containment
spray pump discharge valves, and begins the injection phase. A manual
actuation of the Containment Spray System requires the operator to
actuate two separate switches on the main control board to begin the
same sequence. The injection phase continues until an RWST level
Low-Low alarm is received. The Low-Low alarm for the RWST signals
the operator to manually align the system to the recirculation mode. The
Containment Spray System in the recirculation mode maintains an
equilibrium temperature between the containment atmosphere and the
recirculated sump water. Operation of the Containment Spray System in
the recirculation mode is controlled by the operator in accordance with the
emergency operating procedures.
The RHR spray operation is initiated manually, when required by the
emergency operating procedures, after the Emergency Core Cooling
System (ECCS) is operating in the recirculation mode. The RHR sprays
are available to supplement the Containment Spray System, if required, in
limiting containment pressure. This additional spray capacity would
typically be used after the ice bed has been depleted and in the event that
containment pressure rises above a pre-determined limit.
The Containment Spray System is an ESF system. It is designed to
ensure that the heat removal capability required during the post accident
period can be attained.
The operation of the ice condenser is adequate to assure pressure
suppression during the initial blowdown of steam and water from a DBA.
During the post blowdown period, the Air Return System (ARS) is
automatically started. The ARS returns upper compartment air through
the divider barrier to the lower compartment. This serves to equalize
pressures in containment and to continue circulating heated air and
steam through the ice condenser, where heat is removed by the
remaining ice and by the Containment Spray System after the ice has
melted.
The Containment Spray System limits the temperature and pressure that
could be expected following a DBA. Protection of containment integrity
limits leakage of fission product radioactivity from containment to the
environment.
Containment Spray System B 3.6.6 BASES (continued)
(continued)
Watts Bar - Unit 2 B 3.6-36 (developmental)
B APPLICABLE SAFETY ANALYSES The limiting DBAs considered relative to containment are the loss of
coolant accident (LOCA) and the steam line break (SLB). The DBA
LOCA and SLB are analyzed using computer codes designed to predict
the resultant containment pressure and temperature transients. No two
DBAs are assumed to occur simultaneously or consecutively. The
postulated DBAs are analyzed, in regard to containment ESF systems, assuming the loss of one ESF bus, which is the worst case single active
failure, resulting in one train of the Containment Spray System, the RHR
System, and the ARS being rendered inoperable (Ref. 2).
The DBA analyses show that the maximum peak containment pressure of
10.23 psig results from the LOCA analysis, and is calculated to be less than the containment design pressure. The maximum peak containment
atmosphere temperature results from the SLB analysis. The calculated
transient containment atmosphere temperatures are acceptable for the
DBA SLB.
The modeled Containment Spray System actuation from the containment analysis is based on a response time associated with exceeding the
containment High-High pressure signal setpoint to achieving full flow
through the containment spray nozzles. A delayed response time
initiation provides conservative analyses of peak calculated containment
temperature and pressure responses. The Containment Spray System
total response time of 234 seconds is composed of signal delay, diesel generator startup, and system startup time.
For certain aspects of transient accident analyses, maximizing the
calculated containment pressure is not conservative. In particular, the
ECCS cooling effectiveness during the core reflood phase of a LOCA
analysis increases with increasing containment backpressure. For these
calculations, the containment backpressure is calculated in a manner
designed to conservatively minimize, rather than maximize, the calculated
transient containment pressures in accordance with 10 CFR 50, Appendix K (Ref. 3).
Inadvertent actuation of the Containment Spray System is evaluated in
the analysis, and the resultant reduction in containment pressure is
calculated. The maximum calculated steady state pressure differential
relative to the Shield Building annulus is 1.4 psid, which is below the
containment design external pressure load of 2.0 psid.
The Containment Spray System satisfies Criterion 3 of the NRC Policy
Statement.
Containment Spray System B 3.6.6 BASES (continued)
(continued)
Watts Bar - Unit 2 B 3.6-37 (developmental)
B LCO During a DBA, one train of Containment Spray System and RHR Spray System is required to provide the heat removal capability assumed in the safety analyses. To ensure that these requirements are met, two
containment spray trains and two RHR spray trains must be OPERABLE
with power from two safety related, independent power supplies.
Therefore, in the event of an accident, at least one train in each system
operates.
Each containment spray train typically includes a spray pump, header, valves, a heat exchanger, nozzles, piping, instruments, and controls to
ensure an OPERABLE flow path capable of taking suction from the
RWST upon an ESF actuation signal and transferring suction to the containment sump. This suction path realignment is accomplished by manual operator action upon receipt of a Low-Low level alarm for the RWST.
Each RHR spray train includes a pump, header, valves, a heat
exchanger, nozzles, piping, instruments, and controls to ensure an
OPERABLE flow path capable of taking suction from the containment
sump and supplying flow to the spray header.
APPLICABILITY In MODES 1, 2, 3, and 4, a DBA could cause a release of radioactive material to containment and an increase in containment pressure and
temperature requiring the operation of the Containment Spray System.
A Note has been added which states the RHR spray trains are not
required in MODE 4. The containment spray system does not require
supplemental cooling from the RHR spray in MODE 4.
In MODES 5 and 6, the probability and consequences of these events are
reduced because of the pressure and temperature limitations of these
MODES. Thus, the Containment Spray System is not required to be
OPERABLE in MODE 5 or 6.
ACTIONS A.1 and B.1
With one containment spray train and/or RHR spray train inoperable, the
affected train must be restored to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
The components in this degraded condition are capable of providing
100% of the heat removal needs after an accident. The 72-hour
Completion Time was developed taking into account the redundant heat
removal capabilities afforded by the OPERABLE train and the low
probability of a DBA occurring during this period.
Containment Spray System B 3.6.6 BASES (continued)
Watts Bar - Unit 2 B 3.6-38 (developmental)
B ACTIONS (continued)
C.1 and C.2 If the affected containment spray train and/or RHR spray train 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
6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to MODE 5 within 84 hours9.722222e-4 days <br />0.0233 hours <br />1.388889e-4 weeks <br />3.1962e-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. The extended interval to reach
MODE 5 allows additional time and is reasonable when considering that
the driving force for a release of radioactive material from the Reactor
Coolant System is reduced in MODE 3.
SURVEILLANCE
REQUIREMENTS SR 3.6.6.1
Verifying the correct alignment of manual, power operated, and automatic
valves, excluding check valves, in the Containment Spray System
provides assurance that the proper flow path exists for Containment
Spray System operation. This SR does not apply to valves that are
locked, sealed, or otherwise secured in position since they were verified
in the correct position prior to being secured. This SR does not require
any testing or valve manipulation. Rather, it involves verification that those valves outside containment and capable of potentially being mis-
positioned, are in the correct position.
Verifying that each containment spray pump's developed head at the flow
test point is greater than or equal to the required developed head ensures
that spray pump performance has not degraded during the cycle. Flow
and differential head are normal tests of centrifugal pump performance
required by the American Society of Mechanical Engineers (ASME) OM
Code (Ref. 4). Since the containment spray pumps cannot be tested with
flow through the spray headers, they are tested on bypass flow. This test
confirms one point on the pump design curve and is indicative of overall
performance. Such inservice tests confirm component OPERABILITY, trend performance, and detect incipient failures by indicating abnormal
performance. The Frequency of this SR is in accordance with the
Inservice Testing Program.
Containment Spray System B 3.6.6 BASES (continued)
Watts Bar - Unit 2 B 3.6-39 (developmental)
A SURVEILLANCE REQUIREMENTS (continued)
SR 3.6.6.3 and SR 3.6.6.4
These SRs require verification that each automatic containment spray
valve actuates to its correct position and each containment spray pump
starts upon receipt of an actual or simulated containment spray actuation
signal. This Surveillance is not required for valves that are locked, sealed, or otherwise secured in the required position under administrative
control. Containment spray pump start verification may be performed by
testing breaker actuation without pump start (breaker is racked out in its "test position") and observation of the local or remote pump start lights (breaker energization light). The 18-month Frequency is based on the
need to perform these Surveillances under the conditions that apply
during a plant outage and the potential for an unplanned transient if the
Surveillances were performed with the reactor at power. Operating
experience has shown these components usually pass the Surveillances
when performed at the 18-month Frequency. Therefore, the Frequency
was concluded to be acceptable from a reliability standpoint.
The surveillance of containment sump isolation valves is also required by
SR 3.6.6.3. A single surveillance may be used to satisfy both
requirements.
With the containment spray inlet valves closed and the spray header
drained of any solution, low pressure air or smoke can be blown through
test connections. This SR ensures that each spray nozzle required by the
design bases is unobstructed and that spray coverage of the containment
during an accident is not degraded. Because of the passive design of the
nozzle, a test at the first refueling and at 10 year intervals are considered
adequate to detect obstruction of the spray nozzles.
The Surveillance descriptions from Bases 3.5.2 for SR 3.5.2.2 and
SR 3.5.2.4 apply as applicable to the RHR spray system.
Containment Spray System B 3.6.6 BASES (continued)
Watts Bar - Unit 2 B 3.6-40 (developmental)
A REFERENCES 1. Title 10, Code of Federal Regulations, Part 50, Appendix A, "General Design Criterion (GDC) 38, "Containment Heat Removal," GDC 39, "Inspection of Containment Heat Removal System,"
GDC 40, "Testing of Containment Heat Removal Systems, and
GDC 50, "Containment Design Basis." 2. Watts Bar FSAR, Section 6.2, "Containment Systems." 3. Title 10, Code of Federal Regulations, Part 50, Appendix K, "ECCS Evaluation Models." 4. American Society of Mechanical Engineers (ASME) OM Code, "Code for Operation and Maintenance of Nuclear Power Plants."
Hydrogen Recombiners B 3.6.7 BASES (continued)
(continued)
Watts Bar - Unit 2 B 3.6-41 (developmental)
B B 3.6 CONTAINMENT SYSTEMS B 3.6.7 The Bases for Specification 3.6.7 have been Deleted
HMS B 3.6.8 (continued)
Watts Bar - Unit 2 B 3.6-42 (developmental)
A B 3.6 CONTAINMENT SYSTEMS B 3.6.8 Hydrogen Mitigation System (HMS)
BASES BACKGROUND The HMS consists of two groups of 34 ignitors distributed throughout the containment. The HMS reduces the potential for breach of primary
containment due to a hydrogen oxygen reaction in post accident
environments. The HMS is required by 10 CFR 50.44, "Standards for
Combustible Gas Control Systems in Light-Water-Cooled Reactors" (Ref. 1), and Appendix A, GDC 41, "Containment Atmosphere Cleanup" (Ref. 2), to reduce the hydrogen concentration in the primary containment
following a degraded core accident. The HMS must be capable of
handling an amount of hydrogen equivalent to that generated from a
metal water reaction involving 75% of the fuel cladding surrounding the active fuel region (excluding the plenum volume).
10 CFR 50.44 (Ref. 1) requires plants with ice condenser containments to
install suitable hydrogen control systems that would accommodate an
amount of hydrogen equivalent to that generated from the reaction of
75% of the fuel cladding with water. The HMS provides this required
capability. This requirement was placed on ice condenser plants because
of their small containment volume and low design pressure (compared
with pressurized water reactor dry containments). Calculations indicate
that if hydrogen equivalent to that generated from the reaction of 75% of
the fuel cladding with water were to collect in the primary containment, the resulting hydrogen concentration would be far above the lower
flammability limit such that, if ignited from a random ignition source, the
resulting hydrogen burn would seriously challenge the containment and
safety systems in the containment.
The HMS is based on the concept of controlled ignition using thermal
ignitors, designed to be capable of functioning in a post accident
environment, seismically supported, and capable of actuation from the
control room. A total of 68 ignitors are distributed throughout the various
regions of containment in which hydrogen could be released or to which it
could flow in significant quantities. The ignitors are arranged in two independent trains such that each containment region has at least two ignitors, one from each train, controlled and powered redundantly so
that ignition would occur in each region even if one train failed to
energize.
HMS B 3.6.8 BASES (continued)
Watts Bar - Unit 2 B 3.6-43 (developmental)
A BACKGROUND (continued)
When the HMS is initiated, the ignitor elements are energized and heat up to a surface temperature 1700 F. At this temperature, they ignite the hydrogen gas that is present in the airspace in the vicinity of the ignitor.
The HMS depends on the dispersed location of the ignitors so that local
pockets of hydrogen at increased concentrations would burn before
reaching a hydrogen concentration significantly higher than the lower
flammability limit. Hydrogen ignition in the vicinity of the ignitors is
assumed to occur when the local hydrogen concentration reaches a
minimum 5.0 volume percent (v/o).
APPLICABLE
SAFETY ANALYSES The HMS causes hydrogen in containment to burn in a controlled manner
as it accumulates following a degraded core accident (Ref. 3). Burning
occurs at the lower flammability concentration, where the resulting
temperatures and pressures are relatively benign. Without the system, hydrogen could build up to higher concentrations that could result in a
violent reaction if ignited by a random ignition source after such a buildup.
The hydrogen ignitors are not included for mitigation of a Design Basis
Accident (DBA) because an amount of hydrogen equivalent to that
generated from the reaction of 75% of the fuel cladding with water is far in excess of the hydrogen calculated for the limiting DBA loss of coolant accident (LOCA). The hydrogen concentration resulting from a DBA can
be maintained less than the flammability limit using the hydrogen
recombiners. The hydrogen ignitors, however, have been shown by
probabilistic risk analysis to be a significant contributor to limiting the
severity of accident sequences that are commonly found to dominate risk
for plants with ice condenser containments. As such, the hydrogen
ignitors are considered to be risk significant in accordance with the NRC
Policy Statement.
HMS B 3.6.8 BASES (continued)
(continued)
Watts Bar - Unit 2 B 3.6-44 (developmental)
A LCO Two HMS trains must be OPERABLE with power from two independent, safety related power supplies.
For this plant, an OPERABLE HMS train consists of 33 of 34 ignitors
energized on the train.
Operation with at least one HMS train ensures that the hydrogen in
containment can be burned in a controlled manner. Unavailability of both
HMS trains could lead to hydrogen buildup to higher concentrations, which could result in a violent reaction if ignited. The reaction could take
place fast enough to lead to high temperatures and overpressurization of
containment and, as a result, breach containment or cause containment
leakage rates above those assumed in the safety analyses. Damage to
safety related equipment located in containment could also occur.
APPLICABILITY Requiring OPERABILITY in MODES 1 and 2 for the HMS ensures its immediate availability after safety injection and scram actuated on a
LOCA initiation. In the post accident environment, the two HMS
subsystems are required to control the hydrogen concentration within
containment to near its flammability limit of 4.0 v/o assuming a worst case
single failure. This prevents overpressurization of containment and
damage to safety related equipment and instruments located within
containment.
In MODES 3 and 4, both the hydrogen production rate and the total
hydrogen production after a LOCA would be significantly less than that calculated for the DBA LOCA. Also, because of the limited time in these MODES, the probability of an accident requiring the HMS is low.
Therefore, the HMS is not required in MODES 3 and 4.
In MODES 5 and 6, the probability and consequences of a LOCA are
reduced due to the pressure and temperature limitations of these
MODES. Therefore, the HMS is not required to be OPERABLE in
MODES 5 and 6.
HMS B 3.6.8 BASES (continued)
(continued)
Watts Bar - Unit 2 B 3.6-45 (developmental)
A ACTIONS A.1 and A.2 With one HMS train inoperable, the inoperable train must be restored to
OPERABLE status within 7 days or the OPERABLE train must be verified
OPERABLE frequently by performance of SR 3.6.8.1. The 7-day
Completion Time is based on the low probability of the occurrence of a
degraded core event that would generate hydrogen in amounts equivalent
to a metal water reaction of 75% of the core cladding, the length of time
after the event that operator action would be required to prevent hydrogen
accumulation from exceeding this limit, and the low probability of failure of
the OPERABLE HMS train. Alternative Required Action A.2, by frequent surveillances, provides assurance that the OPERABLE train continues to be OPERABLE.
B.1 Condition B is one containment region with no OPERABLE hydrogen
ignitor. Thus, while in Condition B, or in Conditions A and B
simultaneously, there would always be ignition capability in the adjacent
containment regions that would provide redundant capability by flame
propagation to the region with no OPERABLE ignitors.
Required Action B.1 calls for the restoration of one hydrogen ignitor in
each region to OPERABLE status within 7 days. The 7-day Completion
Time is based on the same reasons given under Required Action A.1.
C.1 If the HMS subsystem(s) 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 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. The Completion Time of
6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is reasonable, based on operating experience, to reach MODE 3
from full power conditions in an orderly manner and without challenging
plant systems.
HMS B 3.6.8 BASES (continued)
(continued)
Watts Bar - Unit 2 B 3.6-46 (developmental)
A SURVEILLANCE REQUIREMENTS SR 3.6.8.1
This SR confirms that 33 of 34 hydrogen ignitors can be successfully energized in each train. The ignitors are simple resistance elements.
Therefore, energizing provides assurance of OPERABILITY. The
allowance of one inoperable hydrogen ignitor is acceptable because, although one inoperable hydrogen ignitor in a region would compromise
redundancy in that region, the containment regions are interconnected so
that ignition in one region would cause burning to progress to the others (i.e., there is overlap in each hydrogen ignitor's effectiveness between
regions). The Frequency of 92 days has been shown to be acceptable through operating experience.
This SR confirms that the two inoperable hydrogen ignitors allowed by
SR 3.6.8.1 (i.e., one in each train) are not in the same containment
region. The containment regions and hydrogen ignitor locations are
provided in Reference 3. The Frequency of 92 days is acceptable based
on the Frequency of SR 3.6.8.1, which provides the information for
performing this SR.
A more detailed functional test is performed every 18 months to verify
system OPERABILITY. Each glow pl ug is visually examined to ensure that it is clean and that the electrical circuitry is energized. All ignitors (glow plugs), including normally inaccessible ignitors, are visually
checked for a glow to verify that they are energized. Additionally, the
surface temperature of each glow plug is measured to be 1700 F to demonstrate that a temperature sufficient for ignition is achieved. The 18-
month Frequency is based on the 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. Operating experience has s hown that these components usually pass the SR when performed at the 18-month Frequency, which is based on the refueling cycle. Therefore, the Frequency was concluded to be
acceptable from a reliability standpoint.
HMS B 3.6.8 BASES (continued)
Watts Bar - Unit 2 B 3.6-47 (developmental)
A REFERENCES
- 1. Title 10, Code of Federal Regulations, Part 50.44, "Standards for Combustible Gas Control Systems in Light Water-Cooled Power Reactors." 2. Title 10, Code of Federal Regulations, Part 50, Appendix A, General Design Criterion 41, "Containment Atmosphere Cleanup." 3. Watts Bar FSAR, Section 6.2.5A, "Hydrogen Mitigation System Description."
EGTS B 3.6.9 (continued)
Watts Bar - Unit 2 B 3.6-48 (developmental)
A B 3.6 CONTAINMENT SYSTEMS
B 3.6.9 Emergency Gas Treatment System (EGTS)
BASES BACKGROUND The EGTS is required by 10 CFR 50, Appendix A, GDC 41, "Containment Atmosphere Cleanup" (Ref. 1), to ensure that radioactive materials that leak from the primary containment into the shield building (secondary containment) following a Design Basis Accident (DBA) are filtered and adsorbed prior to exhausting to the environment.
The containment has a secondary containment called the shield building, which is a concrete structure that surrounds the steel primary containment vessel. Between the containment vessel and the shield building inner wall is an annular space that collects any containment leakage that may occur following a loss of coolant accident (LOCA). This space also allows for periodic inspection of the outer surface of the steel containment vessel.
The EGTS establishes a negative pressure in the annulus between the shield building and the steel containment vessel. Filters in the system then control the release of radioactive contaminants to the environment. Shield building OPERABILITY is required to ensure retention of primary containment leakage and proper operation of the EGTS.
The EGTS consists of two separate and redundant trains. Each train includes a heater, a prefilter, moisture separators, a high efficiency particulate air (HEPA) filter, an activated charcoal adsorber section for removal of radioiodines, and a fan. Ductwork, valves and/or dampers, and instrumentation also form part of the system. The moisture separators function to reduce the moisture content of the airstream.
A second bank of HEPA filters follows the adsorber section to collect carbon fines and provide backup in case of failure of the main HEPA filter bank. Only the upstream HEPA filter and the charcoal adsorber section are credited in the analysis. The system initiates and maintains a negative air pressure in the shield building by means of filtered exhaust ventilation of the shield building following receipt of a safety injection (SI) signal. The system is described in Reference 2.
EGTS B 3.6.9BASES (continued)
Watts Bar - Unit 2 B 3.6-49 (developmental)
B BACKGROUND (continued) The prefilters remove large particles in the air, and the moisture separators remove entrained water droplets present, to prevent excessive loading of the HEPA filters and charcoal absorbers. Heaters are included to reduce the relative humidity of the airstream on systems that operate in high humidity. Continuous operation of each train, for at least 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />
per month, with heaters on, reduces moisture buildup on their HEPA filters and adsorbers. Cross-over flow ducts are provided between the two trains to allow the active train to draw air through the inactive train and cool the air to keep the charcoal beds on the inactive train from becoming too hot due to absorption of fission products.
The containment annulus vacuum fans maintain the annulus at -5 inches water gauge vacuum during normal operations. During accident conditions, the containment annulus vacuum fans are isolated from the air cleanup portion of the system.
The EGTS reduces the radioactive content in the shield building atmosphere following a DBA. Loss of the EGTS could cause site boundary doses, in the event of a DBA, to exceed the values given in the licensing basis.
APPLICABLE
SAFETY ANALYSES The EGTS design basis is established by the consequences of the limiting DBA, which is a LOCA. The accident analysis (Ref. 3) considers two different single failure scenarios. The first one assumes that only one train of the EGTS is functional due to a postulated single failure that disables the other train. An alternate scenario assumes a single failure of the pressure control loop associated with one train of PCOs. The first scenario is bounding for thyroid dose while the alternate scenario is bounding for beta and gamma doses. The accident analysis accounts for the reduction in airborne radioactive material provided by the number of filter trains in operation for each failure scenario. The amount of fission products available for release from containment is determined for a LOCA.
The safety analysis conservatively assumes the annulus is at atmospheric pressure prior to the LOCA. The analysis further assumes that upon receipt of a Containment Isolation Phase A (CIA) signal from the RPS, the EGTS fans automatically start and achieve a minimum flow of 3600 cfm (per train) within 18 seconds (20 seconds from the initiating event.) This does not include 10 seconds for diesel generator startup.
EGTS B 3.6.9BASES (continued)Watts Bar - Unit 2 B 3.6-50 (developmental)
B APPLICABLE SAFETY ANALYSES (continued) The analysis shows that the annulus pressure will rise to a positive value and then decrease to the EGTS control point for a single failure of one EGTS train, or slightly more negative for a single failure of a pressure control loop associated with one train of PCOs. The normal alignment for both EGTS control loops is the A-Auto position. With both EGTS control loops in A-Auto, both trains will function upon initiation of a CIA signal. In the event of a LOCA, the annulus va cuum control system isolates and both trains of the EGTS pressure control loops will be placed in service to maintain the required negative pressure. If annulus vacuum is lost during normal operations, the A-Auto position is unaffected by the loss of vacuum. This operational configuration is acceptable because the accident dose analysis conservatively assumes the annulus is at atmospheric pressure at event initiation.
The EGTS satisfies Criterion 3 of the NRC Policy Statement.
LCO In the event of a DBA, one EGTS train is required to provide the minimum particulate iodine removal assumed in the safety analysis. Two trains of the EGTS must be OPERABLE to ensure that at least one train will operate, assuming that the other train is disabled by a single active failure.
See TS Bases 3.6.15, "Shield Building," for additional information on EGTS.
APPLICABILITY In MODES 1, 2, 3, and 4, a D BA could lead to fission product release to containment that leaks to the shield building. The large break LOCA, on which this system's design is based, is a full power event. Less severe LOCAs and leakage still require the system to be OPERABLE throughout these MODES. The probability and severity of a LOCA decrease as core power and Reactor Coolant System pressure decrease. With the reactor shut down, the probability of release of radioactivity resulting from such an accident is low.
In MODES 5 and 6, the probability and consequences of a DBA are low due to the pressure and temperature limitations in these MODES. Under these conditions, the Filtration System is not required to be OPERABLE (although one or more trains may be operating for other reasons, such as habitability during maintenance in the shield building annulus).
EGTS B 3.6.9BASES (continued)Watts Bar - Unit 2 B 3.6-51 (developmental)
A ACTIONS (continued)
A.1 With one EGTS train inoperable, the inoperable train must be restored to OPERABLE status within 7 days. The components in this degraded condition are capable of providing 100% of the iodine removal needs after a DBA. The 7-day Completion Time is based on consideration of such factors as the availability of the OPERABLE redundant EGTS train and the low probability of a DBA occurring during this period. The Completion Time is adequate to make most repairs.
B.1 and B.2
If the EGTS train 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 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to MODE 5 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 SR 3.6.9.1 Operating each EGTS train for 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> ensures that all trains are OPERABLE and that all associated controls are functioning properly. It also ensures that blockage, fan or motor failure, or excessive vibration can be detected for corrective action. Operation with the heaters on (automatic heater cycling to maintain temperature) for 10 continuous hours eliminates moisture on the adsorbers and HEPA filters. Experience from filter testing at operating units indicates that the 10-hour period is
adequate for moisture elimination on the adsorbers and HEPA filters.
The 31-day Frequency was developed in consideration of the known reliability of fan motors and controls, the two train redundancy available.
This SR verifies that the required EGTS filter testing is performed in accordance with the Ventilation Filter Testing Program (VFTP - Technical Specification Section 5.7.2.14). The EGTS filter tests are in accordance with Regulatory Guide 1.52 (Ref. 4). The VFTP includes testing HEPA filter performance, charcoal adsorber efficiency, minimum system flow rate, and the physical properties of the activated charcoal (general use and following specific operations).
EGTS B 3.6.9BASES (continued)Watts Bar - Unit 2 B 3.6-52 (developmental)
B SURVEILLANCE REQUIREMENTS (continued)
SR 3.6.9.2 (continued)
Specific test frequencies and additional information are discussed in detail in the VFTP. It should be noted that for the EGTS, the VFTP pressure drop value across the entire filtration unit does not account for
instrument error.
The automatic startup ensures that each EGTS train responds properly.
This testing includes the automatic swapping logic of the EGTS pressure control isolation valves in response to the actuation signal. Performance of this swapping logic test will ensure the availability of EGTS functions in the event of an initial single failure of one of the pressure control loops. The 18-month Frequency is based on the 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. Operating experience has shown that these components usually pass the Surveillance when performed at the 18-month Frequency. Therefore the Frequency was concluded to be acceptable from a reliability standpoint. Furthermore, the SR interval was developed considering that the EGTS equipment OPERABILITY is demonstrated at a 31-day Frequency by SR 3.6.9.1.
The proper functioning of the fans, dampers, filters, adsorbers, etc., as a system is verified by the ability of each train to produce the required system flow rate within the specified timeframe. The 18-month Frequency on a STAGGERED TEST BASIS is consistent with Regulatory Guide 1.52 (Ref. 4) guidance for functional testing.
EGTS B 3.6.9BASES (continued)Watts Bar - Unit 2 B 3.6-53 (developmental)
B REFERENCES 1. Title 10, Code of Federal Regulations, Part 50, Appendix A, General Design Criterion 41, "Containment Atmosphere Cleanup." 2. Watts Bar FSAR, Section 6.5, "Fission Product Removal and Control Systems." 3. Watts Bar FSAR, Section 15.0, "Accident Analysis." 4. Regulatory Guide 1.52, Rev. 2, "Design, Testing and Maintenance Criteria for Post Accident Engineered-Safety-Feature Atmospheric Cleanup System Air Filtration and Absorption Units of Light-Water Cooled Nuclear Power Plants."
ARS B 3.6.10 (continued)
B 3.6 CONTAINMENT SYSTEMS B 3.6.10 Air Return System (ARS)
BASES BACKGROUND The ARS is designed to assure the rapid return of air from the upper to the lower containment compartment after the initial blowdown following a Design Basis Accident (DBA). The return of this air to the lower
compartment and subsequent recirculation back up through the ice
condenser assists in cooling the containment atmosphere and limiting
post accident pressure and temperature in containment to less than
design values. Limiting pressure and temperature reduces the release of
fission product radioactivity from containment to the environment in the
event of a DBA.
The ARS provides post accident hydrogen mixing in selected areas of containment. The ARS draws air from the dome of the containment vessel, from the reactor cavity, and from the ten dead ended (pocketed)
spaces in the containment where there is potential for the accumulation
of hydrogen. The minimum design flow from each potential hydrogen
pocket is sufficient to limit the local concentration of hydrogen.
The ARS consists of two separate trains of equal capacity, each capable
of meeting the design bases. Each train includes a 100% capacity air
return fan, associated damper, and hydrogen collection headers. Each
train is powered from a separate Engineered Safety Features (ESF) bus.
The ARS fans are automatically started by the containment isolation
Phase B signal 8 to 10 minutes after the containment pressure reaches
the pressure setpoint. The time delay ensures that no energy released
during the initial phase of a DBA will bypass the ice bed through the ARS
fans into the upper containment compartment.
After starting, the fans displace air from the upper compartment to the
lower compartment, thereby returning the air that was displaced by the high energy line break blowdown from the lower compartment and equalizing pressures throughout containment. After discharge into the lower compartment, air flows with steam produced by residual heat through the ice condenser doors into the ice condenser compartment
where the steam portion of the flow is condensed. The air flow returns to
the upper compartment through the top deck doors in the upper portion
of the ice condenser compartment. The ARS fans operate continuously
after actuation, circulating air through the containment volume and Watts Bar - Unit 2 B 3.6-54 (developmental)
A ARS B 3.6.10 BASES (continued)
Watts Bar - Unit 2 B 3.6-55 (developmental)
A BACKGROUND (continued) purging all potential hydrogen pockets in containment. When the containment pressure falls below a predetermined value, the ARS fans
are manually de-energized. Thereafter, the fans are manually cycled on and off if necessary to control any additional containment pressure
The ARS also functions, after all the ice has melted, to circulate any
steam still entering the lower compartment to the upper compartment
where the Containment Spray System can cool it.
The ARS is an ESF system. It is designed to ensure that the heat
removal capability required during the post accident period can be
attained. The operation of the ARS, in conjunction with the ice bed, the
Containment Spray System, and the Residual Heat Removal (RHR)
System spray, provides the required heat removal capability to limit post accident conditions to less than the containment design values.
APPLICABLE
SAFETY ANALYSES The limiting DBAs considered relative to containment temperature and
pressure are the loss of coolant accident (LOCA) and the steam line
break (SLB). The LOCA and SLB are analyzed using computer codes
designed to predict the resultant containment pressure and temperature
transients. DBAs are assumed not to occur simultaneously or
consecutively. The postulated DBAs are analyzed, in regard to ESF
systems, assuming the loss of one ESF bus, which is the worst case
single active failure and results in one train each of the Containment
Spray System, RHR System, and ARS being inoperable (Ref. 1). The
DBA analyses show that the maximum peak containment pressure results
from the LOCA analysis and is calculated to be less than the containment design pressure.
For certain aspects of transient accident analyses, maximizing the
calculated containment pressure is not conservative. In particular, the
cooling effectiveness of the Emergency Core Cooling System during the
core reflood phase of a LOCA analysis increases with increasing
containment backpressure. For these calculations, the containment
backpressure is calculated in a manner designed to conservatively
minimize, rather than maximize, the calculated transient containment
pressures, in accordance with 10 CFR 50, Appendix K (Ref. 2).
ARS B 3.6.10 BASES (continued)
Watts Bar - Unit 2 B 3.6-56 (developmental)
A APPLICABLE SAFETY ANALYSES (continued)
The modeled ARS actuation from the containment analysis is based upon
a response time associated with exceeding the containment pressure
High-High signal setpoint to achieving full ARS air flow. A delayed
response time initiation provides conservative analyses of peak
calculated containment temperature and pressure responses. The ARS
total response time of 540
+ 60 seconds consists of the built in signal delay.
The ARS satisfies Criterion 3 of the NRC Policy Statement.
LCO In the event of a DBA, one train of the ARS is required to provide the minimum air recirculation for heat removal and hydrogen mixing assumed
in the safety analyses. To ensure this requirement is met, two trains of
the ARS must be OPERABLE. This will ensure that at least one train will
operate, assuming the worst case single failure occurs, which is in the
ESF power supply.
APPLICABILITY In MODES 1, 2, 3, and 4, a DBA could cause an increase in containment pressure and temperature requiring the operation of the ARS. Therefore, the LCO is applicable in MODES 1, 2, 3, and 4.
In MODES 5 and 6, the probability and consequences of these events are
reduced due to the pressure and temperature limitations of these
MODES. Therefore, the ARS is not required to be OPERABLE in these
MODES.
ACTIONS A.1 If one of the required trains of the ARS is inoperable, it must be restored
to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The components in this degraded
condition are capable of providing 100% of the flow capability after an
accident. The 72-hour Completion Time was developed taking into
account the redundant flow and hydrogen mixing capability of the
OPERABLE ARS train and the low probability of a DBA occurring in this
period.
ARS B 3.6.10 BASES (continued)
Watts Bar - Unit 2 B 3.6-57 (developmental)
A ACTIONS (continued)
B.1 and B.2 If the ARS train 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 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to MODE 5 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 SR 3.6.10.1
Verifying that each ARS fan starts on an actual or simulated actuation
signal, after a delay of 8.0 minutes and 10.0 minutes, and operates for 15 minutes is sufficient to ensure that all fans are OPERABLE and that all associated controls and time delays are functioning properly. It also
ensures that blockage, fan and/or motor failure, or excessive vibration
can be detected for corrective action. The 92 day Frequency was
developed considering the known reliability of fan motors and controls
and the two train redundancy available.
Verifying ARS fan motor current with the return air backdraft dampers
closed confirms one operating condition of the fan. This test is indicative
of overall fan motor performance. Such inservice tests confirm
component OPERABILITY, trend performance, and detect incipient
failures by indicating abnormal performance. The Frequency of 92 days
conforms with the testing requirements for similar ESF equipment and
considers the known reliability of fan motors and controls and the two
train redundancy available.
Verifying the OPERABILITY of the air return damper to the proper
opening torque (Ref. 3) provides assurance that the proper flow path will
exist when the fan is started. By applying the correct torque to the
damper shaft, the damper operation can be confirmed. The Frequency of
92 days was developed considering the importance of the dampers, their
location, physical environment, and probability of failure. Operating
experience has also shown this Frequency to be acceptable.
ARS B 3.6.10 BASES Watts Bar - Unit 2 B 3.6-58 (developmental)
A REFERENCES
- 1. Watts Bar FSAR, Section 6.8, "Air Return Fans." 2. Title 10, Code of Federal Regulations, Part 50, Appendix K, "ECCS Evaluation Models." 3. System Description N3-30RB-4002.
Ice Bed B 3.6.11 (continued)
Watts Bar - Unit 2 B 3.6-59 (developmental)
A B 3.6 CONTAINMENT SYSTEMS B 3.6.11 Ice Bed
BASES BACKGROUND The ice bed consists of over 2,158,000 lbs of ice stored in 1944 baskets within the ice condenser. Its primary purpose is to provide a large heat
sink in the event of a release of energy from a Design Basis Accident (DBA) in containment. The ice would absorb energy and limit
containment peak pressure and temperature during the accident
transient. Limiting the pressure and temperature reduces the release of
fission product radioactivity from containment to the environment in the
event of a DBA.
The ice condenser is an annular compartment enclosing approximately
300 of the perimeter of the upper containment compartment, but penetrating the operating deck so that a portion extends into the lower
containment compartment. The lower portion has a series of hinged
doors exposed to the atmosphere of the lower containment compartment, which, for normal plant operation, are designed to remain closed. At the
top of the ice condenser is another set of doors exposed to the
atmosphere of the upper compartment, which also remain closed during
normal plant operation. Intermediate deck doors, located below the top
deck doors, form the floor of a plenum at the upper part of the ice
condenser. These doors also remain closed during normal plant
operation. The upper plenum area is used to facilitate surveillance and
maintenance of the ice bed.
The ice baskets contain the ice within the ice condenser. The ice bed is
considered to consist of the total volume from the bottom elevation of the
ice baskets to the top elevation of the ice baskets. The ice baskets
position the ice within the ice bed in an arrangement to promote heat
transfer from steam to ice. This arrangement enhances the ice
condenser's primary function of condensing steam and absorbing heat
energy released to the containment during a DBA.
Ice Bed B 3.6.11 BASES (continued)
Watts Bar - Unit 2 B 3.6-60 (developmental)
A BACKGROUND (continued)
In the event of a DBA, the ice condenser inlet doors (located below the operating deck) open due to the pressure rise in the lower compartment.
This allows air and steam to flow from the lower compartment into the ice
condenser. The resulting pressure increase within the ice condenser
causes the intermediate deck doors and the top deck doors to open, which allows the air to flow out of the ice condenser into the upper
compartment. Steam condensation within the ice condenser limits the
pressure and temperature buildup in containment. A divider barrier
separates the upper and lower compartments and ensures that the steam
is directed into the ice condenser.
The ice, together with the containment spray, is adequate to absorb the
initial blowdown of steam and water from a DBA and the additional heat
loads that would enter containment during several hours following the
initial blowdown. The additional heat loads would come from the residual
heat in the reactor core, the hot piping and components, and the
secondary system, including the steam generators. During the post
blowdown period, the Air Return System (ARS) returns upper
compartment air through the divider barrier to the lower compartment.
This serves to equalize pressures in containment and to continue
circulating heated air and steam from the lower compartment through the
ice condenser where the heat is removed by the remaining ice.
As ice melts, the water passes through the ice condenser floor drains into
the lower compartment. Thus, a second function of the ice bed is to be a
large source of borated water (via the containment sump) for long term
Emergency Core Cooling System (ECCS) and Containment Spray
System heat removal functions in the recirculation mode.
A third function of the ice bed and melted ice is to remove fission product
iodine that may be released from the core during a DBA. Iodine removal
occurs during the ice melt phase of the accident and continues as the
melted ice is sprayed into the containment atmosphere by the
Containment Spray System. The ice is adjusted to an alkaline pH that
facilitates removal of radioactive iodine from the containment atmosphere.
The alkaline pH also minimizes the occurrence of the chloride and
caustic stress corrosion on mechanical systems and components
exposed to ECCS and Containment Spray System fluids in the
recirculation mode of operation.
It is important for the ice to be uniformly distributed around the 24 ice
condenser bays and for open flow paths to exist around ice baskets. This
is especially important during the initial blowdown so that the steam and Ice Bed B 3.6.11 BASES (continued)
Watts Bar - Unit 2 B 3.6-61 (developmental)
A BACKGROUND (continued) water mixture entering the lower compartment do not pass through only part of the ice condenser, depleting the ice there while bypassing the ice
in other bays.
Two phenomena that can degrade the ice bed during the long service
period are:
- a. Loss of ice by melting or sublimation; and
- b. Obstruction of flow passages through the ice bed due to buildup of frost or ice. Both of these degrading phenomena are reduced by
minimizing air leakage into and out of the ice condenser.
The ice bed limits the temperature and pressure that could be expected
following a DBA, thus limiting leakage of fission product radioactivity from
containment to the environment.
APPLICABLE
SAFETY ANALYSES The limiting DBAs considered relative to containment temperature and
pressure are the loss of coolant accident (LOCA) and the steam line
break (SLB). The LOCA and SLB are analyzed using computer codes
designed to predict the resultant containment pressure and temperature
transients. DBAs are not assumed to occur simultaneously or
consecutively.
Although the ice condenser is a passive system that requires no electrical
power to perform its function, the Containment Spray System and the
ARS also function to assist the ice bed in limiting pressures and
temperatures. Therefore, the postulated DBAs are analyzed in regards to
containment Engineered Safety Feature (ESF) systems, assuming the
loss of one ESF bus, which is the worst case single active failure and
results in one train each of the Containment Spray System and ARS
being inoperable.
The limiting DBA analyses (Ref. 1) show that the maximum peak
containment pressure results from the LOCA analysis and is calculated to
be less than the containment design pressure. For certain aspects of the
transient accident analyses, maximizing the calculated containment
pressure is not conservative. In particular, the cooling effectiveness of
the ECCS during the core reflood phase of a LOCA analysis increases
with increasing containment backpressure.
Ice Bed B 3.6.11 BASES (continued)
Watts Bar - Unit 2 B 3.6-62 (developmental)
A APPLICABLE SAFETY ANALYSES (continued)
For these calculations, the containment backpressure is calculated in a
manner designed to conservatively minimize, rather than maximize, the
calculated transient containment pressures, in accordance with
10 CFR 50, Appendix K (Ref. 2). The maximum peak containment
atmosphere temperature results from the SLB analysis and is discussed
in the Bases for LCO 3.6.5, "Containment Air Temperature."
In addition to calculating the overall peak containment pressures, the
DBA analyses include calculation of the transient differential pressures
that occur across subcompartment walls during the initial blowdown
phase of the accident transient. The internal containment walls and
structures are designed to withstand these local transient pressure
differentials for the limiting DBAs.
The ice bed satisfies Criterion 3 of the NRC Policy Statement.
LCO The ice bed LCO requires the existence of the required quantity of stored ice, appropriate distribution of the ice and the ice bed, open flow paths
through the ice bed, and appropriate chemical content and pH of the
stored ice. The stored ice functions to absorb heat during a DBA, thereby
limiting containment air temperature and pressure. The chemical content
and pH of the ice provide core SDM (boron content) and remove
radioactive iodine from the containment atmosphere when the melted ice
is recirculated through the ECCS and the Containment Spray System, respectively.
APPLICABILITY In MODES 1, 2, 3, and 4, a DBA could cause an increase in containment pressure and temperature requiring the operation of the ice bed.
Therefore, the LCO is applicable in MODES 1, 2, 3, and 4.
In MODES 5 and 6, the probability and consequences of these events are
reduced due to the pressure and temperature limitations of these
MODES. Therefore, the ice bed is not required to be OPERABLE in
these MODES.
Ice Bed B 3.6.11 BASES (continued)
(continued)
Watts Bar - Unit 2 B 3.6-63 (developmental)
B ACTIONS A.1 If the ice bed is inoperable, it must be restored to OPERABLE status
within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. The Completion Time was developed based on
operating experience, which confirms that due to the very large mass of
stored ice, the parameters comprising OPERABILITY do not change
appreciably in this time period. Because of this fact, the Surveillance
Frequencies are long (months), except for the ice bed temperature, which
is checked every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. If a degraded condition is identified, even for
temperature, with such a large mass of ice it is not possible for the
degraded condition to significantly degrade further in a 48-hour period.
Therefore, 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> is a reasonable amount of time to correct a degraded
condition before initiating a shutdown.
B.1 and B.2
If the ice bed 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 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to MODE 5 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 SR 3.6.11.1
Verifying that the maximum temperature of the ice bed is 27 F (value does not account for instrument error) ensures that the ice is kept well below the melting point. The 12-hour Frequency was based on operating
experience, which confirmed that, due to the large mass of stored ice, it is
not possible for the ice bed temperature to degrade significantly within a
12-hour period and was also based on assessing the proximity of the
LCO limit to the melting temperature.
Furthermore, the 12-hour Frequency is considered adequate in view of
indications in the control room, including the alarm, to alert the operator to
an abnormal ice bed temperature condition. This SR may be satisfied by
use of the Ice Bed Temperature Monitoring System.
Ice Bed B 3.6.11 BASES (continued)
Watts Bar - Unit 2 B 3.6-64 (developmental)
A SURVEILLANCE REQUIREMENTS (continued)
The weighing program is designed to obtain a representative sample of
the ice baskets. The representative sample shall include 6 baskets from
each of the 24 ice condenser bays and shall consist of one basket from
radial rows 1, 2, 4, 6, 8, and 9. If no basket from a designated row can be
obtained for weighing, a basket from the same row of an adjacent bay
shall be weighed.
The rows chosen include the rows nearest the inside and outside walls of
the ice condenser (rows 1 and 2, and 8 and 9, respectively), where heat
transfer into the ice condenser is most likely to influence melting or
sublimation. Verifying the total weight of ice ensures that there is
adequate ice to absorb the required amount of energy to mitigate the
DBAs.
If a basket is found to contain less than 1100 lb of ice, a representative
sample of 20 additional baskets from the same bay shall be weighed.
The average weight of ice in these 21 baskets (the discrepant basket and
the 20 additional baskets) shall be 1100 lb at a 95% confidence level.
[Value does not account for instrument error.]
Weighing 20 additional baskets from the same bay in the event a
Surveillance reveals that a single basket contains less than 1100 lb
ensures that no local zone exists that is grossly deficient in ice. Such a
zone could experience early melt out during a DBA transient, creating a
path for steam to pass through the ice bed without being condensed. The
Frequency of 18 months was based on ice storage tests and the
allowance built into the required ice mass over and above the mass
assumed in the safety analyses. Operating experience has verified that, with the 18 month Frequency, the weight requirements are maintained
with no significant degradation between surveillances.
This SR ensures that the azimuthal distribution of ice is reasonably
uniform, by verifying that the average ice weight in each of three
azimuthal groups of ice condenser bays is within the limit. The
Frequency of 18 months was based on ice storage tests and the
allowance built into the required ice mass over and above the mass
assumed in the safety analyses. Operating experience has verified that, with the 18-month Frequency, the weight requirements are maintained
with no significant degradation between surveillances.
Ice Bed B 3.6.11 BASES (continued)
Watts Bar - Unit 2 B 3.6-65 (developmental)
A SURVEILLANCE REQUIREMENTS (continued)
This SR ensures that the air/steam flow channels through the ice bed
have not accumulated ice blockage that exceeds 15 percent of the total
flow area through the ice bed region. The allowable 15 percent buildup of
ice is based on the analysis of the subcompartment response to a design
basis LOCA with partial blockage of the ice bed flow channels. The
analysis did not perform detailed flow area modeling, but rather lumped
the ice condenser bays into six sections ranging from 2.75 bays to
6.5 bays. Individual bays are acceptable with greater than 15 percent
blockage, as long as 15 percent blockage is not exceeded for any
analysis section.
To provide a 95 percent confidence that flow blockage does not exceed
the allowed 15 percent, the visual inspection must be made for at least
54 (33 percent) of the 162 flow channels per ice condenser bay. The
visual inspection of the ice bed flow channels is to inspect the flow area, by looking down from the top of the ice bed, and where view is achievable
up from the bottom of the ice bed. Flow channels to be inspected are
determined by random sample. As the most restrictive flow passage
location is found at a lattice frame elevation, the 15 percent blockage
criteria only applies to "flow channels" that comprise the area:
- a. between ice baskets, and
- b. past lattice frames and wall panels.
Due to a significantly larger flow area in the regions of the upper deck
grating and the lower inlet plenum and turning vanes, it would require a
gross buildup of ice on these structures to obtain a degradation in
air/steam flow. Therefore, these structures are excluded as part of a flow
channel for application of the 15 percent blockage criteria. Plant and
industry experience have shown that removal of ice from the excluded
structures during the refueling outage is sufficient to ensure they remain
operable throughout the operating cycle. Thus, removal of any gross ice
buildup on the excluded structures is performed following outage
maintenance activities.
Ice Bed B 3.6.11 BASES (continued)
Watts Bar - Unit 2 B 3.6-66 (developmental)
B SURVEILLANCE REQUIREMENTS
SR 3.6.11.4 (continued)
Operating experience has demonstrated that the ice bed is the region that
is the most flow restrictive, due to the normal presence of ice
accumulation on lattice frames and wall panels. The flow area through
the ice basket support platform is not a more restrictive flow area because
it is easily accessible from the lower plenum and is maintained clear of ice
accumulation. There is not a mechanistically credible method for ice to
accumulate on the ice basket support platform during plant operation.
Plant and industry experience has shown that the vertical flow area
through the ice basket support platform remains clear of ice accumulation
that could produce blockage. Normally, only a glaze may develop or exist
on the ice basket support platform which is not significant to blockage of
flow area. Additionally, outage maintenance practices provide measures
to clear the ice basket support platform following maintenance activities of
any accumulation of ice that could block flow areas.
Frost buildup or loose ice is not to be considered as flow channel
blockage, whereas attached ice is considered blockage of a flow channel.
Frost is the solid form of water that is loosely adherent, and can be
brushed off with the open hand.
The Frequency of 18 months was based on ice storage tests and the
allowance built into the required ice mass over and above the mass
assumed in the safety analyses.
Verifying the chemical composition of the stored ice ensures that the
stored ice has a boron concentration of 1800 ppm and 2000 ppm as sodium tetraborate and a high pH, 9.0 and 9.5, in order to meet the requirement for borated water when the melted ice is used in the ECCS
recirculation mode of operation. Additionally, the minimum boron
concentration setpoint is used to assure reactor subcriticality in a post
LOCA environment, while the maximum boron concentration is used as
the bounding value in the hot leg switchover timing calculation (Ref. 3).
This is accomplished by obtaining at least 24 ice samples. Each sample
is taken approximately one foot from the top of the ice of each randomly
selected ice basket in each ice condenser bay. The SR is modified by a
NOTE that allows the boron concentration and pH value obtained from
averaging the individual samples' analysis results to satisfy the
requirements of the SR. If either the average boron concentration or the
average pH value is outside their prescribed limit, then entry into ACTION
Condition A is required. Sodium tetraborate has been proven effective in Ice Bed B 3.6.11 BASES (continued)
Watts Bar - Unit 2 B 3.6-67 (developmental)
A SURVEILLANCE REQUIREMENTS
SR 3.6.11.5 (continued)
maintaining the boron content for long storage periods, and it also
enhances the ability of the solution to remove and retain fission product
iodine. The high pH is required to enhance the effectiveness of the ice
and the melted ice in removing iodine from the containment atmosphere.
This pH range also minimizes the occurrence of chloride and caustic
stress corrosion on mechanical systems and components exposed to
ECCS and Containment Spray System fluids in the recirculation mode of
operation. The Frequency of 54 months is intended to be consistent with
the expected length of three fuel cycles, and was developed considering
these facts:
- a. Long term ice storage tests have determined that the chemical composition of the stored ice is extremely stable;
- b. There are no normal operating mechanisms that decrease the boron concentration of the stored ice, and pH remains within a 9.0 through
9.5 range
when boron concentrations are above approximately
1200 ppm.
- c. Operating experience has demonstrated that meeting the boron concentration and pH requirements has never been a problem; and
- d. Someone would have to enter the containment to take the sample, and, if the unit is at power, that person would receive a radiation
dose.
This SR ensures that a representative sampling of ice baskets, which are
relatively thin walled, perforated cylinders, have not been degraded by
wear, cracks, corrosion, or other damage. Each ice basket must be
raised at least 10 feet for this inspection. However, for baskets where
vertical lifting height is restricted due to overhead obstruction, a camera
shall be used to perform the inspection. The Frequency of 40 months for
a visual inspection of the structural soundness of the ice baskets is based
on engineering judgment and considers such factors as the thickness of
the basket walls relative to corrosion rates expected in their service
environment and the results of the long term ice storage testing.
Ice Bed B 3.6.11 BASES Watts Bar - Unit 2 B 3.6-68 (developmental)
B SURVEILLANCE REQUIREMENTS (continued)
This SR ensures that initial ice fill and any subsequent ice additions meet
the boron concentration and pH requirements of SR 3.6.11.5. The SR is
modified by a NOTE that allows the chemical analysis to be performed on
either the liquid or resulting ice of each sodium tetraborate solution
prepared. If ice is obtained from offsite sources, then chemical analysis
data must be obtained for the ice supplied.
REFERENCES 1. Watts Bar FSAR, Section 6.2, "Containment Systems" 2. Title 10, Code of Federal Regulations, Part 50, Appendix K, "ECCS Evaluation Models" 3. Westinghouse Letter, WAT-D-10686, "Upper Limit Ice Boron Concentration In Safety Analysis"
Ice Condenser Doors B 3.6.12 (continued)
Watts Bar - Unit 2 B 3.6-69 (developmental)
A B 3.6 CONTAINMENT SYSTEMS B 3.6.12 Ice Condenser Doors
BASES BACKGROUND The ice condenser doors consist of the inlet doors, the intermediate deck doors, and the top deck doors. The functions of the doors are to:
- a. Seal the ice condenser from air leakage during the lifetime of the plant; and
- b. Open in the event of a Design Basis Accident (DBA) to direct the hot steam air mixture from the DBA into the ice bed, where the ice would
absorb energy and limit containment peak pressure and temperature
during the accident transient.
Limiting the pressure and temperature following a DBA reduces the
release of fission product radioactivity from containment to the
environment.
The ice condenser is an annular compartment enclosing approximately
300 of the perimeter of the upper containment compartment, but penetrating the operating deck so that a portion extends into the lower
containment compartment. The inlet doors separate the atmosphere of
the lower compartment from the ice bed inside the ice condenser. The
top deck doors are above the ice bed and exposed to the atmosphere of
the upper compartment. The intermediate deck doors, located below the
top deck doors, form the floor of a plenum at the upper part of the ice
condenser. This plenum area is used to facilitate surveillance and
maintenance of the ice bed.
The ice baskets held in the ice bed within the ice condenser are arranged
to promote heat transfer from steam to ice. This arrangement enhances
the ice condenser's primary function of condensing steam and absorbing
heat energy released to the containment during a DBA.
Ice Condenser Doors B 3.6.12 BASES (continued)
Watts Bar - Unit 2 B 3.6-70 (developmental)
A BACKGROUND (continued)
In the event of a DBA, the ice condenser inlet doors (located below the operating deck) open due to the pressure rise in the lower compartment.
This allows air and steam to flow from the lower compartment into the ice
condenser. The resulting pressure increase within the ice condenser
causes the intermediate deck doors and the top deck doors to open, which allows the air to flow out of the ice condenser into the upper
compartment. Steam condensation within the ice condensers limits the
pressure and temperature buildup in containment. A divider barrier
separates the upper and lower compartments and ensures that the steam
is directed into the ice condenser.
The ice, together with the containment spray, serves as a containment
heat removal system and is adequate to absorb the initial blowdown of
steam and water from a DBA as well as the additional heat loads that
would enter containment during the several hours following the initial blowdown. The additional heat loads would come from the residual heat in the reactor core, the hot piping and components, and the secondary
system, including the steam generators. During the post blowdown
period, the Air Return System (ARS) returns upper compartment air
through the divider barrier to the lower compartment. This serves to
equalize pressures in containment and to continue circulating heated air
and steam from the lower compartment through the ice condenser, where
the heat is removed by the remaining ice.
The water from the melted ice drains into the lower compartment where it
serves as a source of borated water (via the containment sump) for the
Emergency Core Cooling System (ECCS) and the Containment Spray
System heat removal functions in the recirculation mode. The ice (via the
Containment Spray System) and the recirculated ice melt also serve to
clean up the containment atmosphere.
The ice condenser doors ensure that the ice stored in the ice bed is
preserved during normal operation (doors closed) and that the ice
condenser functions as designed if called upon to act as a passive heat
sink following a DBA.
APPLICABLE
SAFETY ANALYSES The limiting DBAs considered relative to containment pressure and
temperature are the loss of coolant accident (LOCA) and the steam line
break (SLB). The LOCA and SLB are analyzed using computer codes
designed to predict the resultant containment pressure and temperature
transients. DBAs are assumed not to occur simultaneously or
consecutively.
Ice Condenser Doors B 3.6.12 BASES (continued)
Watts Bar - Unit 2 B 3.6-71 (developmental)
A APPLICABLE SAFETY ANALYSES (continued)
Although the ice condenser is a passive system that requires no electrical
power to perform its function, the Containment Spray System and ARS
also function to assist the ice bed in limiting pressures and temperatures.
Therefore, the postulated DBAs are analyzed with respect to Engineered
Safety Feature (ESF) systems, assuming the loss of one ESF bus, which
is the worst case single active failure and results in one train each of the
Containment Spray System and the ARS being rendered inoperable.
The limiting DBA analyses (Ref. 1) show that the maximum peak
containment pressure results from the LOCA analysis and is calculated to
be less than the containment design pressure. For certain aspects of
transient accident analyses, maximizing the calculated containment
pressure is not conservative. In particular, the cooling effectiveness of
the ECCS during the core reflood phase of a LOCA analysis increases
with increasing containment backpressure. For these calculations, the containment backpressure is calculated in a manner designed to conservatively minimize, rather than maximize, the calculated transient
containment pressures, in accordance with 10 CFR 50, Appendix K (Ref. 2).
The maximum peak containment atmosphere temperature results from
the SLB analysis and is discussed in the Bases for LCO 3.6.5, "Containment Air Temperature."
An additional design requirement was imposed on the ice condenser door
design for a small break accident in which the flow of heated air and
steam is not sufficient to fully open the doors.
For this situation, the doors are designed so that all of the doors would
partially open by approximately the same amount. Thus, the partially opened doors would modulate the flow so that each ice bay would receive
an approximately equal fraction of the total flow.
This design feature ensures that the heated air and steam will not flow
preferentially to some ice bays and deplete the ice there without utilizing
the ice in the other bays.
Ice Condenser Doors B 3.6.12 BASES (continued)
Watts Bar - Unit 2 B 3.6-72 (developmental)
A APPLICABLE SAFETY ANALYSES (continued)
In addition to calculating the overall peak containment pressures, the
DBA analyses include the calculation of the transient differential
pressures that would occur across subcompartment walls during the initial
blowdown phase of the accident transient. The internal containment walls
and structures are designed to withstand the local transient pressure
differentials for the limiting DBAs.
The ice condenser doors satisfy Criterion 3 of the NRC Policy Statement.
LCO This LCO establishes the minimum equipment requirements to assure that the ice condenser doors perform their safety function. The ice
condenser inlet doors, intermediate deck doors, and top deck doors must
be closed to minimize air leakage into and out of the ice condenser, with
its attendant leakage of heat into the ice condenser and loss of ice
through melting and sublimation. The doors must be OPERABLE to
ensure the proper opening of the ice condenser in the event of a DBA.
OPERABILITY includes being free of any obstructions that would limit
their opening, and for the inlet doors, being adjusted such that the
opening and closing torques are within limits. The ice condenser doors function with the ice condenser to limit the pressure and temperature that could be expected following a DBA.
APPLICABILITY In MODES 1, 2, 3, and 4, a DBA could cause an increase in containment pressure and temperature requiring the operation of the ice condenser
doors. Therefore, the LCO is applicable in MODES 1, 2, 3, and 4.
The probability and consequences of these events in MODES 5 and 6 are
reduced due to the pressure and temperature limitations of these
MODES. Therefore, the ice condenser doors are not required to be
OPERABLE in these MODES.
Ice Condenser Doors B 3.6.12 BASES (continued)
(continued)
Watts Bar - Unit 2 B 3.6-73 (developmental)
A ACTIONS A Note provides clarification that, for this LCO, separate Condition entry is allowed for each ice condenser door.
A.1 If one or more ice condenser inlet doors are inoperable due to being
physically restrained from opening, the door(s) must be restored to
OPERABLE status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. The Required Action is necessary to
return operation to within the bounds of the containment analysis. The
1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Completion Time is consistent with the ACTIONS of LCO 3.6.1, "Containment," which requires containment to be restored to OPERABLE
status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
B.1 and B.2
If one or more ice condenser doors are determined to be partially open or
otherwise inoperable for reasons other than Condition A or if a door is
found that is not closed, it is acceptable to continue plant operation for up to 14 days, provided the ice bed temperature instrumentation is monitored
once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to ensure that the open or inoperable door is not
allowing enough air leakage to cause the maximum ice bed temperature
to approach the melting point. The Frequency of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is based on the
fact that temperature changes cannot occur rapidly in the ice bed
because of the large mass of ice involved. The 14-day Completion Time
is based on long term ice storage tests that indicate that if the temperature is maintained below 27 F, there would not be a significant loss of ice from sublimation. If the maximum ice bed temperature is > 27 F at any time, or ice bed temperature is not verified to be within the specified Frequency as augmented by the provisions of SR 3.0.2, the
situation reverts to Condition C and a Completion Time of 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> is
allowed to restore the inoperable door to OPERABLE status or enter into
Required Actions D.1 and D.2. [NOTE: Entry into Condition B is not
required due to personnel standing on or opening an intermediate deck or
upper deck door for short durations to perform required surveillances, minor maintenance such as ice removal, or routine tasks such as system
walkdowns.]
Ice Condenser Doors B 3.6.12 BASES (continued)
Watts Bar - Unit 2 B 3.6-74 (developmental)
A ACTIONS (continued)
C.1 If Required Actions or Completion Times of B.1 or B.2 are not met, the
doors must be restored to OPERABLE status and closed positions within
48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. The 48-hour Completion Time is based on the fact that, with
the very large mass of ice involved, it would not be possible for the
temperature to decrease to the melting point and a significant amount of
ice to melt in a 48-hour period.
D.1 and D.2
If the ice condenser doors 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 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to MODE 5 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 SR 3.6.12.1
Verifying, by means of the Inlet Door Position Monitoring System, that the inlet doors are in their closed positions makes the operator aware of an
inadvertent opening of one or more doors. The Frequency of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />
ensures that operators on each shift are aware of the status of the doors.
Verifying, by visual inspection, that each intermediate deck door is closed
and not impaired by ice, frost, or debris provides assurance that the
intermediate deck doors (which form the floor of the upper plenum where
frequent maintenance on the ice bed is performed) have not been left
open or obstructed. The Frequency of 7 days is based on engineering
judgment and takes into consideration such factors as the frequency of
entry into the intermediate ice condenser deck, the time required for
significant frost buildup, and the probability that a DBA will occur.
Ice Condenser Doors B 3.6.12 BASES (continued)
Watts Bar - Unit 2 B 3.6-75 (developmental)
A SURVEILLANCE REQUIREMENTS (continued)
Verifying, by visual inspection, that the ice condenser inlet doors are not
impaired by ice, frost, or debris provides assurance that the doors are
free to open in the event of a DBA. For this unit, the Frequency of
18 months (3 months during the first year after receipt of license - the
3 month performances during the first year after receipt of license may be
extended to coincide with plant outages) is based on door design, which
does not allow water condensation to freeze, and operating experience, which indicates that the inlet doors very rarely fail to meet their SR
acceptance criteria. Because of high radiation in the vicinity of the inlet doors during power operation, this Surveillance is normally performed during a shutdown.
Verifying the opening torque of the inlet doors provides assurance that no
doors have become stuck in the closed position. The value of 675 in-lb is
based on the design opening pressure on the doors of 1.0 lb/ft
- 2. For this unit, the Frequency of 18 months (3 months during the first year after
receipt of license - the 3 month performances during the first year after
receipt of license may be extended to coincide with plant outages) is
based on the passive nature of the closing mechanism (i.e., once
adjusted, there are no known factors that would change the setting, except possibly a buildup of ice; ice buildup is not likely, however, because of the door design, which does not allow water condensation to
freeze). Operating experience indicates that the inlet doors usually meet
their SR acceptance criteria. Because of high radiation in the vicinity of
the inlet doors during power operation, this Surveillance is normally
performed during a shutdown.
Ice Condenser Doors B 3.6.12 BASES (continued)
Watts Bar - Unit 2 B 3.6-76 (developmental)
A SURVEILLANCE REQUIREMENTS (continued)
The torque test Surveillance ensures that the inlet doors have not
developed excessive friction and that the return springs are producing a
door return torque within limits. The torque test consists of the following:
- 1. Verify that the torque, T(OPEN), required to cause opening motion at the 40 open position is 195 in-lb;
- 2. Verify that the torque, T(CLOSE), required to hold the door stationary (i.e., keep it from closing) at the 40 open position is 78 in-lb; and
- 3. Calculate the frictional torque,
T(FRICT) = 0.5 {T(OPEN) - T(CLOSE)},
and verify that the T(FRICT) is 40 in-lb.
The purpose of the friction and return torque Specifications is to ensure
that, in the event of a small break LOCA or SLB, all of the 24 door pairs
open uniformly. This assures that, during the initial blowdown phase, the
steam and water mixture entering the lower compartment does not pass
through part of the ice condenser, depleting the ice there, while bypassing
the ice in other bays. The Frequency of 18 months (3 months during the
first year after receipt of license - the 3 month performances during the
first year after receipt of license may be extended to concide with plant
outages) is based on the passive nature of the closing mechanism (i.e.,
once adjusted, there are no known factors that would change the setting, except possibly a buildup of ice; ice buildup is not likely, however, because of the door design, which does not allow water condensation to
freeze). Operating experience indicates that the inlet doors very rarely
fail to meet their SR acceptance criteria. Because of high radiation in the
vicinity of the inlet doors during power operation, this Surveillance is normally performed during a shutdown.
Ice Condenser Doors B 3.6.12 BASES (continued)
Watts Bar - Unit 2 B 3.6-77 (developmental)
A SURVEILLANCE REQUIREMENTS (continued)
Verifying the OPERABILITY of the intermediate deck doors provides
assurance that the intermediate deck doors are free to open in the event
of a DBA. The verification consists of visually inspecting the intermediate
doors for structural deterioration, verifying free movement of the vent
assemblies, and ascertaining free movement of each door when lifted
with the applicable force shown below:
DOOR LIFTING FORCE a. Adjacent to crane wall
< 37.4 lb
- b. Paired with door adjacent to crane wall 33.8 lb c. Adjacent to containment wall 31.8 lb d. Paired with door adjacent to containment wall 31.0 lb The above test lifting forces were established based upon test results
gathered on newly manufactured Intermediate Deck Doors set up in
fixturing to simulate plant installation tolerances. The lifting force values
developed were to account for and envelope expected door panel
variations in weight and hinge friction and alignments. The intent of the
surveillance is to establish a method of detecting abnormalities or
deteriorating conditions of the door panels or hinges after completion of
refueling outage maintenance activities.
The 18-month Frequency (3 months during the first year after receipt of
license) is based on the passive design of the intermediate deck doors, the frequency of personnel entry into the intermediate deck, and the fact that SR 3.6.12.2 confirms on a 7 day Frequency that the doors are not impaired by ice, frost, or debris, which are ways a door would fail the
opening force test (i.e., by sticking or from increased door weight).
Ice Condenser Doors B 3.6.12 BASES Watts Bar - Unit 2 B 3.6-78 (developmental)
A SURVEILLANCE REQUIREMENTS (continued)
Verifying, by visual inspection, that the top deck doors are in place, not
obstructed, and verifying free movement of the vent assembly provides
assurance that the doors are performing their function of keeping warm
air out of the ice condenser during normal operation, and would not be
obstructed if called upon to open in response to a DBA. The Frequency of
92 days is based on engineering judgment, which considered such
factors as the following:
- a. The relative inaccessibility and lack of traffic in the vicinity of the doors make it unlikely that a door would be inadvertently left open;
- b. Excessive air leakage would be detected by temperature monitoring in the ice condenser; and
- c. The light construction of the doors would ensure that, in the event of a DBA, air and gases passing through the ice condenser would find a
flow path, even if a door were obstructed.
REFERENCES
- 1. Watts Bar FSAR, Section 15.0, "Accident Analysis." 2. Title 10, Code of Federal Regulations, Part 50, Appendix K, "ECCS Evaluation Models."
Divider Barrier Integrity B 3.6.13 (continued)
Watts Bar - Unit 2 B 3.6-79 (developmental)
A B 3.6 CONTAINMENT SYSTEMS B 3.6.13 Divider Barrier Integrity
BASES BACKGROUND The divider barrier consists of the operating deck and associated seals, personnel access doors, and equipment hatches that separate the upper
and lower containment compartments. Divider barrier integrity is
necessary to minimize bypassing of the ice condenser by the hot steam
and air mixture released into the lower compartment during a Design
Basis Accident (DBA). This ensures that most of the gases pass through
the ice bed, which condenses the steam and limits pressure and
temperature during the accident transient. Limiting the pressure and
temperature reduces the release of fission product radioactivity from
containment to the environment in the event of a DBA.
In the event of a DBA, the ice condenser inlet doors (located below the
operating deck) open due to the pressure rise in the lower compartment.
This allows air and steam to flow from the lower compartment into the ice
condenser. The resulting pressure increase within the ice condenser
causes the intermediate deck doors and the door panels at the top of the
condenser to open, which allows the air to flow out of the ice condenser
into the upper compartment. The ice condenses the steam as it enters, thus limiting the pressure and temperature buildup in containment. The
divider barrier separates the upper and lower compartments and ensures
that the steam is directed into the ice condenser. The ice, together with
the containment spray, is adequate to absorb the initial blowdown of
steam and water from a DBA as well as the additional heat loads that
would enter containment over several hours following the initial
blowdown. The additional heat loads would come from the residual heat
in the reactor core, the hot piping and components, and the secondary
system, including the steam generators. During the post blowdown
period, the Air Return System (ARS) returns upper compartment air
through the divider barrier to the lower compartment. This serves to
equalize pressures in containment and to continue circulating heated air
and steam from the lower compartment through the ice condenser, where
the heat is removed by the remaining ice.
Divider barrier integrity ensures that the high energy fluids released
during a DBA would be directed through the ice condenser and that the
ice condenser would function as designed if called upon to act as a
passive heat sink following a DBA.
Divider Barrier Integrity B 3.6.13 BASES (continued)
(continued)
Watts Bar - Unit 2 B 3.6-80 (developmental)
A APPLICABLE SAFETY ANALYSES Divider barrier integrity ensures the functioning of the ice condenser to
the limiting containment pressure and temperature that could be
experienced following a DBA. The limiting DBAs considered relative to
containment temperature and pressure are the loss of coolant accident (LOCA) and the steam line break (SLB). The LOCA and SLB are
analyzed using computer codes designed to predict the resultant
containment pressure and temperature transients. DBAs are assumed
not to occur simultaneously or consecutively.
Although the ice condenser is a passive system that requires no electrical
power to perform its function, the Containment Spray System and the
ARS also function to assist the ice bed in limiting pressures and
temperatures. Therefore, the postulated DBAs are analyzed, with respect
to containment Engineered Safety Feature (ESF) systems, assuming the
loss of one ESF bus, which is the worst case single active failure and results in the inoperability of one train in both the Containment Spray System and the ARS.
The limiting DBA analyses (Ref. 1) show that the maximum peak
containment pressure results from the LOCA analysis and is calculated to
be less than the containment design pressure. The maximum peak
containment temperature results from the SLB analysis and is discussed
in the Bases for LCO 3.6.5, "Containment Air Temperature."
In addition to calculating the overall peak containment pressures, the
DBA analyses include calculation of the transient differential pressures
that occur across subcompartment walls during the initial blowdown
phase of the accident transient. The internal containment walls and
structures are designed to withstand these local transient pressure
differentials for the limiting DBAs.
The divider barrier satisfies Criterion 3 of the NRC Policy Statement.
Divider Barrier Integrity B 3.6.13 BASES (continued)
(continued)
Watts Bar - Unit 2 B 3.6-81 (developmental)
A LCO This LCO establishes the minimum equipment requirements to ensure that the divider barrier performs its safety function of ensuring that bypass leakage, in the event of a DBA, does not exceed the bypass leakage
assumed in the accident analysis. Included are the requirements that the
personnel access doors and equipment hatches in the divider barrier are
OPERABLE and closed and that the divider barrier seal is properly
installed and has not degraded with time. An exception to the
requirement that the doors be closed is made to allow personnel transit
entry through the divider barrier. The basis of this exception is the
assumption that, for personnel transit, the time during which a door is
open will be short (i.e., shorter than the Completion Time of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> for
Condition A). The divider barrier functions with the ice condenser to limit
the pressure and temperature that could be expected following a DBA.
APPLICABILITY In MODES 1, 2, 3, and 4, a DBA could cause an increase in containment pressure and temperature requiring the integrity of the divider barrier.
Therefore, the LCO is applicable in MODES 1, 2, 3, and 4.
The probability and consequences of these events in MODES 5 and 6 are
low due to the pressure and temperature limitations of these MODES. As
such, divider barrier integrity is not required in these MODES.
ACTIONS A.1 If one or more personnel access doors or equipment hatches are
inoperable or open, except for personnel transit entry, 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> is allowed to
restore the door(s) and equipment hatches to OPERABLE status and the
closed position. The 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Completion Time is consistent with
LCO 3.6.1, "Containment," which requires that containment be restored to
OPERABLE status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
Condition A has been modified by a Note to provide clarification that, for
this LCO, separate Condition entry is allowed for each personnel access door or equipment hatch.
Divider Barrier Integrity B 3.6.13 BASES (continued)
Watts Bar - Unit 2 B 3.6-82 (developmental)
A ACTIONS (continued)
B.1 If the divider barrier seal is inoperable, 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> is allowed to restore the
seal to OPERABLE status. The 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Completion Time is consistent
with LCO 3.6.1, which requires that containment be restored to
OPERABLE status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
C.1 and C.2
If the divider barrier integrity 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 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to MODE 5 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 SR 3.6.13.1
Verification, by visual inspection, that all personnel access doors and
equipment hatches between the upper and lower containment
compartments are closed provides assurance that divider barrier integrity
is maintained prior to the reactor being taken from MODE 5 to MODE 4.
The visual inspection shall include the canal gate and control rod drive
missile shield which penetrate the divider barrier. This SR is necessary
because many of the doors and hatches may have been opened for
maintenance during the shutdown.
Divider Barrier Integrity B 3.6.13 BASES (continued)
Watts Bar - Unit 2 B 3.6-83 (developmental)
A SURVEILLANCE REQUIREMENTS (continued)
Verification, by visual inspection, that the personnel access door and
equipment hatch seals, sealing surfaces, and alignments are acceptable
provides assurance that divider barrier integrity is maintained. This
inspection cannot be made when the door or hatch is closed. Therefore, SR 3.6.13.2 is required for each door or hatch that has been opened, prior to the final closure. Some doors and hatches may not be opened for
long periods of time. Those that use resilient materials in the seals must
be opened and inspected at least once every 10 years to provide
assurance that the seal material has not aged to the point of degraded performance. The Frequency of 10 years is based on the known resiliency of the materials used for seals, the fact that the openings have
not been opened (to cause wear), and operating experience that confirms
that the seals inspected at this Frequency have been found to be
acceptable.
Verification, by visual inspection, after each opening of a personnel
access door or equipment hatch that it has been closed makes the
operator aware of the importance of closing it and thereby provides
additional assurance that divider barrier integrity is maintained while in
applicable MODES.
The divider barrier seal can be field spliced for repair purposes utilizing a
cold bond procedure rather than the original field splice technique of
vulcanization. However, the cold bond adhesive, which works in
conjunction with a bolt array to splice the field joint, could not be heat
aged to 40 years plant life prior to acceptability testing. Prolonged
exposure to the elevated temperatures required for heat aging the seal
material was destructive to the adhesive. The seal material was heat
aged to 40 years equivalent age, and the entire joint assembly was
irradiated to 40 year normal operation plus accident integrated dose.
Conducting periodic peel tests on the test specimens provides assurance that the adhesive has not degraded in the containment environment. The Frequencies of 18 months for the first two outages after fabrication of the
joint, followed by 18 months if the peel length is greater than 1/2" and
36 months if the peel length is less than or equal to 1/2" is based upon
the original vendor's recommendation which is based upon baseline
examination of the strength of the adhesive. Therefore, the Frequency
was concluded to be acceptable from a reliability standpoint.
Divider Barrier Integrity B 3.6.13 BASES Watts Bar - Unit 2 B 3.6-84 (developmental)
A SURVEILLANCE REQUIREMENTS
Visual inspection of the seal around the perimeter provides assurance
that the seal is properly secured in place. The Frequency of 18 months
was developed considering such factors as the inaccessibility of the seals
and absence of traffic in their vicinity, the strength of the bolts and
mechanisms used to secure the seal, and the plant conditions needed to
perform the SR. Operating experience has shown that these components
usually pass the Surveillance when performed at the 18 month
Frequency. Therefore, the Frequency was concluded to be acceptable
from a reliability standpoint.
REFERENCES
- 1. Watts Bar FSAR, Section 6.2, "Containment Systems."
Containment Recirculation Drains B 3.6.14 (continued)
Watts Bar - Unit 2 B 3.6-85 (developmental)
A B 3.6 CONTAINMENT SYSTEMS B 3.6.14 Containment Recirculation Drains
BASES BACKGROUND The containment recirculation drai ns consist of the ice condenser drains and the refueling canal drains. The ice condenser is partitioned into
24 bays, each having a pair of inlet doors that open from the bottom
plenum to allow the hot steam-air mixture from a Design Basis Accident (DBA) to enter the ice condenser. Twenty of the 24 bays have an ice
condenser floor drain at the bottom to drain the melted ice into the lower
compartment (in the 4 bays that do not have drains, the water drains
through the floor drains in the adjacent bays). Each drain leads to a drain
pipe that drops down several feet, then makes one or more 90 bends and exits into the lower compartment. A check (flapper) gate at the end of each pipe keeps warm air from entering during normal operation, but when the water exerts pressure, it opens to allow the water to spill into
the lower compartment. This prevents water from backing up and
interfering with the ice condenser inlet doors. The water delivered to the
lower containment serves to cool the atmosphere as it falls through to the
floor and provides a source of borated water at the containment sump for
long term use by the Emergency Core Cooling System (ECCS) and the
Containment Spray System during the recirculation mode of operation.
The two refueling canal drains are at low points in the refueling canal.
During a refueling, plugs are installed in the drains and the canal is
flooded to facilitate the refueling process. The water acts to shield and
cool the spent fuel as it is transferred from the reactor vessel to storage.
After refueling, the canal is drained and the plugs removed. In the event
of a DBA, the refueling canal drains are the main return path to the lower
compartment for Containment Spray Sy stem water sprayed into the upper compartment.
The ice condenser drains and the refueling canal drains function with the
ice bed, the Containment Spray System, and the ECCS to limit the
pressure and temperature that could be expected following a DBA.
Containment Recirculation Drains B 3.6.14 BASES (continued)
(continued)
Watts Bar - Unit 2 B 3.6-86 (developmental)
A APPLICABLE SAFETY ANALYSES The limiting DBAs considered relative to containment pressure and
temperature are the loss of coolant accident (LOCA) and the steam line
break (SLB) respectively. The LOCA and SLB are analyzed using
computer codes designed to predict the resultant containment pressure
and temperature transients. DBAs are assumed not to occur
simultaneously or consecutively. Although the ice condenser is a passive
system that requires no electrical power to perform its function, the
Containment Spray System and the Air Return System (ARS) also
function to assist the ice bed in limiting pressures and temperatures.
Therefore, the analysis of the postulated DBAs, with respect to
Engineered Safety Feature (ESF) systems, assumes the loss of one ESF
bus, which is the worst case single active failure and results in one train
of the Containment Spray System and one train of the ARS being
rendered inoperable.
The limiting DBA analyses (Ref. 1) show that the maximum peak containment pressure results from the LOCA analysis and is calculated to
be less than the containment design pressure. The maximum peak
containment atmosphere temperature results from the SLB analysis and
is discussed in the Bases for LCO 3.6.5, "Containment Air Temperature."
In addition to calculating the overall peak containment pressures, the
DBA analyses include calculation of the transient differential pressures
that occur across subcompartment walls during the initial blowdown
phase of the accident transient. The internal containment walls and
structures are designed to withstand these local transient pressure
differentials for the limiting DBAs.
The containment recirculation drains satisfy Criterion 3 of the NRC Policy
Statement.
LCO This LCO establishes the minimum requirements to ensure that the containment recirculation drains perform their safety functions. The ice
condenser floor drain valve gates must be closed to minimize air leakage
into and out of the ice condenser during normal operation and must open
in the event of a DBA when water begins to drain out. The refueling canal
drains must have their plugs removed and remain clear to ensure the
return of Containment Spray System water to the lower containment in
the event of a DBA. The containment recirculation drains function with
the ice condenser, ECCS, and Containment Spray System to limit the
pressure and temperature that could be expected following a DBA.
Containment Recirculation Drains B 3.6.14 BASES (continued)
(continued)
Watts Bar - Unit 2 B 3.6-87 (developmental)
A APPLICABILITY In MODES 1, 2, 3, and 4, a DBA could cause an increase in containment pressure and temperature, which would require the operation of the containment recirculation drains. Therefore, the LCO is applicable in
MODES 1, 2, 3, and 4.
The probability and consequences of these events in MODES 5 and 6 are
low due to the pressure and temperature limitations of these MODES. As
such, the containment recirculation drains are not required to be
OPERABLE in these MODES.
ACTIONS A.1 If one ice condenser floor drain is inoperable, 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> is allowed to restore
the drain to OPERABLE status. The Required Action is necessary to
return operation to within the bounds of the containment analysis. The
1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Completion Time is consistent with the ACTIONS of LCO 3.6.1, "Containment," which requires that containment be restored to
OPERABLE status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
B.1 If one refueling canal drain is inoperable, 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> is allowed to restore the
drain to OPERABLE status. The Required Action is necessary to return
operation to within the bounds of the containment analysis. The 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />
Completion Time is consistent with the ACTIONS of LCO 3.6.1, which
requires that containment be restored to OPERABLE status in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
C.1 and C.2
If the affected drain(s) 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 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to MODE 5 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.
Containment Recirculation Drains B 3.6.14 BASES (continued)
Watts Bar - Unit 2 B 3.6-88 (developmental)
A SURVEILLANCE REQUIREMENTS SR 3.6.14.1
Verifying the OPERABILITY of the refueling canal drains ensures that
they will be able to perform their functions in the event of a DBA. This
Surveillance confirms that the refueling canal drain plugs have been
removed and that the drains are clear of any obstructions that could
impair their functioning. In addition to debris near the drains, attention
must be given to any debris that is located where it could be moved to the
drains in the event that the Containment Spray System is in operation and
water is flowing to the drains. SR 3.6.14.1 must be performed before
entering MODE 4 from MODE 5 after every filling of the canal to ensure that the plugs have been removed and that no debris that could impair the drains was deposited during the time the canal was filled. The 92 day
Frequency was developed considering such factors as the inaccessibility
of the drains, the absence of traffic in the vicinity of the drains, and the
redundancy of the drains.
Verifying the OPERABILITY of the ice condenser floor drains ensures that
they will be able to perform their functions in the event of a DBA.
Inspecting the drain valve gate ensures that the gate is performing its
function of sealing the drain line from warm air leakage into the ice
condenser during normal operation, yet will open if melted ice fills the line
following a DBA. Verifying that the drain lines are not obstructed ensures
their readiness to drain water from the ice condenser. The 18 month
Frequency was developed considering such factors as the inaccessibility
of the drains during power operation; the design of the ice condenser, which precludes melting and refreezing of the ice; and operating experience that has confirmed that the drains are found to be acceptable
when the Surveillance is performed at an 18 month Frequency. Because
of high radiation in the vicinity of the drains during power operation, this
Surveillance is normally done during a shutdown.
REFERENCES
- 1. Watts Bar FSAR, Section 6.2, "Containment Systems."
Shield Building B 3.6.15 (continued)
Watts Bar - Unit 2 B 3.6-89 (developmental)
A B 3.6 CONTAINMENT SYSTEMS B 3.6.15 Shield Building
BASES BACKGROUND The shield building is a concrete structure that surrounds the steel containment vessel. Between the containment vessel and the shield
building inner wall is an annular space that collects containment leakage
that may occur following a loss of coolant accident (LOCA) as well as
other design basis accidents (DBAs) that release radioactive material.
This space also allows for periodic inspection of the outer surface of the
steel containment vessel.
The Emergency Gas Treatment System (EGTS) establishes a negative
pressure in the annulus between the shield building and the steel
containment vessel. Filters in the system then control the release of
radioactive contaminants to the environment. The shield building is
required to be OPERABLE to ensure retention of containment leakage
and proper operation of the EGTS.
APPLICABLE
SAFETY ANALYSES The design basis for shield building OPERABILITY is a LOCA.
Maintaining shield building OPERABILITY ensures that the release of
radioactive material from the containment atmosphere is restricted to
those leakage paths and associated leakage rates assumed in the
accident analyses.
The shield building satisfies Criterion 3 of the NRC Policy Statement.
LCO Shield building OPERABILITY must be maintained to ensure proper operation of the EGTS and to limit radioactive leakage from the
containment to those paths and leakage rates assumed in the accident
analyses.
Shield Building B 3.6.15 BASES (continued)
(continued)
Watts Bar - Unit 2 B 3.6-90 (developmental)
A APPLICABILITY Maintaining shield building OPERABILITY prevents leakage of radioactive material from the shield building. Radioactive material may enter the shield building from the containment following a DBA. Therefore, shield
building OPERABILITY is required in MODES 1, 2, 3, and 4 when DBAs
could release radioactive material to the containment atmosphere.
In MODES 5 and 6, the probability and consequences of these events are
low due to the Reactor Coolant System temperature and pressure
limitations in these MODES. Therefore, shield building OPERABILITY is
not required in MODE 5 or 6.
ACTIONS A.1
In the event shield building OPERABILITY is not maintained, shield
building OPERABILITY must be restored within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is a
reasonable Completion Time considering the limited leakage design of
containment and the low probability of a Design Basis Accident occurring
during this time period.
B.1 The Completion Time of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> is based on engineering judgment. The
normal alignment for both EGTS control loops is the A-Auto position.
With both EGTS control loops in A-Auto, both trains will function upon
initiation of a Containment Isolation Phase A (CIA) signal. In the event of
a LOCA, the annulus vacuum control system isolates and both trains of
the EGTS pressure control loops will be placed in service to maintain the
required negative pressure. If annulus vacuum is lost during normal
operations, the A-Auto position is unaffected by the loss of vacuum. This
operational configuration is acceptable because the accident dose
analysis conservatively assumes the annulus is at atmospheric pressure
at event initiation. A Note has been provided which makes the
requirement to maintain the annulus pressure within limits not applicable
during venting operations, required annulus entries, or Auxiliary Building
isolations not exceeding 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> in duration.
Shield Building B 3.6.15 BASES (continued)
Watts Bar - Unit 2 B 3.6-91 (developmental)
B ACTIONS (continued)
C.1 and C.2 If the shield building 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 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to MODE 5 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 SR 3.6.15.1
Verifying that shield building annulus negative pressure is within limit (equal to or more negative than -5 inches water gauge; value does not
account for instrument error) ensures that operation remains within the limit assumed in the containment analysis. The 12-hour Frequency of this
SR was developed considering operating experience related to shield
building annulus pressure variations and pressure instrument drift during
the applicable MODES.
Maintaining shield building OPERABILITY requires maintaining each door
in the access opening closed, except when the access opening is being
used for normal transient entry and exit. The 31-day Frequency of this
SR is based on engineering judgment and is considered adequate in view
of the other indications of door status that are available to the operator.
This SR would give advance indication of gross deterioration of the
concrete structural integrity of the shield building. The Frequency of this
SR is the same as that of SR 3.6.1.1. The verification is done during
shutdown.
Shield Building B 3.6.15 BASES Watts Bar - Unit 2 B 3.6-92 (developmental)
B SURVEILLANCE REQUIREMENTS (continued)
The EGTS is required to maintain a pressure equal to or more negative
than -0.50 inches water gauge ("wg) in the annulus at an elevation
equivalent to the top of the Auxiliary Building. At elevations higher than
the Auxiliary Building, the EGTS is required to maintain a pressure equal
to or more negative than -0.25 "wg. The low pressure sense line for the
pressure controller is located in the annulus at elevation 783. By verifying
that the annulus pressure is equal to or more negative than -0.61 "wg at
elevation 783, the annulus pressurization requirements stated above are
met. The ability of a EGTS train with final flow 3600 cfm and 4400 cfm to produce the required negative pressure during the test operation provides assurance that the building is adequately sealed. The
negative pressure prevents leakage from the building, since outside air
will be drawn in by the low pressure at a maximum rate 250 cfm. The 18 month Frequency on a STAGGERED TEST BASIS is consistent with
Regulatory Guide 1.52 (Ref. 1) guidance for functional testing.
REFERENCES 1. Regulatory Guide 1.52, Revision 2, "Design, Testing and Maintenance Criteria for Post Accident Engineered-Safety-Feature
Atmospheric Cleanup System Air Filtration and Adsorption Units of
Light-Water Cooled Nuclear Power Plants."