ML20056F490
| ML20056F490 | |
| Person / Time | |
|---|---|
| Site: | 05200001 |
| Issue date: | 08/10/1993 |
| From: | Crutchfield D Office of Nuclear Reactor Regulation |
| To: | Marriot P NORTHEAST NUCLEAR ENERGY CO. |
| References | |
| NUDOCS 9308270317 | |
| Download: ML20056F490 (120) | |
Text
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UNITED STATES NUCLEAR REGULATORY COMMISSION Ef..g y
W ASWNGT ON, D.C. 20555-m01
.e August 10, 1993 Docket No.52-001 Mr. Patrick W. Marriott, Manager Licensing & Consulting Services GE Nuclear Energy 175 Curtner Avenue San Jose, California 95125
Dear Mr. Harriott:
SUBJECT:
ADVANCED BOILING WATER REACTOR (ABWR) TECHNICAL SPECIFICATIONS Enclosed are the proof and review ABWR technical specifications and their bases for the following sections:
3.6 Containment Systems 3.7 Plant Systems As we discussed at our management meeting on June 10, 1993, the Nuclear Regulatory Commission (NRC) staff will be providing GE Nuclear Energy (GE),
until August 31, selected sections of proof and review ABWR technical specifi-cations. These sections are based on the NRC staff review of the GE mark-up of the BWR-6 and BWR-4 Standard Technical Specifications; the sections, as provided, are acceptable to the NRC staff. As discussed, we anticipate that GE will interface very closely with the staff to resolve any issues on these sections prior to August 31, 1993.
Under this arrangement, we anticipate that formal comments to proof and review ABWR technical specifications made by September 20, 1993, will be few.
The electronic text of these sections is available on the NRC Technical Specifications Branch electronic bulletin board (OTSB-BBS) in Wordperfect 5.1 format. The data telephone number for the OTSB-BBS is (301) 504-1778, and the system operator is Tom Dunning who is available for assistance at (301) 504-1189. Also, in accordance with our agreements, GE will maintain these sections in Wordperfect 5.1 format and will produce subsequent issues of the ABWR technical specifications in Wordperfect 5.1 format.
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Mr. Patrick W. Marriott August 10, 1993 If you have any questions about technical specifications please contact Mark Reinhart with the Nuclear Reactor Regulation Technical Specifications Branch.
He may be reached at (301) 504-11B5.
Sincerely, 9tiginal Signed By Dennis M. Crutchfield, Associate Director for Advanced Reactors and License Renewal Office of Nuclear Reactor Regulation
Enclosure:
As stated cc w/ enclosure:
See next page DISTRIBUTION w/ enclosure:
\\ Docket File CPoslusny PShea FMReinhart PDR PDST R/F DORS R/F OTSB R/F DISTRIBUTION w/o enclosure:
TEMurley/FJMiraglia WTRussell JGPartlow DMCrutchfield BKGrimes CIGrimes BABoger JWiggins ACThadani FJCongel JSWermiel RBBarrett CEMcCracken RCJones CEBerlinger AEthaffee GHMarcus WBHardin, RES LCShao, RES JA0'Brien RWBorchardt JNWilson SNinh SMLMagruder TGody, 17G21 JEMoore, 15B18 RHLo PCHearn ACRS (11)
- See previous concurrence OFC:
LA:PDST:ADAR OTSB SC:PTSB:
C:OTSB NAME:
PSh PCHearn*
FMReinhart*
CIGrirnes
- DATE:
08/
9 08/06/93 08/06/93 08/09/93 OFC:
(A)SC:PDST:A-(A)D:PDST:ADAR AA :NRR NAME:
CPostusny:sD
- JNWilson*
DMCrutchfield DATE:
08/09/93 08/09/93 08/d!93 OFFICIAL RECORD COPY:
Mr. Patrick W. Marriott August 10, 1993 If you have any questions about technical specifications please contact Mark Reinhart with the Nuclear Reactor Regulation Technical Specifications Branch.
He may be reached at (301) 504-1185.
Sincerely,
/,
/ISW ennis M. Crutchfi cia e irector for Advanced Reactors ind License Renewal Office of Nuclear Reactor Regulation
Enclosure:
As stated cc w/ enclosure:
See next page
Mr. Patrick W. Harriott Docket No.52-001 General Electric Company cc: Mr. Robert Mitchell Mr. Joseph Quirk General Electric Company GE Nuclear Energy 175 Curtner Avenue General Electric Company San Jose, California 95125 175 Curtner Avenue, Mail Code 782 San Jose, California 95125 Mr. L. Gifford, Program Manager Regulatory Programs GE Nuclear Energy 12300 Twinbrook Parkway Suite 315 Rockville, Maryland 20852 Director, Criteria & Standards Division Office of Radiation Programs U. S. Environmental Protection Agency 401 M Street, S.W.
Washington, D.C.
20460 Mr. Sterling Franks U. S. Department of Energy NE-42 Washington, D.C.
20585 l
Marcus A. Rowden, Esq.
Fried, Frank, Harris, Shriver & Jacobson 2001 Pennsylvania Avenue, N.W.
Suite 800 Washington, D.C.
20004 Jay M. Gutierrez, Esq.
Newman & Holtzinger, P.C.
1615 L Street, N.W.
Suite 1000 Washington, D.C.
20036 I
i i
m 3
Primary Containment B 3.6.1.1 B 3.6 CONTAINMENT SYSTEMS B 3.6.1.1 Primary Containment BASES BACKGROUND The function of the primary containment is to isolate and contain fission products released from the Reactor Primary System following a Design Basis Accident (DBA) and to confine the postulated release of radioactive material. The primary containment consists of a steel lined, reinforced concrete vessel, which surrounds the Reactor Primary System and provides an essentially leak tight barrier against an uncontrolled release of radioactive material to the environment.
The isolation devices for the penetrations in the primary 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 LC0 3.6.1.3, " Primary Containment Isolation Valves (PCIVs)";
b.
The primary containment air lock is OPERABLE, except as provided in LCO 3.6.1.2, " Primary Containment Air Lock";
c.
The sealing mechanism associated with a penetration (e.g., welds, bellows, or o-rings) is OPERABLE.
This Specification ensures that the performance of the primary containment, in the event of a DBA, meets the assumptions used in the safety analyses of References 1 and 2.
SR 3.6.1.1.1 leakage rate requirements are in (continued)
ABWR TS B 3.6-1 P&R, 08/03/93 4:15pm
O Primary Containment B 3.6.1.1 BASES BACKGROUND conformance with 10 CFR 50, Appendix J (Ref. 3), as modified (continued) by approved exemptions.
APPLICABLE The safety design basis for the primary containment is that SAFETY ANALYSES it must withstand the pressures and temperatures of the limiting DBA without exceeding the design leakage rate.
The DBA that postulates the maximum release of radioactive material within primary containment is a LOCA.
In the analysis of this accident, it is assumed that primary containment is OPERABLE such that release of fission products to the environment is controlled by the rate of primary containment leakage.
Analytical methods and assumptions involving the primary containment are presented in References I and 2.
The safety analyses assume a nonmechanistic fission product release following a DBA, which forms the basis for determination of offsite doses. The fission product release is, in turn, based on an assumed '.aakage rate from the primary containment.
OPERABILITY of the primary containment ensures that the leakage rate assumed in the safety analyses is not exceeded.
The maximum allowable leakage rate for the primary containment (L ) is [0.5]% by weight of the containment air per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> al the maximum peak containment pressure (P,)
2 of 2.75Kg/cm g (39 psig) or [
]% by weight of the containmentajrper24hoursatthereducedpressureofPt of [ ] Kg/cm g ([
] psig) (Ref. 1).
Primary containment satisfies Criterion 3 of the NRC Policy Statement.
LCO Primary containment OPERABILITY is maintained by limiting leakage to within the acceptance criteria of 10 CFR 50, Appendix J (Ref. 3).
Compliance with this LC0 will ensure a primary containment configuration, including equipment hatches, that is structurally sound and that will limit leakage to those leakage rates assumed in the safety analyses.
Individual leakage rates specified for the primary containment air lock are addressed in LCO 3.6.1.2.
(continued)
ABWR TS B 3.6-2 P&R, 08/03/93 4:15pm
I Primary Containment B 3.6.1.1 BASES (continued)
APPLICABILITY In MODES 1, 2, and 3, a DBA could cause a release of radioactive material to primary containment.
In MODES 4 and 5, the probability and consequences of these events are reduced due to the pressure and temperature limitations of these MODES.
Therefore, primary containment is not required to be OPERABLE in MODES 4 and 5 to prevent leakage of radioactive material from primary containment.
ACTIONS Al z
In the event primary containment is inoperable, primary containment must be restored to OPERABLE status within I hour.
The I hour Completion Time provides a period of time to correct the problem commensurate with the importance of maintaining primary containment OPERABILITY during MODES 1, 2, and 3.
This time period also ensures that the probability of an accident (requiring primary containment OPERABILITY) occurring during periods where primary containment is inoperable is minimal.
B.1 and B,2 If primary containment cannot be restored to OPERABLE status within the required Completion Time, the plant must be brought to a MODE in which the LC0 does not apply.
To achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and to MODE 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
SURVEILLANCE SR 3.6.1.1.I REQUIREMENTS Maintaining the primary containment OPERABLE requires compliance with the visual examinations and leakage rate test requirements of 10 CFR 50, Appendix J (Ref. 3), as modified by approved exemptions.
Failure to meet air lock leakage testing (SR 3.6.1.2.1), [ resilient seal primary containment purge valve leakage testing (SR 3.6.1.3.7),]
(continued)
ABWR TS B 3.6-3 P&R, 08/03/93 5:05pm
I Primary Containment B 3.6.1.1 BASES SURVEILLANCE SR 3.6.1.1.1 (continued)
REQUIREMENTS main steam isolation valve leakage (SR 3.6.1.3.12), or hydrostatically tested valve leakage (SR 3.6.1.3.13) does not necessarily result in a failure of this SR.
The impact of the failure to meet these SRs must be evaluated against the Type A, B, and C acceptance criteria of 10 CFR 50, Appendix J.
The Frequency is required by 10 CFR 50, Appendix J (Ref. 3), as modified by approved exemptions.
Thus, SR 3.0.2 (which allows Frequency extensions) does not apply.
SR 3.6.1.1.2 Maintaining the pressure suppression function of primary containment requires limiting the leakage from the drywell to the suppression chamber.
Thus, if an event were to occur that pressurized the drywell, the steam would be directed through the horizontal vents into the suppression pool.
This SR measures drywell to suppression chamber differential pressure during a 10 minute period to ensure that the leakage paths that would bypass the suppression pool are within allowable limits.
Satisfactory performance of this SR can be achieved by establishing a known differential pressure between the drywell and the suppression chamber and verifying that the pressure in either the suppression chamber or the drywell does not change by more than 6 mm (0.25 inch) of water per minute over a 10 minute period. The leakage test is performed every 18 months. The 18 month Frequency was developed considering it is prudent that this Surveillance be performed during a unit outage and also in view of the fact that component failures that might have affected this test are identified by other primary containment SRs.
Two consecutive test failures, however, would indicate unexpected primary containment degradation; in this event, as the Note indicates, increasing the Frequency to once every 9 months is required until the situation is remedied as evidenced by passing two consecutive tests.
(continued)
ABWR TS B 3.6-4 P&R, 08/03/93 4:27pm
I Primary Containment B 3.6.1.1 BASES REFERENCES 1.
2.
3.
ABWR TS B 3.6-5 P&R, 08/03/93 4:27pm
E Primary Containment Air Locks B 3.6.1.2 B 3.6 CONTAINMENT SYSTEMS B 3.6.1.2 Primary Containment Air Locks i
BASES BACKGROUND One double door primary containment air lock has been built into the primary containment to provide personnel access to the drywell and to provide primary containment isolation during the process of personnel entering and exiting the d rywell. The air lock is designed to withstand the same loads, temperatures, and peak design internal and external pressures as the primary containment (Ref.1). As part of the primary containment, the air lock limits the release of radioactive material to the environment during normal unit operation and through a range of transients and accidents up to and including postulated Design Basis Accidents (DBAs).
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 DBA in primary containment.
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 primary containment internal pressure results in increased sealing force on each door.)
Each air lock is nominally a right circular cylinder,10 ft in diameter, with doors at each end that are interlocked to prevent simultaneous opening. The air lock is provided with limit switches on both doors that provide control room indication of door position.
[ Additionally, control room indication is provided to alert the operator whenever an air lock interlock mechanism is defeated.] During periods when primary containment is not required to be OPERABLE, the air lock interlock mechanism may be disabled, allowing both doors of an air lock to remain open for extended periods when frequent primary containment entry is necessary.
Under some conditions, as allowed by this LCO, the primary containment may be accessed through the air lock when the door interlock mechanism has failed, by manually performing the interlock function.
(continued)
ABWR TS B 3.6-1 P&R, 08/04/93 8:36am
t j
Primary Containment Air Locks B 3.6.1.2 l
l BASES l
BACKGROUND The primary containment air lock forms part of the primary (continued) containment pressure boundary.
As such, air lock integrity and leak tightness are essential for maintaining primary containment leakage rate to within limits 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 unit safety analysis.
SR 3.6.1.1.1 leakage rate requirements conform with 10 CFR 50, Appendix J (Ref. 2), as modified by approved exemptions.
a APPLICABLE The DBA that postulates the maximum release of radioactive SAFETY ANALYSES material within primary containment is a LOCA.
In the analysis of this accident, it is assumed that primary containment is OPERABLE, such that release of fission products to the environment is controlled by the rate of primary containment leakage. The primary containment is designed with a maximum allowable leakage rate (L ) of 0.5%
byweightofthecontainmentairper24hoursatIhe calculated maximum peak containment pressure (P ) of 2.74 2
Kg/cm g (39 psig) Ref. 3).
This allowable leaEage rate forms the basis for the acceptance criteria imposed on the SRs associated with the air lock.
Primary containment air lock OPERABILITY is also required to minimize the amount of fission product gases that may escape primary containment through the air lock and contaminate and pressurize the secondary containment.
The primary containment air lock satisfies Criterion 3 of the NRC Policy Statement.
LCO As part of the primary containment, the air lock's safety function is related to control of containment leakage rates following a DBA.
Thus, the air lock's structural integrity and leak tightness are essential to the successful mitigation of such an event.
The primary containment air locks are required to be OPERABLE.
For each air lock to be considered OPEPABLE, 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 (continued)
ABWR TS B 3.6-2 P&R, 08/04/93 8:36am
L j
Primary Containment Air Locks B 3.6.1.2 BASES LCO allows only one air lock door to be opened at a time.
This (continued) provision ensures that a gross breach of primary containment does not exist when primary containment is required to be OPERABLE. Closure of a single door in each air lock is sufficient to provide a leak tight barrier following i
postulated events.
Nevertheless, both doors are kept closed when the air lock is not being used for normal entry into and exit from primary containment.
APPLICABILITY In MODES 1, 2, and 3, a DBA could cause a release of radioactive material to primary containment.
In MODES 4 and 5, the probability and consequences of these events are reduced due to the pressure and temperature limitations of these MODES. Therefore, the primary containment air lock is not required to be OPERABLE in MODES 4 and 5 to prevent i
leakage of radioactive material from primary containment.
ACTIONS The ACTIONS are modified by Note 1, which allows entry and exit to perform repairs of the affected air lock component.
If the outer door is inoperable, then it may be easily accessed to repair.
If the inner door is the one that is inoperable, however, then it is preferred that the air lock be accessed from inside primary containment by entering through the other OPERABLE air lock.
If this is not practical, however, then it is permissible to enter the air lock through the OPERABLE outer door, which means there is a short time during which the primary containment boundary is not intact (during access through the outer door).
The ability to open the OPERABLE door, even if it means the primary containment boundary is temporarily not intact, is acceptable due to the low probability of an event that could pressurize the primary containment during the short time in which the OPERABLE door is expected to be open.
The OPERABLE door must be immediately closed after each entry and exit.
Note 2 has been included to provide clarification that, for this LCO, separate Condition entry is allowed for each air lock.
The ACTIONS are modified by a third Note, which ensures appropriate remedial measures are taken when necessary.
(continued)
ABWR TS B 3.6-3 P&R, 08/04/93 8:36am l
s Primary Containment Air Locks B 3.6.1.2 BASES ACTIONS Pursuant to LCO 3.0.6, actions are not required, even if (continued) primary containment is exceeding its leakage limit.
Therefore, the Note is added to require ACTIONS for LC0 3.6.1.1, " Primary Containment," to be taken in this event.
A.I. A.2. and A.3 With one primary containment air lock door inoperable in one or more primary containment air locks, the OPERABLE door must be verified closed (Required Action A.1) in each-affected air lock. This ensures that a leak tight primary containment barrier is maintained by the use of an OPERABLE i
air lock door. This action must be completed within I hour.
The I hour Completion Time is consistent with the ACTIONS of LCO 3.6.1.1, which requires that primary containment be t
restored to OPERABLE status within I hour.
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 considered reasonable for locking the OPERABLE air lock door, considering that the OPERABLE door of the affected air lock is being maintained closed.
Required Action A.3 ensures that the affected air lock with an inoperable door has been isolated by the use of a locked closed OPERABLE air lock door. This ensures that an acceptable primary 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 or areas with limited access due to inerting and allows these doors to be verified locked closed by use of administrative controls.
Allowing verification by administrative controls 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 prope" position, is small.
(continued)
ABWR TS B 3.6-4 P&R, 08/04/93 8:36am
1 Primary Containment Air Locks B 3.6.1.2 BASES ACTIONS A.I. A.2. and A.3 (continued) l 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 air lock are inoperable. With both doors in the 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 use of the air lock for entry and exit for 7 days under administrative contro's.
Primary containment entry may be required to perform Technical Specifications (TS) Surveillances and Required Actions, as well as other activities on equipment inside primary containment that are required by TS or activities on equipment that support TS-required equipment.
This Note is I
not intended to preclude performing other activities (i.e.,
non-TS-related activities) if the primary containment was 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 primary containment during the short time that the OPERABLE door is expected to be open.
B.I. B.2. and B.3 With an air lock interlock mechanism inoperable in one or both primary containment 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 one air lock are inoperable. With both doors in the 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 the primary 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).
nequired Action B.3 is modified by a Note that applies to air lock doors located in high radiation areas or areas with (continued)
ABWR TS B 3.6-5 P&R, 08/04/93 8:39am
e Primary Containment Air Locks B 3.6.1.2 BASES ACTIONS B.l. B.2. and B.3 { continued) limited access due to inerting and that allows these doors to be verified locked closed by use of administrative controls. Allowing verification by administrative controls 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.I. C.2. and C.3 l
With one or more air lecks inoperable for reasons other than those described in Condition A or B, Required Action C.1 requires action to be immediately initiated to evaluate containment overall leakage rates using current air lock leakage test results. An evaluation is acceptable since it is overly conservative to immediately declare the primary 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) primary containment remains OPERABLE, yet only I hour (according to LC0 3.6.1.1) 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 primary containment air locks must be verified closed.
This action must be completed within the I hour Completion Time.
This specified time period is consistent with the ACTIONS of LC0 3.6.1.1, which require that primary containment be f
restored to OPEPABLE status within I hour.
i Additionally, the air lock must be restored to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. 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 restoring an inoperable air lock to OPERABLE status considering that at least one door is maintained closed in each affected air lock.
(continued)
ABWR TS B 3.6-6 P&R, 08/04/93 8:39am
o Primary Containment Air Locks B 3.6.1.2 BASES ACTIONS D.1 and D.2 (continued)
If the inoperable primary containment air lock cannot be restored to OPERABLE status within the associated Completion Time, the plant must be brought to a MODE in which the LC0 does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and to MODE 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
SURVEILLANCE SR 3.6.1.2.1 REQUIREMENTS Maintaining primary containment air locks OPERABLE requires compliance with the leakage rate test requirements of 10 CFR 50, Appendix J (Ref. 2), as modified by approved exemptions. This SR reflects the leakage rate testing requirements with respect to air lock leakage (Type B leakage tests). The acceptance criteria were established
[during initial air lock and primary containment OPERABILITY testing]. Tne periodic testing requirements verify that the air lock leakage does not exceed the allowed fraction of the overall primary containment leakage rate. The Frequency is required by 10 CFR 50, Appendix J (Ref. 2), as modified by approved exemptions. Thus, SR 3.0.2 (which allows Frequency extensions) does not apply.
The SR has been modified by two Notes. Note I 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 has been added to this SR, requiring the results to be evaluated against the acceptance criteria of SR 3.6.1.1.1.
This ensures that air lock leakage is properly accounted for in determining the overall primary containment leakage rate.
(continued)
ABWR TS B 3.6-7 P&R, OB/04/93 8:36am
e Primary Containment Air Locks B 3.6.1.2 BASES (continued)
SURVEILLANCE SR 3.6.1.2.2 REQUIREMENTS (continued)
The air lock interlock mechanism is designed to prevent j
simultaneous opening of both doors in the air lock.
Since both the inner and outer doors of an air lock are designed to withstand the maximum expected post accident primary containment pressure, closure of either door will support primary containment OPERABILITY.
Thus, the interlock i
feature supports primary 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 a
and that simultaneous inner and outer door opening will not inadvertently occur.
Due to the purely mechanical nature of this interlock, and given that the interlock mechanism is only challenged when primary containment is entered, this test is only required to be performed upon entering primary containment, but is not required more frequently than i
184 days when primary containment is de-inerted. The 184 day Frequency is based on engineering judgment and is considered adequate in view of other administrative controls
[such as indications of interlock mechanism status, available to operations personnel).
REFERENCES 1.
}
2.
3.
t i
i i
1 i
i l
L ABWR TS B 3.6-8 P&R, 08/04/93 8:36am
PCIVs B 3.6.1.3 B 3.6 CONTAINMENT SYSTEMS B 3.6.1.3 Primary Containment Isolation Valves (PCIVs)
BASES BACKGROUND The function of the PCIVs, in combination with other accident mitigation systems, is to limit fission product release during and following postulated Design Basis Accidents (DBAs) to within limits.
Primary containment isolation within the time limits specified for those isolation valves designed to close automatically ensures that the release of radioactive material to the environment will be consistent with the assumptions used in the analyses for a DBA.
The OPERABILITY requirements for PCIVs help ensure that adequate primary containment leak tightness is maintained during and after an accident by minimizing potential leakage paths to the environment. Therefore, the OPERABILITY requirements provide assurance that primary containment leakage rates assumed in the safety analyses will not be exceeded. 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.
The primary containment purge lines are 550 mm (22 inches) in diameter; vent lines are 550 mm (22 inches) in diameter.
The 550 mm (22 inch) primary containment purge valves are normally maintained closed in MODES 1, 2, and 3 to ensure leak tightness. The isolation valves on the 550 mm (22 inch) vent lines have 50 mm (2 inch) bypass lines around them for use during normal reactor operation.
Two additional redundant excess flow isolating dampers are provided on the vent line upstream of the Standby Gas Treatment (SGT) System filter trains.
These isolation (continued)
ABWR TS B 3.6-1 P&R, 08/04/93 3:27pm
PCIVs B 3.6.1.3 BASES BACKGROUND damper.s, together with the PCIVs, will prevent high (continued) pressure from reaching the SGT System filter trains in the unlikely event of a loss of coolant accident (LOCA) during venting.
Closure of the excess flow isolation dampers will not prevent the SGT System from performing its design function (that is, to maintain a negative pressure in the secondary containment). To ensure that a vent path is available, a 50 mm (2 inch) bypass line is provide around the dampers.
l APPLICABLE The PCIVs LCO was derived from the requirements related SAFETY ANALYSES to the control of leakage from the primary containment during major accidents. This LCO is intended to ensure that i
primary containment leakage rates do not exceed the values assumed in the safety analyses.
As part of the primary containment boundary, PCIV OPERABILITY supports leak tightness of primary containment.
Therefore, the safety analysis of any event requiring isolation'of primary containment is applicable to this LCO.
The DBAs that result in a release of radioactive material within primary containment a LOCA and a main steam line break (MSLB).
In the analysis for each of these accidents, it is assumed that PCIVs are either closed or close within the required isolation times following event initiation.
This ensures that potential leakage paths to the environment through PCIVs (and primary containment purge valves) are minimized. Of the events analyzed in Reference 1, the MSLB is the most limiting event due to radiological consequences.
The closure time of the main steam isolation valves (MSIVs) is the most significant variable from a radiological standpoint. The MSIVs are required to close within 3 to 4.5 seconds; therefore, the 4.5 second closure time is assumed in the analysis. The safety analyses assume that the purge valves were closed at event initiation.
- Likewise, it is assumed that the primary containment is isolated such that release of fission products to the environment is controlled by the rate of primary containment leakage.
The DBA analysis assumes that within 60 seconds of the accident, isolation of the primary containment is complete and leakage is terminated, except for the maximum allowable leakage rate, L,.
The primary containment isolation total (continued)
o o
l PCIVs B 3.6.1.3 BASES (continued)
APPLICABLE response time of 60 seconds includes signal delay, diesel SAFETY ANALYSES generator startup (for loss of offsite power), and PCIV l
(continued) stroke times.
l The single failure criterion required to be imposed in the conduct of unit safety analyses was considered in the original design of the primary 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 primary containment purge valves may be unable to close in the environment following a LOCA. Therefore, each of the purge valves is required to remain sealed closed during MODES 1, 2, and 3.]
In this case, the single failure criterion remains applicable to the primary containment purge valve due to failure in the control circuit associated with each valve. Again, the primary containment purge valve design precludes a single failure from compromising primary containment OPERABILITY as long as the system is operated in accordance with this LCO.
j PCIVs satisfy Criterion 3 of the NRC Policy Statement.
l LC0 PCIVs form a part of the primary containment boundary. The PCIV safety function is related to control of primary l
containment leakage rates during a DBA.
The power operated, automatic isolation valves are required i
to have isolation times within limits and actuate on an automatic isolation signal. The 550 mm (22 inch) purge valves must be maintained sealed closed [or blocked to prevent full opening]. The valves covered by this LC0 are listed with their associated stroke times in Reference 2.
The normally closed 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 and devices are those listed in Reference 2.
Purge valves with resilient seals, secondary bypass valves, MSIVs, and hydrostatically tested valves must (continued)
ABWR TS B 3.6-3 P&R, 08/04/93 3:27pm
=
o PCIVs B 3.6.1.3 BASES LCO meet additional leakage rate requirements. Other PCIV (continued) leakage rates are addressed by LCO 3.6.1.1, " Primary Containment," as Type C testing.
This LC0 provides assurance that the PCIVs will perform their designed safety functions to control leakage from the primary containment during accidents.
APPLICABILITY In MODES 1, 2, and 3, a DBA could cause a release of radioactive material to primary containment.
In MODES 4 and 5, the probability and consequences of these events are reduced due to the pressure and temperature limitations of l
these MODES. Therefore, most PCIVs are not required to be OPERABLE [and the primary containment purge valves are not required to be sealed closed] in MODES 4 and 5.
Certain valves, however, are required to be OPERABLE to prevent inadvertent reactor vessel draindown.
These valves are those whose associated instrumentation is required to be OPERABLE per LC0 3.3.6.1, " Primary Containment Isolation Instrumentation."
(This does not include the valves that isolate the associated instrumentation.)
ACTIONS The ACTIONS are modified by a Note allowing penetration flow path (s) [except for purge valve flow path (s)] to be unisolated intermittently under administrative controls.
These controls consist of stationing a dedicated operator at the controls of the valve, who is in continuous communication with the control room.
In this way, the penetration can be rapidly isolated when a need for primary containment isolation is indicated.
Due to the size of the primary containment purge line penetratic. and the fact that those penetrations exhaust directly from the containment atmosphere to the environment, the penetration flow path containing these valves is not allowed to be opened under administrative controls.
A single purge valve in a penetration flow path may be opened to effect repairs to an inoperable valve, as allowed by SR 3.6.1.3.1.
A second Note has been added to provide clarification that, for the purpose of this LCO, separate Condition entry is allowed for each penetration flow path.
(continued)
ABWR TS B 3.6-4 P&R, 08/04/93 3:27pm
e PCIVs B 3.6.1.3 BASES (continued)
ACTIONS The ACTIONS are modified by a third Note, which ensures (continued) that appropriate remedial actions are taken, if necessary, if the affected system (s) are rendered inoperable by an I
inoperable PCIV (e.g., an Emergency Core Cooling Systems subsystem is inoperable due to a failed open test return valve).
1 Note 4 ensures appropriate remedial actions are taken when the primary containment leakage limits are exceeded.
Pursuant to LCO 3.0.6, these actions are not required even i
I when the associated LCO is not met. Therefore, Notes 3 l
and 4 are added to require that the proper actions are i
taken.
l A.1 and A.2 l
With one or more penetration flow paths with one PCIV inoperable [except for purge valve leakage not within limit], the affected penetration flow paths must be isolated. The method of isolation must include the use of at least one isolation barrier that cannot be adversely i
I affected by a single active failure.
Isolation barriers that meet this criterion are a closed and de-activated automatic valve, a closed manual valve, a blind flange, and a check valve with flow through the valve secured.
For i
penetration isolated in accordance with Required Action A.1, j
the valve used to isolate the penetration should be the closest available valve to the primary containment. The Required Action must be completed within the 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Completion Time (8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> for main steam lines).
The Completion Time of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is reasonable considering the time required to isolate the penetration and the relative importance of supporting primary containment OPERABILITY during MODES 1, 2, and 3.
For main steam lines, an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> Completion Time is allowed.
The Completion Time of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> for the main steam lines allows a period of time to restore the MSIVs to OPERABLE status given the fact that MSIV closure will result in isolation of the main steam line(s) and a potential for plant shutdown.
For affected penetrations that have been isolated in accordance with Required Action A.1, the affected penetration flow path (s) must be verified to be isolated on a periodic basis. This is necessary to ensure that primary containment penetrations required to be isolated following (continued)
ABWR TS B 3.6-5 P&R, 08/04/93 3:27pm
e PCIVs B 3.6.1.3 BASES ACTIONS A.1 and A.2 (continued) 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 valve manipulation.
Rather, it involves verification, through a system walkdown, that those valves outside containment and capable of potentially being mispositioned are in the correct position. The Completion Time of "once 31 days for isolation devices outside primary containment,"
is appropriate because the valves are operated under administrative controls and the probability of their misalignment is low.
For valves inside primary containment, the time period " prior to entering MODE 2 or 3 from MODE 4, if primary containment was de-inerted while in MODE 4, if ggg not performed within the previous 92 days," is based on engineering judgment and is considered reasorable in view of i
the inaccessibility of the valves and other administrative controls ensuring that valve misalignment is an unlikely possibility.
Condition A is modified by a Note indicating that this Condition is only applicable to those penetration flow paths with two PCIVs.
For penetration flow paths with one PCIV, Condition C provides the appropriate Required Actions.
Required Action A.2 is modified by a Note that applies to valves and blind flanges located in high radiation areas, and allows them to be verified by use of administrative means. Allowing verification by administrative means is considered acceptable, since access to thase areas is typically restricted.
Therefore, the probability of misalignment of these valves, once they have been verified to be in the proper position, is low.
161 With one or more penetration flow paths with two PCIVs inoperable, either the inoperable PCIVs must be restored to OPERABLE status or 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 (continued)
ABWR TS B 3.6-6 P&R, 08/04/93 3:27pm
O O
PCIVs B 3.6.1.3 BASES t
L ACTIONS 161 (continued)
[
de-activated automatic valve, a closed manual valve, and a blind flinge. The I hour Completion Time is consistent with the ACTIONS of LC0 3.6.1.1.
Condition B is modified by a Note indicating this Condition is only applicable to penetration flow paths with two PCIVs.
For penetration flow paths with one PCIV, Condition C provides the appropriate Required Actions.
C.] and C.2 With one or more penetration flow paths with one PCIV inoperable, the inoperable valve 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.
i A check valve may not be used to isolate the affected penetration.
Required Action C.1 must be completed within i
l the 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Completion Time. The Completion Time of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is reasonable considering the relative stability of the i
closed system (hence, reliability) to act as a penetration isolation boundary and the relative importance of supporting primary containment OPERABILITY during MODES 1, 2, and 3.
The Completion Time of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is reasonable considering the instrument and the small pipe diameter of penetration (hence, reliability) to act as a penetration isolation boundary and the small pipe diameter of the affected penetrations.
In the event the affected penetration flow t
path is isolated in accordance with Required Action C.1, the i
affected penetration must be verified to be isolated on a j
periodic basis. This is necessary to ensure that primary j
containment penetrations required to be isolated following an accident are isolated.
The Completion Time of once per 31 days for verifying each affected penetration is isolated is appropriate because the valves are operated under administrative controls and the probability of their misalignment is low.
(continued)
ABWR TS B 3.6-7 P&R, 08/04/93 3:27pm 9
W PCIVs B 3.6.1.3 BASES ACTIONS C.1 and C.2 (continued)
Condition C is modified by a Note indicating that this i
Condition is only applicable to penetration flow paths with only one PCIV.
For penetration flow paths with two PCIVs, Conditions A and B provide the appropriate Required Actions.
t Required Action C.2 is modified by a Note that applies to valves and blind flanges located in high radiation areas and l
allows them to be verified by use of administrative means.
l Allowing verification by administrative means is considered acceptable, since access to these areas is typically restricted. Therefore, the probability of misalignment of these valves, once they have been verified to be in the l
proper position, is low.
D.I. D.2. and D.3 In the event one or more containment purge valves are not within the purge valve leakage limits, purge valve leakage must be restored to within limits or the affected penetration 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, and blind fl ange]. A purge valve with resilient seals utilized to satisfy Required Action D.1 must have been demonstrated to meet the leakage requirements of SR 3.6.1.3.7.
The specified Completion Time is reasonabic, considering that one containment purge valva remains closed (refer to the SR 3.6.1.3.1), so that a res, '+each of containment does not exist.
In accordance with Requirea Action D.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, through a system walkdown, that those isolation devices outside containment and potentially capable of being mispositioned are in the correct position.
(continued)
ABWR TS B 3.6-8 P&R, 08/04/93 3:27pm i
e o
PCIVs B 3.6.1.3 BASES I
ACTIONS D.I. D.2. and 0.3 (continued)
For the isolation devices inside containment, the time period specified as " prior to entering MODE 4, from MODE 5 4
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.
For the containment purge valve with resilient seal that is isolated in accordance with Required Action D.1, SR 3.6.1.3.7 must be performed at least once every [92]
days. This provides assurance 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 3.6.1.3.7, 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.
E.1 and E.2 If any Required Action and associated Completion Time cannot be met in MODE 1, 2, or 3, the plant must be brought to a MODE in which the LC0 does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and to MODE 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full l
power conditions in an orderly manner and without challenging plant systems.
i F.1. G.I. H.l. and H.2 If any Required Action and associated Completion Time cannot be met, the unit must be placed in a condition in which the LCO does not apply.
If applicable, CCRE ALTERATIONS and i
i movement of irradiated fuel assemblies must be immediately suspended.
Suspension of these activities shall not i
preclude completion of movement of a component to a safe l
(continued) i ABWR TS B 3.6-9 P&R, 08/04/93 3:27pm
9 PCIVs B 3.6.1.3 BASES l
\\
ACTIONS F.1. G.I. H.l. and H.2 (continued) f condition. Also, if applicable, action must be immediately initiated to suspend operations with a potential for i
draining the reactor vessel (0PDRVs) to minimize the probability of a vessel draindown and subsequent potential for fission product release. Actions must continue until OPDRVs are suspended and valve (s) are restored to OPERABLE status.
If suspending an OPDRVs would result in closing the residual heat removal (RHR) shutdown cooling isolation valves, an alternative Required Action is provided to immediately initiate action to restore the valve (s) to OPERABLE status. This allows RHR to remain in service while l
actions are being taken to restore the valve.
SURVEILLANCE SR 3.6.1.3.1 REQUIREMENTS Each 550 mm (22 inch) primary containment purge valve is required to be verified sealed closed at 31 day intervals.
This SR is designed to ensure that a gross breach of primary containment is not caused by an inadvertent or spurious j
opening of a primary containment purge valve.
Detailed analysis of the purge valves failed to conclusively l
demonstrate their ability to close during a LOCA in time to l
limit offsite doses. Primary containment purge valves that are sealed closed must have motive power to the valve operator removed.
This can be accomplished by de-energizing the source of electric power or removing the air supply to the valve operator.
In this application, the term " sealed" has no connotation of leak tightness.
The 31 day Frequency is a result of an NRC initiative, Generic Issue B-24 (Ref. 4), related to primary containment purge valve use during unit operations.
i This SR allows a valve that is open under administrative controls to not meet the SR during the time the valve is open. Opening a purge valve under administrative controls is restricted to one valve in a penetration flow path at a given time (refer to discussion for Note 1 of the ACTIONS) i in order to effect repairs to that valve. This allows one purge valve to be opened without resulting in a failure of the Surveillance and resultant entry into the ACTIONS for this purge valve, provided the stated restrictions are met.
Condition E must be entered during this allowance, and the (continued) s h
ABWR TS B 3.6-10 P&R, 08/04/93 3:27pm
O O
PCIVs B 3.6.1.3 BASES SURVEILLANCE SR 3.6.1.3.1 (continued)
REQUIREMENTS valve opened only as necessary for effecting repairs.
Each purge valve in the penetration flow path may be alternately opened, provided one remains sealed closed, if necessary, to complete repairs on the penetration.
The SR is modified by a Note stating that primary j
containment purge valves are only required to be sealed closed in MODES 1, 2, and 3.
If a LOCA inside primary containment occurs in these MODES, the purge valves may not be capable of closing before the pressure pulse affects systems downstream of the purge valves or the release of radioactive material will exceed limits prior to the closing of the purge valves. At other times when the purge valvns j
are required to be capable of closing (e.g., during handling of irradiated fuel), pressurization concerns are not present and the purge valves are allowed to be open.
SR 3.6.1.3.2 This SR ensures that the primary containment purge valves are closed as required or, if open, open for an allowable j
reason.
j i.
[The SR is also modified by a Note (Note 1), stating that primary containment purge valves are only required to be closed in MODES 1, 2, and 3.
If a LOCA inside primary containment occurs in these MODES, the purge valves may not be capable of closing before the pressure pulse affects systems downstream of the purge valves, or the release of radioactive material will exceed limits prior to the purge valves closing. At other times when the purge valves are required to be capable of closing (e.g., during handling of irradiated fuel), pressurization concerns are not present 4
and the purge valves are allowed to be open.]
The SR is modified by a Note (Note 2) stating that the SR is not required to be met when the purge valves are open for the stated reasons. The Note states that these valves may be opened for inerting, de-inerting, pressure control, ALARA, or air quality considerations for personnel entry, or Surveillances that require the valves to be open. The 550 mm (22 inch) purge valves are capable of closing in the environment following a LOCA. Therefore, these valves are (continued)
ABWR TS B 3.6-11 P&R, 08/04/93 3:42pm j
O PCIVs B 3.6.1.3 BASES SURVEILLANCE SR 3.6.1.3.2 (continued)
REQUIREMENTS allowed to be open for limited periods of time. The 31 day Frequency is consistent with other PCIV requirements discussed in SR 3.6.1.3.3.
SR 3.6.1.3.3 This SR verifies that each primary containment isolation manual valve and blind flange that is located outside primary containment, and is required to be closed during accident conditions, is closed. The SR helps to ensure that post accident leakage of radioactive fluids or gases outside the primary containment boundary is within design limits.
This SR does not require any testing or valve manipulation.
Rather, it involves verification, through a system walkdown, that those valves outside primary containment, and capable of being mispositioned, are in the correct position. Since verification of valve position for valves outside primary containment is relatively easy, the 31 day Frequency was chosen to provide added assurance that the valves are in the correct positions.
Two Notes have been added to this SR. The first Note allows valves and blind flanges located in high radiation areas to be verified by use of administrative controls. Allowing verification by administrative controls is considered acceptable, since the primary containment is inerted and access to these areas is-typically restricted during MODES 1, 2, and 3 for ALARA reasons. Therefore, the probability of misalignment of these valves, once they have been verified to be in the proper position, is low. A second Note has been included to clarify that valves that are open under administrative controls are not required to meet the SR during the time that the valves are open.
SR 3.6.1.3.4 This SR verifies that each primary containment manual isolation valve and blind flange that is located inside primary containment, and is required to be closed during accident conditions, is closed. The SR helps to ensure that-post accident leakage of radioactive fluids or gases outside the primary containment boundary is within design limits.
(continued)
ABWR TS B 3.6-12 P&R, 08/04/93 3:44pm
o PCIVs B 3.6.1.3 BASES SURVEILLANCE SR 3.6.1.3.4 (continued)
REQUIREMENTS For valves inside primary containment, the Frequency defined as " prior to entering MODE 2 or 3 from MODE 4 if primary i
containment was de-inerted while in MODE 4, if not performed within the previous 92 days," is appropriate since these valves and flanges are operated under administrative I
controls and the probability of their misalignment is low.
Two Notes have been added to this SR. The first Note allows l
valves and blind flanges located in high radiation areas to be verified by use of administrative controls. Allowing verification by administrative controls is considered acceptable since the primary containment is inerted and access to these areas is typically restricted during MODES 1, 2, and 3 for ALARA reasons. Therefore, the probability of misalignment of these valves, once they have been verified to be in their proper position, is low. A second Note has been included to clarify _ that valves that are open under administrative controls are not required to meet the SR during the time that the valves are open.
[
The traversing incore probe (TIP) shear isolation valves are actuated by explosive charges. Surveillance of explosive charge continuity provides assurance that TIP valves will actuate when required. Other administrative controls, such t
as those that limit the shelf life of the explosive charges, must be followed. The 32 day Frequency is based on i
operating experience that has demonstrated the reliability of the explosive charge continuity.
SR 3.6.1.3.6 Verifying the isolation time of each power operated and each i
automatic PCIV is within limits is required to demonstrate OPERABILITY. MSIVS may be excluded from this SR since MSIV full closure isolation time is demonstrated by SR 3.6.1.3.8.
The isolation time test ensures that 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 requirements of the Inservice Testing Program or 92 days.
(continued)
ABWR TS B 3.6-13 P&R, 08/04/93 3:47pm
PCIVs B 3.6.1.3 BASES SURVEILLANCE SR 3.6.1.3.7 REQUIREMENTS (continued)
For primary containment purge valves with resilient seals, 4
additional leakage rate testing beyond the test requirements of 10 CFR 50, Appendix J (Ref. 5), 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 primary containment and the environment), a Frequency of 184 days was established as part of the NRC resolution of Generic Issue B-20 (Ref. 3). Additionally, this SR must be performed once 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 which occurs 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.
The SR is modified by a Note stating that the primary containment purge valves are only required to meet leakage rate testing requirements in MODES 1, 2, and 3.
If a LOCA inside primary containment occurs in these MODES, purge valve leakage must be minimized to ensure offsite radiological release is within limits. At other times when the purge valves are required to be capable of closing (e.g., during handling of irradiated fuel), pressurization concerns are not present and the purge valves are allowed to be open.
i 1
- 1 A second Note has been added to this SR requiring that the results be evaluated against the acceptance criteria of SR 3.6.1.1.1.
This ensures that primary containment purge valve leakage is properly accounted for in determining the overall primary containment leakage rate.
)
Verifying the isolation time of each MSIV is within the specified limits is required to demonstrate OPERABILITY.
The isolation time test ensures that the MSIV will isolate in a time period that does not exceed the times assumed in the DBA analyses.
This ensures that the calculated radiological consequences of these events remain within (continued)
[
i ABWR TS B 3.6-14 P&R, 08/04/93 3:49pm
O s
PCIVs B 3.6.1.3 BASES SURVEILLANCE SR 3.6.1.3.8 (continued)
REQUIREMENTS 10 CFR 100 limits. The Frequency of this SR is in accordance with the requirements of the Inservice Testing Program or 18 months.
SR 3.6.1.3.9 Automatic PCIVs close on a primary containment isolation signal to prevent leakage of radioactive material from primary containment following a DBA.
This SR ensures that each automatic PCIV will actuate to its isolation position on a primary containment isolation signal. The LOGIC SYSTEM FUNCTIONAL TEST in SR 3.3.6.3.6 overlaps this SR to provide complete testing of the safety function. The 18 month Frequency was developed considering it is prudent that this Surveillance be performed only during a unit outage since isolation of penetrations would eliminate cooling water flow and disrupt the normal operation of many critical components. 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.
SR 3.6.1.3.10 This SR requires a demonstration that each reactor instrumentation line excess flow check valve (EFCV) is OPERABLE by verifying that the valve reduces flow to 3
s; 3.8 E-3 m /hr (1 gph) on a simulated instrumer.t line break. This SR provides assurance that the instrumentation line EFCVs will perform so that predicted radiological consequences will not be exceeded during the postulated instrument line break event evaluated in Reference 6.
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 l
Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.
(continued)
ABWR TS B 3.6-15 P&R, 08/05/93 11:14am
e PCIVs B 3.6.1.3 BASES SURVEILLANCE SR 3.6.1.3.11 REQUIREMENTS (continued)
The TIP shear isolation valves are actuated by explosive charges. An in place functional test is not possible with this design. The explosive squib is removed and tested to provide assurance that the valves will actuate when required.
The replacement charge for the explosive squib shall be from the same manufactured batch as the one fired or from another batch that has been certified by having one of the batch successfully fired. The Frequency of 18 months on a STAGGERED TEST BASIS is considered adequate given the administrative controls on replacement charges and the frequent checks of circuit continuity (SR 3.6.
3.5).
SR 3.6.1.3.12
- 2 The analyses in References 2 and 6 are based on leakage that is less than the specified leakage rate.
Leakage 3
through each MSIV must be <; I m /hr (35 scfh) when tested at 2
2 Pt of 1.76 Kg/cm g (25 psig). The MSIV leakage rate must be verified to be in accordance with the leakage test requirements of 10 CFR 50, Appendix J (Ref. 5), as modified by approved exemptions. A Note has been added to this SR requiring the results to be evaluated against the acceptance criteria of SR 3.6.1.1.1.
This ensures that MSIV leakage is properly accounted for in determining the overall primary containment leakage rate.
The Frequency is required by 10 CFR 50, Appendix J, as modified by approved exemptions; thus, SR 3.0.2 (which allows Frequency extensions) does not apply.
SR 3.6.1.3.13 Surveillance of hydrostatically tested lines provides assurance that the calculation assumptions of Reference 2 are met. Note also that dual function valves must pass all applicable SRs, including the Type C leakage rate test (SR 3.6.1.1.1), if appropriate.
The combined leakage rates must be demonstrated in accordance with the leakage rate test requirements of 10 CFR 50, Appendix J (Ref. 5), as modified by approved exemptions.
This SR has been modified by two Notes.
Note I states that these valves are only required to meet the combined leakage (continued)
ABWR TS B 3.6-16 P&R, 08/04/93 3:56pm
^
P PCIVs B 3.6.1.3 BASES SURVEILLANCE REQUIREMENTS SR 3.6.1.3.13 (continued)
Coolant System is pressurized and primary c or required.
capable of automatically closing during MOD nment is c
MODES 1, 2, and 3.
However, their leak tightness under l
accident conditions is not required in these other MODES conditions.
Note 2 has been added to this SR requiring the results to be evaluated against the acceptance crit or SR 3.6.1.1.1.
accounted for in determining the overall primary eria of leakage rate.
a nment SR 3.6.1.3.14 Reviewer's Note:
with purge valves with resilient seals allowed to beT during [ MODE 1 are not permane,ntly installed on the valves 2, 3, or 4) an open Verifying each 550 mm valve is blocked to res(trict opening to s; [50]% i22 inch) to ensure that the valves can close under DBA c s required within the times assumed in the analysis of Referenc onditions and 6.
es 3 required to be met in MODES 1, 2, and 3.[The S s only the purge valves must close to maintain c]ontaiIf a LOCA occurs, within the values assumed in the accident analysis nment leakage other times w closing (e.g. hen purge valves are required to be capable of At
, during movement of irradiated fuel assemblies the purge v)a,lves can be fully open.s are not present, thu pressurization concern typically removed only during a refueling outaF The (18 month ces are ge.
REFERENCES 1.
2.
3.
ABWR TS (continued)
B 3.6-17 P&R, 08/04/93 3:27pm
-M
A PCIVs B 3.6.1.3 BASES (continued)
REFERENCES 4.
Generic Issue B-20, " Containment Leakage Due to Seal (continued)
Deterioration."
5.
Generic Issue B-24.
6.
ABWR TS B 3.6-18 P&R, 08/04/93 3:27pm
e Drywell Pressure B 3.6.1.4 8 3.6 CONTAINMENT SYSTEMS B 3.6.1.4 Drywell Pressure BASES BACKGROUND The drywell pressure is limited during normal operations to preserve the initial conditions assumed in the accident analysis for a Design Basis Accident (DBA) or loss of coolant accident (LOCA).
Primary contaiimant performance is evaluated for the entire spectrum of b"eak sizes for postulated LOCAs (Ref.1).
Among the inrats to the DBA is the initial primary containment internal pressure (Ref.1)g.
Analyses assume an initial drywell pressure of, 05 Kg/cm (0.75 psig).
This limitation ensures that the safety analysis remains valid by maintaining the expected initial conditions and ensures drywell interpal pressure does not exceed maximum allowable of 3.16 Kg/cm g (45 psig). The maximum calculated drywell pressure occurs during the rector blowdown phase of the DBA, which assumes a feedwater line break. Thecalculatedpeak drywell pressure for this limiting event is 2.74 Kg/cm g (39 psig) (Ref. 1).
APPLICABLE Drywell pressure satisfies Criterion 2 of the NRC Policy SAFETY ANALYSES Statement.
LC0 In the event of a DBA, with an initial drywell pressure 2
s.05 Kg/cm g (0.75 psig), the resultant peak drywell accident pressure will be maintained below the drywell design pressure.
APPLICABILITY In MODES 1, 2, and 3, a DBA could cause a release of radioactive material to primary containment.
In MODES 4 and 5, the probability and consequences of these events are reduced due to the pressure and temperature limitations of these MODES. Therefore, maint-ining drywell pressure within limits is not required in MODE or 5.
(continued)
ABWR TS B 3.6-1 P&R, 08/05/93 11:19am
s>
Drywell Pressure B 3.6.1.4 l
BASES ACTIONS A1 With drywell pressure not within the limit of the LCO, drywell pressure must be restored 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 primary 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 j
LC0 3.6.1.1, " Primary Containment," which requires that primary containment be restored to OPERABLE status within i
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 drywell pressure cannot be restored to within limit within the required Completion Time, the plant must be brought to a MODE in which the LCO does not apply.
To achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and to MODE 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
SVRVEILLANCE SR 3.6.1.4.1 REQUIREMENTS Verifying that drywell pressure is within limit ensures that unit operation remains within the limit assumed in the primary 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 drywell pressure variations and pressure instrument drift during the applicable MODES.
Furthermore, the 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-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 drywell i
pressure condition.
REFERENCES 1.
i ABWR TS B 3.6-2 P&R, 08/05/93 7:45am i
i i
l i
Containment Systems B 3.6.1.5 l
B 3.6 CONTAINMENT SYSTEMS B 3.6.1.5 Drywell Air Temperature BASES BACKGROUND The drywell contains the reactor vessel and piping, which add heat to the airspace.
Drywell coolers remove heat and i
maintain a suitable environment. The average airspace temperature affects the calculated response to postulated Design Basis Accidents (DBAs). The limitation on the drywell average air temperature was developei as reasonable, based on operating experience. The limitation on drywell air temperature is used in the Reference 1 safety analyses.
I APPLICABLE Primary containment performance is evaluated for a spectrum SAFETY ANALYSES of break sizes for postulated loss of coolant accidents i
(LOCAs) (Ref. 1). Among the inputs to the design basis analysis is the initial drywell average air temperature (Ref. 1). Analyses assume an initial average drywell air temperature of 57*C (135'F). This limitation ensures that the safety analysis remains valid be maintaining the expected initial conditions and ensures that the peak LOCA drywell temperature does not exceed the maximum allowable i
temperature of 171*C (340*F) (Ref. 2).
Exceeding this design temperature may result in the degradation of the j
primary containment structure under accident loads.
i Equipment inside primary containment, required to mitigate the effects of a DBA, it designed to operate and be capable of operating under environmental conditions expected for the accident.
The most severe drywell temperature condition occurs as a l
result of a small Reactor Coolant System rupture above the reactor water level, which results in the blowdown of reactor steam to the drywell. The drywell temperature analysis considers main steam line breaks occurring inside the drywell and having various break areas. The maximum calculated drywell average temperature for the worst case i
break area is provided in Reference 2.
i Drywell air temperature satisfies Criterion 2 of the NRC Policy Statement.
l (continued)
ABWR TS B 3.6-1 P&R, 08/05/93 8:17am l
s Containment Systems r
B 3.6.1.5 BASES LC0 In the event of a DBA, with an initial drywell average air temperature less than or equal to the LC0 temperature limit, the resultant peak accident temperature is maintained below the drywell design temperature. As a result, the ability of primary containment to perform its design function is ensured.
APPLICABILITY In MODES 1, 2, and 3, a DBA could cause a release of radioactive material to primary containment.
In MODES 4 and 5, the probability and consequences of these events are reduced due to the pressure and temperature limitations of these MODES. Therefore, maintaining drywell average air temperature within the limit is not required in MODE 4 or 5.
ACTIONS A.1 With drywell average air temperature not within the limit of the LCO, drywell average temperature must be restored within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. The Required Action is necessary to return operation to within the bounds of the primary containment analysis. The 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> Completion Time is acceptable, considering the sensitivity of the analysis to variations in this parameter, and provides sufficient time to correct minor problems.
i B.I and B.2 If the drywell average air temperature cannot be restored to within limit within the required Completion Time, the plant must be brought to a MODE in which the LC0 does not apply.
To achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and to MODE 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
(continued)
ABWR TS B 3.6-2 P&R, 08/05/93 8:17am f
._~
f
\\
Containment Systems B 3.6.1.5 BASES SURVEILLANCE SR 3.6.1.5.1 REQUIREMENTS Verifying that the drywell average air temperature is within the LCO limit ensures that operation remains within the limits assumed for the primary containment analyses.
Drywell air temperature is monitored in all quadrants and at various elevations (referenced to mean sea level).
Due to the shape of the drywell, a volumetric average is used to determine an accurate representation of the actual average temperature.
I The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Frequency of the SR was developed based on operating experience related to drywell average air temperature variations and temperature instrument drift i
during the applicable MODES and the low probability of a DBA occurring between surveillances.
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 i
l indications available in the control room, including alarms, to alert the operator to an abnormal drywell air temperature condition.
l l
REFERENCES 1.
2.
4 r
I t
i i
i i
I f
ABWR TS B 3.6-3 P&R, 08/05/93 8:17am
1 Containment Systems B 3.6.1.6 B 3.6 CONTAINMENT SYSTEMS B 3.6.1.6 Suppressien-Chamber-to-Drywell Vacuum Breakers 1
i BASES l
BACKGROUND The function of the suppression-chamber-to-drywell vacuum breakers is to relieve vacuum in the drywell. There are [8]
internal vacuum breakers between the drywell and the suppression chamber, which allow air and steam flow from the suppression chamber to the drywell when the drywell is at a negative pressure with respect to the suppression chamber.
Therefore, suppression-chamber-to-drywell vacuum breakers prevent an excessive negative differential pressure across the wetwell drywell boundary.
Each vacuum breaker is a self actuating valve, similar to a check valve, which can be manually operated for testing purposes.
A negative differential pressure across the drywell wall is caused by rapid depressurization of the drywell.
Events that cause this rapid depressurization are cooling cycles, inadvertent drywell spray actuation and steam condensation from sprays or subcooled water reflood of a break in the event of a primary system rupture.
Cooling cycles result in minor pressure transients in the drywell that occur slowly and are normally controlled by heating and ventilation equipment. Spray actuation or spill of subcooled water out of a break results in more significant pressure transients and becomes important in sizing the internal vacuum breakers.
In the event of a primary system rupture, steam condensation
[
within the drywell results in the most severe pressure transient.
Following a primary system rupture, air in the drywell is purged into the suppression chamber free airspace, leaving the drywell full of steam.
Subsequent condensation of the steam can be caused in two possible l
ways, namely, Emergency Core Cooling System flow from a ruptured pipe, or containment spray actuation following a loss of coolant accident (LOCA).
These two cases determine t
the maximum depressurization rate of the drywell.
In addition, the waterleg in the vertical vents of the vent system is controlled by the drywell-to-suppression chamber
^
differential pressure.
If the drywell pressure is less than 4
the suppression chamber pressure, there will be an increase l
in the vent waterleg.
This will result in an increase in (continued) i ABWR TS B 3.6-1 P&R, 08/05/93 1:06pm
)
Containment Systems B 3.6.1.6 BASES j
BACKGROUND the water clearing inertia in the event of a postulated (continued)
LOCA, resulting in an increase in the peak drywell pressure.
This in turn will result in an increase in the pool swell dynamic loads. The internal vacuum breakers limit the height of the waterleg in the vent system during normal operation.
Analytical methods and assumptions involving the suppression-chamber-to-drywell vacuum breakers are presented in Reference 1 as part of the accident response of the primary containment systems. The vacuum breakers are provided as part of the primary containment to limit the negative differential pressure across the drywell and suppression chamber walls that form part of the primary containment boundary.
APPLICABLE The safety analyses assume that the internal vacuum breakers SAFETY ANALYSES are closed initially anp are fully open at a differential pressure of.0352 Kg/cm d (0.5 psid) (Ref.1).
Additionally,1 of the 8 internal vacuum breakers are assumed to fail in a closed position (Ref. 1). The results of the analyses show that the design pressure is not exceeded even under the worst case accident scenario. The vacuum breaker opening differential pressure setpoint and the requirement that all 8 vacuum breakers be OPERABLE are a result of the requirement placed on the vacuum breakers to limit the vent system waterleg height.
Design Basis Accident (DBA) analyses require the vacuum breakers to be closed initially and to remain closed and leak tight, with the suppression pool at a positive pressure relative to the drywell.
The suppression-chamber-to-drywell vacuum breakers satisfy Criterion 3 of the NRC Policy Statement.
LCO All 8 of the vacuum breakers must be OPERABLE for opening.
All suppression-chamber-to-drywell vacuum breakers, however, are required to be closed (except during testing or when the vacuum breakers are performing the intended design function).
The vacuum breaker OPERABILITY requirement provides assurance that the drywell-to-suppression chamber negative differential pressure remains below the design (continued)
ABWR TS B 3.6-2 P&R, 08/05/93 11:41am
j i
s Containment Systems B 3.6.1.6 BASES LCO value. The requirement that the vacuum breakers be closed (continued) ensures that there is no excessive bypass leakage should a LOCA occur.
APPLICABILITY In MODES 1, 2, and 3, the containment sprays of the residual heat removal system are required to be OPERABLE to mitigate the effects of a DBA.
Excessive negative pressure inside the drywell could occur due to inadvertent actuation of the drywell spray. The vacuum breakers, therefore, are required to be OPERABLE in MODES 1, 2, and 3, when the containment sprays are required to be OPERABLE, to mitigate the effects of inadvertent actuation of the drywell spray. Also, in MODES 1, 2, and 3, a DBA could result in excessive negative differential pressure across the drywell wall, caused by the rapid depressurization of the drywell.
i The event that results in the limiting rapid depressurization of the drywell is the primary system rupture that purges the drywell of air and fills the drywell free airspace with steam.
Subsequent condensation of the steam would result in depressurization of the drywell.
The limiting pressure and temperature of the primary system prior to a DBA occur in MODES 1, 2, and 3.
In MODES 4 and 5, the probability and consequences of these events are reduced by the pressure and temperature limitations in these MODES; therefore, maintaining suppression-chamber-to-drywell vacuum breakers OPERABLE is not required in MODE 4 or 5.
ACTIONS A.1 With one of the required vacuum breakers inoperable for opening (e.g., the vacuum breaker is not open and may be l
stuck closed or not within its opening setpoint limit, so that it would not function as designed during an event that depressurized the drywell), the remaining [seven] OPERABLE vacuum breakers are capable of providing the vacuum relief function. However, overall system reliability is reduced because a single failure in one of the remaining vacuum breakers could result in an excessive suppression-chamber-to-drywell differential pressure during a DBA.
(continued)
ABWR TS B 3.6-3 P&R, 08/05/93 11:41am
s L
Containment Systems B 3.6.1.6 r
BASES ACTIONS A.1 (continued)
Therefore, with one of the [eight] required vacuum breakers inoperable, 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is allowed to restore at least one of the inoperable vacuum breakers to OPERABLE status so that 4
plant conditions are consistent with those assumed for the design basis analysis. The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Completion Time is considered acceptable due to the low probability of an event in which the remaining vacuum breaker capability would not be adequate.
B.1 and B.2 An open vacuum breaker allows communication between the drywell and suppression chamber airspace, and, as a result, there is the potential for suppression chamber overpressurization due to this bypass leakage if a LOCA were to occur. Therefore, the open vacuum breaker must be closed. A short time is allowed to close the vacuum breaker due to the low probability of an event that would pressurize primary containment.
If vacuum breaker position indication is not reliable, an alternate method of verifying that the vacuum breakers are clpsed is to verify that a differential pressure of.035 Kg/cm d (0.5 psid) between the suppression chamber and drywell is maintained for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> without makeup.
The required 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Completion Time is considered adequate to perform this test.
C1 and C.2 If the inoperable suppression-chamber-to-drywell vacuum breaker cannot be closed or restored to OPERABLE status within the required Completion Time, the plant must be brought to a MODE in which the LC0 does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and to MODE 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant cor.ditions from full power conditions in an orderly manner and without challenging plant systems.
(continued)
ABWR TS B 3.6-4 P&R, 08/05/93 11:41am
t 1
Containment Systems B 3.6.1.6 BASES SURVEILLANCE SR 3.6.1.6.1 REQUIREMENTS Each vacuum breaker is verified closed (except when being tested in accordance with SR 3.6.1.6.2 or when performing its intended function) to ensure that this potential large bypass leakage path is not present. This Surveillance is performed by observing the vacuum breaker position indication or by verifying that a differential pressure of 2
.035 Kg/cm d (0.5 psid) between the suppression chamber and drywell is maintained for I hour without makeup. The 14 day Frequency is based on engineering judgment, is considered adequate in view of other indications of vacuum breaker status available to operations personnel, and has been shown 1
to be acceptable through operating experience.
This i
verification is also required within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after any discharge of steam to the suppression chamber from the safety / relief valves or any operation that causes the drywell-to-suppression phamber differential pressure to be reduced by 2.035 Kg/cm d (0.5 psid). A footnote is added to provide additional assurance of closure if position indication instruments indicate one or more vacuum breakers are not closed.
SR 3.6.1.6.2 Each required vacuum breaker must be cycled to ensure that it opens adequately to perform its design function and returns to the fully closed position.
This ensures that the safety analysis assumptions are valid. The 18 month Frequency of this SR is based on the need to perform the surveillance during an outage.
The vacuum breakers can only be manually actuated and are only accessible during an outage.
SR 3.6.1.6.3 Verification of the vacuum breaker opening setpoint is necessary to ensure that the safety analysis assumption regarding vacuum breaker full open differential pressure of 2
.035 Kg/cm g (0.5 psid) is valid.
The 18 month Frequency is i
based on the need to perform this Surveillance under the l
conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance (continued)
ABWR TS B 3.6-5 P&R, 08/05/93 11:41am
5
+
Containment Systems B 3.6.1.6 i,
BASES l
SURVEILLANCE SR 3.6.1.6.3 (continued)
REQUIREMENTS were performed with the reactor at power.
For this facility, the 18 month Frequency has been shown to be acceptable, based on operating experience, and is further justified because of other surveillances performed at shorter Frequencies that convey the proper functioning status for each vacuum breaker.
REFERENCE 1.
l i
1 l
1 l
ABWR TS B 3.6-6 P&R, 08/05/93 II:41am l
A Suppression Pool Average Temperature B 3.6.2.1 B 3.6 CONTAINMENT SYSTEMS B 3.6.2.1 Suppression Pool Average Temperature BASES BACKGROUND The suppression chamber is a steel lined reinforced concrete pressure vessel containing a volume of water called the suppression pool. The suppression pool is designed to absorb the decay heat and sensible energy released during a reactor blowdown from safety / relief valve discharges or from Design Basis Accidents (DBAs). The suppression pool must quench all the steam released through the vent lines during a loss of coolant accident (LOCA).
This is the essential mitigative feature of a pressure suppression containment that ensures that the peak containment pressure is maintained below the maximum allowable pressure for DBAs of 2
3.16 Kg/cm g (45 psig).
The suppression pool must also condense steam from steam exhaust lines in the turbine driven systems (i.e., the Reactor Core Isolation Cooling System.
Suppression pool average temperature (along with LCO 3.6.2.2, " Suppression Pool Water Level") is a key indication of the capacity of the suppression pool to fulfill these requirements.
The technical concerns that lead to the development of suppression pool average temperature limits are as follows:
a.
Complete steam condensation-the original limit for the end of a LOCA blowdown was 76.67'C (170*F), based on the Bodega Bay and Humboldt Bay Tests; b.
Primary containment peak prespure and temperature-the design pressure is 3.16 Kg/cm g (45 psig) and design temperature is 171*C (340*F) (Ref. 1).
c.
Condensation oscillation loads maximum allowable initial temperature is 49'C (120*F).
(continued)
ABWR TS B 3.6-1 P&R, 08/05/93 1:38pm
o Suppression Pool Average Temperature B 3.6.2.1 BASES (continued)
APPLICABLE The postulated DBA against which the primary containment SAFETY ANALYSES performance is evaluated is the entire spectrum of postulated pipe breaks within the primary containment.
Inputs to the safety analyses include initial suppression pool water volume and suppression pool temperature (Reference 1 for LOCAs and Reference 2 for the pool temperature analyses required by Reference.3). An initial pool temperature of 43*C (110*F) is assumed for the Reference I and 2 analyses.
Reactor shutdown at a pool temperature of 49'C (120*F) and vessel depressurization at a pool temperature of 54*C (130*F) are assumed for the Reference 2 analyses.
The limit of 46*C (ll5'F), at which testing is terminated, is not used in the safety analyses because DBAs are assumed to not initiate during unit testing.
Suppression pool average temperature satisfies Criteria 2 and 3 of the NRC Policy Statement.
LCO A limitation on the suppression pool average temperature is required to provide assurance that the containment conditions assumed for the safety analyses are met. This limitation subsequently ensures that peak primary containment pressures and temperatures do not exceed maximum allowable values during a postulated DBA or any transient resulting in heatup of the suppression pool. The LCO requirements are:
a.
Average temperature s 43*C (110*F) when THERMAL POWER is 1% RTP and no testing that adds heat to the suppression pool is being performed. This requirement ensures that licensing bases initial conditions are met.
b.
Average temperature 5 46*C (ll5'F) when THERMAL POWER is 1% RTP and testing that adds heat to the suppression pool is being performed. This required value ensures that the unit has testing flexibility, and was selected to provide margin below the 49'C (120*F) limit at which reactor shutdown is required.
When testing ends, temperature must be restored to s 43*C (110*F) within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> according to Required (continued)
ABWR TS B 3.6-2 P&R, 08/05/93 1:04pm
l N
t Suppression Pool Average Temperature B 3.6.2.1 BASES (continued)
APPLICABLE The postulated DBA against which the primary containment SAFETY ANALYSES performance is evaluated is the entire spectrum of postulated pipe breaks within the primary containment.
Inputs to the safety analyses include initial suppression pool water volume and suppression pool temperature (Reference 1 for LOCAs and Reference 2 for the pool temperature analyses required by Reference 3).
An initial pool temperature of 43*C (110*F) is assumed for the Reference 1 and 2 analyses.
Reactor shutdown at a pool temperature of 49'C (120*F) and vessel depressurization at a pool temperature of 54*C (130*F) are assumed for the Reference 2 analyses. The limit of 46*C (115'F), at which testing is terminated, is not used in the safety analyses because DBAs are assumed to not initiate during unit testing.
Suppression pool average temperature satisfies Criteria 2 and 3 of the NRC Policy Statement.
LC0 A limitation on the suppression pool average temperature is required to provide assurance that the containment conditions assumed for the safety analyses are met.
This limitation subsequently ensures that peak primary containment pressures and temperatures de not exceed maximum allowable values during a postulated DBA or any transient resulting in heatup of the suppression pool. The LCO requirements are:
a.
Average temperature s 43*C (110*F) when THERMAL POWER is 1% RTP and no testing that adds heat to the suppression pool is being performed. This requirement ensures that licensing bases initial conditions are met.
b.
Average temperature s 46*C (ll5'F) when THERMAL POWER is 1% RTP and testing that adds heat to the suppression pool is being performed.
This required value ensures that the unit has testing flexibility, and was selected to provide margin below the 49'C (120*F) limit at which reactor shutdown is required.
When testing ends, temperature must be restored to s 43*C (110*F) within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> according to Required (continued)
ABWR TS B 3.6-2 P&R, 08/05/93 1:04pm r: z ~ ~
~'
i s
Suppression Pool Average Temperature B 3.6.2.1 BASES (continued)
LC0 b.
(continued)
Action A.2.
Therefore, the time period that the temperature is > 43*C (110*F) is short enough not to cause a significant increase in unit risk.
1 c.
Average temperature s 49'c (120*F) when THERMAL POWER is > 1% RTP. This requirement ensures that the unit will be shut down at > 49'C (120*F).
The pool is designed to absorb decay heat and sensible heat but could be heated beyond design limits by the steam generated if the reactor is not shut down.
At the 1% RTP power level, heat input is approximately equal to normal system heat losses.
P APPLICABILITY In MODES 1, 2, and 3, a DBA could cause significant heatup of the suppression pool.
In MODES 4 and 5, the probability and consequences of these events are reduced due to the pressure and temperature limitations in these MODES.
Therefore, maintaining suppression pool average temperature within limits is not required in MODE 4 or 5.
ACTIONS A.1 and A.2 With the suppression pool average temperature above the specified limit when not performing testing that adds heat to the suppression pool and when above the specified power indication, the initial conditions exceed the conditions assumed for the Reference 1, 3, and 4 analyses.
- However, primary containment cooling capability still exists, and the primary containment pressure suppression function will occur at temperatures well above those assumed for safety analyses. Therefore, continued operation is allowed for a limited 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 adequate to allow the suppression pool average temperature to be restored below the limit. Additionally, when suppression pool temperature is > 43*C (110*F), increased nnitoring of-the suppression pool temperature is required tv ensure that it remains s 49'C (120*F).
The once per hour Completion Time is adequate based on past experience, which has shown (continued) i ABWR TS B 3.6-3 P&R, 08/05/93 1:04pm i
l
i 3
Suppression Pool Average Temperature B 3.6.2.1 BASES ACTIONS A.1 and A.2 (continued) that pool temperature increases relatively slowly except when testing that adds heat to the suppression pool is being performed.
Furthermore, the once per hour Completion Time is considered adequate in view of other indications in the control room, including alarms, to alert the operator to 6
an abnormal suppression pool average temperature condition.
B.1 If the suppression pool average temperature 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 power must be reduced to s 1% RTP within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Completion Time is reasonable, based on operating experience, to reduce power from full power conditions in an orderly manner and without challenging plant systems.
C.1 Suppression pool average temperature is allowed to be > 43*C (110*F) when THERHAL POWER is >1% RTP, and when testing that adds heat to the suppression pool is being performed.
However, if temperature is > 46'C (115'F) all testing must be immediately suspended to preserve the heat absorption capability of the suppression pool. With the testing suspended, Condition A is entered and the Required Actions and associated Completion Times are applicable.
D.1 and D.2 When suppression pool average temperature is > 49'C (120*F),
the suppression pool cooling function of the RHR system is automatically initiated. The pool temperature continues to increase due to the mismatch of cooling capacity and steam discharged into the pool.
When the pool temperature reaches 54*C (130*F) a reactor scram is automatically initiated.
Additionally, when suppression pool temperature is > 49'C (120*F), increased monitoring of pool temperature is required to ensu'.e that it remains s 54*C (130*F).
The once per 30 minute Completion Time is adequate, based on (continued)
ABWR TS B 3.6-4 P&R, 08/05/93 1:04pm
i s
l Suppression Pool Average Temperature B 3.6.2.1 BASES ACTIONS D.] and D.2 (continued)
I operating experience. Given the high suppression pool average temperature in this Condition, the monitoring Frequency is increased to twice that of Condition A.
Furthermore, the 30 minute Completion Time is considered adequate in view of other indications available in the control room, including alarms, to alert the operator to an abnormal suppression pool average temperature condition.
I a
E.1 and E.2 If suppression pool average temperature cannot be maintained I
at $; 54*C (130*F), the plant must be brought to a MODE in which the LC0 does not apply. To achieve this status, the reactor pressure must be reduced to < 14.1 Kg/cm*g (200 psig) within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, and the plant must be brought to at least MODE 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion i
Times are reasonable, based on operating experience, to reach the required plant conditions from full power i
conditions in an orderly manner and without challenging plant systems.
Continued addition of heat to the suppression pool with suppression pool temperature > 54*C (130*F) could result in j
exceeding the design basis maximum allowable values for primary containment temperature or pressure.
Furthermore, i
if a blowdown were to occur when the temperature was > 54*C (130*F), the maximum allowable bulk and local temperatures could be exceeded very quickly.
l
]
SURVEILLANCE SR
- 3. 6. 2M REQUIREMENTS The suppression pool average temperature is regularly i
monitored to ensure that the required limits are satisfied.
i The average temperature is determined by taking an arithmetic average of the OPERABLE suppression pool water temperature channels. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Frequency has been shown, based en operating experience, to be acceptable. When heat is being added to the suppression pool by testing, however, i
it is necessary to monitor suppression pool temperature more frequently. The 5 minute Frequency during testing is justified by the rates at which tests will heat up the i
(continued)
ABWR TS B 3.6-5 P&R, 08/05/93 1:04pm
Suppression Pool Average Temperature B 3.6.2.1 BASES SURVEILLANCE SR 3.6.2.1.1 (continued)
REQUIREMENTS suppression pool, has been shown to be acceptable based on operating experience, and provides assurance that allowable pool temperatures are not exceeded. The Frequencies are further justified in view of other indications available in the control room, including alarms, to alert the operator to an abnormal suppression pool average temperature condition.
REFERENCES 1.
2.
3.
" Plant Unique Load Definition Edwin I.
Hatch Nuclear Power Plant Unit 1."
i i
l I
i ABWR TS B 3.6-6 P&R, 08/05/93 1:04pm
e Suppression Pool Water Level B 3.6.2.2 B 3.6 CONTAINMENT SYSTEMS B 3.6.2.2 Suppression Pool Water Level BASES i
BACKGROUND The suppression pool is a steel lined reinforced concrete pressure vessel containing a volume of water called the suppression pool. The suppression pool is designed to absorb the energy associated with decay heat and sensible heat released during a reactor blowdown from safety / relief valve (S/RV) discharges or from a Design Basis Accident (DBA). The suppression pool must quench all the steam released through the vent lines during a loss of coolant accident (LOCA). This is the essential mitigative feature of a pressure suppression containment, which ensures that the peak containment pressure is maintained below he maximum allowable pressure for DBAs of 3.164 Kg/cm}g (45 psig).
The suppression pool must also condense steam from the steam exhaust lines in the turbine driven system (i.e., Reactor Core Isolation Cooling (RCIC) System and provides the main emergency water supply source for the reactor vessel. The suppression pool level ranges between the low water level 3
limit of 7 p))(18 ft 4.5 inches) (at a volume of 3625 m (135,291 ft and the high water level limit of 7.1 m (18 ft 9.75 inches).
If the suppression pool water level is too low, an insufficient amount of water would be available to adequately condense the steam from the S/RV quenchers, main vents, or RCIC turbine exhaust lines.
Low suppression pool water level could also result in an inadequate emergency j
makeup water source to the Emergency Core Cooling System.
The lower volume would also absorb less steam energy before heating up excessively. Therefore, a minimum suppression pool water level is specified.
If the suppression pool water level is too high, it could result in excessive clearing loads from S/RV discharges and excessive pool swell loads during a DBA LOCA. Therefore, a i
maximum pool water level is specified.
This LC0 specifies an acceptable range to prevent the suppression pool water level from being either too high or too low.
(continued)
ABWR TS B 3.6-1 P&R, 08/06/93 10:39am
Suppression Pool Water Level B 3.6.2.2 BASES APPLICABLE Initial suppression pool water level affects suppression SAFETY ANALYSES pool temperature response calculations, calculated drywell pressure during vent clearing for a DBA, calculated pool swell loads for a DBA LOCA, and calculated loads due to S/RV discharges.
Suppression pool water level must be maintained within the limits specified so that the safety analysis of Reference I remains valid.
Suppression pool water level satisfies Criteria 2 and 3 of the NRC Policy Statement.
LC0 A limit that suppression pool water level be 2 7 m 1
j (18 ft 4.5 inches) and s 7.1 m (18 ft 9.75 inches) is required to ensure that the primary containment conditions assumed for the safety analyses are met.
Either the high or low water level limits were used in the safety analyses, depending upon which is more conservative for a particular calculation.
APPLICABILITY In MODES 1, 2, and 3, a DBA could cause significant loads on the primary containment.
In MODES 4 and 5, the probability and consequences of these events are reduced due to the pressure and temperature limitations in these MODES.
Therefore, maintaining suppression pool water level within i
limits is not required in MODE 4 or 5.
ACTIONS A.1 With suppression pool water level outside the limits, the conditions assumed for the safety analyses are not met.
If water level is below the minimum level, the pressure suppression function still exists as long as main vents are covered, RCIC turbine exhausts are covered, and S/RV quenchers are covered.
If suppression pool water level is l
above the maximum level, protection against overpressurization still exists due to the margin in the peak containment pressure analysis or as long as the drywell and containment sprays are OPERABLE. Therefore, continued operation for a limited time is allowed.
The 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> (continued)
ABWR TS B 3.6-2 P&R, 08/06/93 10:38am
a e
Suppression Pool Water Level B 3.6.2.2 BASES ACTIONS A.1 (continued)
Completion Time is sufficient to restore suppression pool water level to within limits. Also, it takes into account the low probability of an event impacting the suppression pool water level occurring during this interval.
B.1 and B.2 If suppression pool water level 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 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and to MODE 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
SURVEILLANCE SR 3.6.2.2.1 REQUIREMENTS Verification of the suppression pool water level is to ensure that the required limits are satisfied. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Frequency of this SR was developed considering operating experience related to trending variations in suppression pool water level and water level instrument drift during the applicable MODES and to assessing the proximity to the specified LC0 level limits.
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 suppression pool water level condition.
REFERENCE 1.
i ABWR TS B 3.6-3 P&R, 08/06/93 10:37am l
l
=
RHR Suppression Pool Cooling B 3.6.2.3 B 3.6 CONTAINMENT SYSTEMS B 3.6.2.3 Residual Heat Removal (RHR) Suppression Pool Cooling l
BASES BACKGROUND Following a Design Basis Accident (DBA), the RHR Suppression 1
Pool Cooling System removes heat from the suppression pool.
The suppression pool is designed to absorb the sudden input of heat from the primary system.
In the long term, the pool continues to absorb residual heat generated by fuel in the reactor core. Some means must be provided to remove heat i
from the suppression pool so that the temperature inside the primary containment remains within design limits.
This function is provided by three redundant RHR suppression pool cooling subsystems. The purpose of this LC0 is to ensure that the three subsystems are OPERABLE in applicable MODES.
Each RHR subsystem contains one pump and one heat exchanger and is manually initiated and independently controlled. The three RHR subsystems perform the suppression pool cooling function by circulating water from the suppression pool through the RHR heat exchangers and returning it to the suppression pool.
Reactor Building Cooling Water (RCW),
S/RV leakage, and high pressure core injection Reactor Core Isolation Cooling System testing increase suppression pool temperature more slowly. The RHR Suppression Pool Cooling System is
'so used to lower the suppression pool water bulk temperatu t following such events.
t APPLICABLE Reference 1 contains the results of analyses used to predict SAFETY ANALYSES primary containment pressure and temperature following large and small break LOCAs. The intent of the analyses is to l
demonstrate that the heat removal capacity of the RHR Suppression Pool Cooling System is adequate to maintain the i
primary containment conditions within design limits. The suppression pool temperature is calculated to remain below the design limit.
Reference 3 contains discussion of additional analyses that was performed to support PRA success criteria for the long term heat removal function. The intent of these analyses was to predict primary containment pressure and temperature following low probability events beyond the DBA and to determine the minimum heat-removal capacity required to (continued)
ABWR TS B 3.6-1 P&R, 08/05/93 3:22pm
RHR Suppression Pool Cooling B 3.6.2.3 U
BASES maintain the primary containment conditions within its ultimate capacity. The results are used to establish the L
minimum amount of RHR (Suppression Pool Cooling) system i
equipment required to prevent ultimate containment failure beyond DBA events.
The RHR Suppression Pool Cooling System satisfies Criterion 3 of the NRC Policy Statement.
i LC0 During a DBA, a minimum of two RHR suppression pool cooling subsystems are required to maintain the primary containment peak pressure and temperature below the design limits (Ref. 1). To ensure that these requirements are met, three RHR suppression pool cooling subsystems must be OPERABLE with power from three safety related independent power supplies. Therefore, in the event of an accident, at least two subsystem are OPERABLE, assuming the worst case single active failure. An RHR suppression pool cooling subsystem is OPERABLE when the pump, the heat exchanger, and associated piping, valves, instrumentation, and controls are OPERABLE.
1 APPLICABILITY In MODES 1, 2, and 3, a DBA could cause a release of radioactive material to primary containment and cause a heatop and pressurization of primary containmer.t.
In MODES 4 and 5, the probability and consequences of these events are reduced due to the pressure and temperature limitations in these MODES. Therefore, the RHR Suppression Pool Cooling System is not required to be OPERABLE in MODE 4 or 5.
j i
ACTIONS AJ t
With one RHR suppression pool cooling subsystem inoperable, the inoperable subsystem must be restored within 7 days.
In this Condition, the remaining RHR suppression pool cooling subsystems are adequate to perform the primary containment cooling function. However, the overall reliability is reduced because a single failure in one of the OPERABLE (continued)
ABWR TS B 3.6-2 P&R, 08/05/93 3:22pm
RHR Suppression Pool Cooling B 3.6.2.3 BASES ACTIONS A.I (continued) subsystem could result in reduced primary containment cooling capability.
The 7 day Completion Time is acceptable in light of the redundant RHR suppression pool cooling capabilities afforded by the OPERABLE subsystems and the low probability of a DBA occurring during this period.
I Additionally, analyses of beyond design basis events demonstrates that one RHR suppression pool cooling subsystem is adequate in maintain containment conditions below the ultimate capacity.
1 fLd With two RHR suppression pool cooling subsystems inoperable, at least one inoperable subsystem must be restored to l
OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
In this Condition, the remaining OPERABLE RHR suppression pool cooling subsystem affords significant primary containment cooling capability and would be sufficient to maintain containment conditions well below its ultimate capacity.
However, the overall reliability is reduced because a single failure in the one OPERABLE subsystem could result in a substantial loss af primary containment cooling capability. The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Completion Time was chosen in light of the redundant RHR suppression pool cooling capability afforded by the OPERABLE train and the low probability of a DBA occurring during this pe: lod.
C.1 and C.2 6
If the Required Action and associated Completion Time of Conditions A and B cannot be met within the required Completion Time or if three RHR suppression pool cooling subsystems are inoperable, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must b brought to at least MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and to MODE 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed t
Completion Times are reasonable, based on operating I
experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
i (continued) j ABWR TS B 3.6-3 P&R, 08/06/93 10:42am i
RHR Suppression Pool Cooling B 3.6.2.3 t
BASES SURVEILLANCE SR 3.6.2.3.1 REQUIREMENTS Verifying the correct alignment for manual, power operated, i
and automatic valves, in the RHR suppression pool cooling mode flow path provides assurance that the proper flow path exists for system operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position since these valves were verified to be in the correct position prior to locking, sealing, or securing. A valve is also allowed to be in the nonaccident position, provided it can be aligned to the accident position within the time assumed in the accident analysis. This is acceptable, since the RHR suppression pool cooling mode is manually initiated. This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of being mispositioned are in the i
correct position. This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves.
i The Frequency of 31 days is justified because the valves are operated under procedural control, improper valve position would affect only a single subsystem, the probability of an event requiring initiation of the system is low, and the subsystem is a manually initiated system. This Frequency has been shown to be acceptable, based on operating experience.
l i
Verifying that each RHR pump develops a flow rate 2 954 m /h (4200 gpm), while operating in the suppression pool cooling mode with fl,e through the associated heat exchanger ensures that pump performance has not degraded during the cycle.
Flow is a normal test of centrifugal pump performance i
required by ASME Code Section XI (Ref. 3). This test i
confirms one point on the pump design curve, and the results are indicative of overall performance. Such inservice inspections confirm component OPERABILITY, trend performance, and detect incipient failures by indicating i
abnormal performance. The Frequency of this SR is [in j
accordance with the Inservice Testing Program or 92 days].
(c.ontinued)
ABWR TS B 3.6-4 P&R,08/05/93 3:22pm
RHR Suppression Pool Cooling B 3.6.2.3 BASES REFERENCES 1.
2.
ASME, Boiler and Pressure Vessel Code,Section XI.
3 ABWR TS B 3.6-5 P&R, 08/05/93 3:24pm i
t
Containment System B 3.6.2.4 i
B 3.6 CONTAINMENT SYSTEMS B 3.6.2.4 Residual Heat Removal (RHR) Wetwell Spray BASES BACKGROUND Following a Design Basis Accident (DBA), the RHR Wetwell Spray System removes heat from the wetwell airspace. The suppression pool is designed to absorb the sudden input of heat from the primary system from a DBA or a rapid depressurization of the reactor pressure vessel (RPV) through safety / relief valves. The heat addition to the suppression pool results in increased steam in the wetwell, which increases primary containment pressure. Steam blowdown from a DBA can also bypass the suppression pool and end up in the wetwell airspace.
Some means must be provided to remove heat from the wetwell so that the pressure and temperature inside primary containment remain within analyzed design limits. This function is provided by two redundant RHR wetwell spray subsystems.
(Only RHR subsystems B and C operate in this mode.) The purpose of this LC0 is to ensure that both subsystems are OPERABLE in 1
applicable MODES.
Each of the two RHR wetwell spray subsystems contains a pump and a heat exchanger, which are manually initiated and independently controlled. The two subsystems perform the wetwell spray function by circulating water from the suppression pool through the RHR heat exchangers and returning it to a common wetwell spray sparger. The sparger only accommodates a small portion of the total RHR pump flow; the remainder of the flow returns to either the suppression pool through the suppression pool cooling return line, or can be routed to the drywell spray sparger.
Reactor Building Cooling Water (RCW) circulating through the shell side of the these exchangers, exchanges heat with the suppression pool water and discharges this heat to the external heat sink via the reactor service water (RSW) system.
Either RHR wetwell spray subsystem is sufficient to condense the steam from small bypass leaks from the drywell to the wetwell airspace during the postulated DBA.
APPLICABLE Reference 1 contains the results of analyses used to predict SAFETY ANALYSES primary containment pressure and temperature following large (continued)
ABWR TS B 3.6-1 P&R, 08/05/93 5:07pm
Containment System B 3.6.2.4 l
BASES APPLICABLE and small break loss of coolant accidents. The intent of the analyses is to demonstrate that the pressure reduction capacity of the RHR wetwell Spray System is adequate to maintain the primary containment conditions within design limits.
The time history for primary containment pressure is calculated to demonstrate that the maximum pressure remains below the design limit.
i The RHR wetwell spray system satisfies Criterion 3 of the i
NRC Policy Statement.
f LCO In the event of a DBA, a minimum of one RHR wetwell spray i
subsystem is required to mitigate potential bypass leakage paths and maintain the primary containment peak pressure below the design limits (Ref. I). To ensure that these requirements are met, two RHR wetwell spray subsystems must t
be OPERABLE with power from two safety related independent power supplies. Therefore, in the event of an accident, at least one subsystem is OPERABLE, assuming the worst case single active failure. An RHR wetwell spray subsystem is OPERABLE when the pump, the heat exchanger, and associated piping, valves, instrumentation, and controls are OPERABLE.
t APPLICABILITY In MODES I, 2, and 3, a DBA could cause heatup and pressurization of the primary containment.
In MODES 4 and 5, the probability and consequences of these events are reduced due to the pressure and temperature limitations in these MODES. Therefore, maintaining the RHR wetwell spray subsystems OPERABLE is not required in MODE 4 or 5.
ACTIONS A1 2
With one RHR wetwell spray subsystem inoperable, the inoperable subsystem must be restored to OPERABLE status within 7 days.
In this Condition, the remaining OPERABLE RHR wetwell spray subsystem is adequate to perform the primary containment bypass leakage mitigation function.
However, the overall reliability is reduced because a single failure in the OPERABLE subsystem cold result in reduced primary containment bypass mitigation capability. The g
(continued)
ABWR TS B 3.6-2 P&R, 08/05/93 5:07pm i
Containment System B 3.6.2.4 i
BASES (continued) i ACTIONS azl (continued) 7 day Completion Time was chosen in light of the redundant RHR wetwell spray capabilities afforded by the OPERABLE subsystem and the low probability of a DBA occurring during this period.
Ed With both RHR wetwell spray subsystems inoperable, at least i
one subsystem must be restored to OPERABLE status within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
In this Condition, there is a substantial loss of the primary containment bypass leakage mitigation function.
The 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> Completion Time is based on this loss of function and is considered acceptable due to the low probability of a DBA and because alternative methods to remove heat from primary containment are available.
C1 If the inoperable RHR wetwell spray subsystem cannot be i
restored to OPERABLE status within the associated Completion Time, the plant must be brought to a MODE in which the LC0 does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and to MODE 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
SURVEILLANCE SR 3.6.2.4.1 REQUIREMENTS Verifying the correct alignment for manual, power operated, and automatic valves in the RHR wetwell spray mode flow path provides assurance that the proper flow paths will exist for system operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position since these valves were verified to be in the correct position prior to leaking, sealing, or securing.
A valve is also allowed to be in the nonaccident position provided it can be aligned to t
(continued)
ABWR TS B 3.6-3 P&R, 08/05/93 5:07pm i
=.
4 Containment System B 3.6.2.4 BASES SURVEILLANCE SR 3.6.2.4.1 (continued)
REQUIREMENTS the accident position within the time assumed in the accident analysis. This is acceptable since the RHR suppression pool cooling mode is manually initiated. This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of being mispositioned are in the correct position.
This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves.
The Frequency of 31 days is justified because the valves are operated under procedural control, improper valve position would affect only a single subsystem, the probability of an event requiring initiation of the system is low, and the subsystem is a manually initiated system.
This Frequency has been shown to be acceptable based on operating i
experience.
SR 3.6.2.4.2 Verifying each associated RHR pump develops a flow rate 2 31.5 1/s (500 gpm) while operating in the wetwell spray mode with flow through the heat exchanger ensures that pump performance has not degraded during the cycle.
Flow is a normal test of centrifugal pump performance required by Section XI of the ASME Code (Ref. 2). This test confirms 4
one point on the pump design curve and is indicative of overall performance. Such inservice inspections 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, but the Frequency must not exceed 92 days.
REFERENCES 1.
2.
ASME, Boiler and Pressure Vessel Code,Section XI.
i i
ABWR TS B 3.6-4 P&R, 08/06/93 10:45am t
e i
Primary Containment Hydrogen Recombiners B 3.6.3.1 B 3.6 CONTAINMENT SYSTEMS B 3.6.3.1 Primary Containment dydrogen Recombiners BASES I
BACKGROUND The primary containment hydrogen recombiner eliminates the potential breach of primary containment due to a hydrogen oxygen reaction and is part of combustible gas control raquired by 10 CFR 50.44, " Standards for Combustible Gas Control Systems in Light-Water-Cooled Reactors" (Ref. 1),
and GDC 41, " Containment Atmosphere Cleanup" (Ref. 2). The primary containment hydrogen recombiner is required to reduce the hydrogen concentration in the primary containment following a loss of coolant accident (LOCA). The primary containment hydrogen recombiner accomplishes this by recombining hydrogen and oxygen to form water vapor. The vapor remains in the primary containment, thus eliminating any discharge to the environment. The primary containment hydrogen recombiner is manually initiated, since flammability limits would not be reached until several days after a Design Basis Accident (DBA).
The primary containment hydrogen recombiner functions to maintain the hydrogen gas concentration within the containment at or below the flammability limit of 4.0 volume percent (v/o) following a postulated LOCA.
It is fully redundant and consists of two 100% capacity subsystems.
Each primary containment hydrogen recombiner consists of an enclosed blower assembly, heater section, reaction chamber, direct contact water spray gas cooler, water separator, and associated piping, valves, and instruments.
The primary containment hydrogen recombiner will be manually initiated from the main control room when the hydrogen gas concentration in the primary containment reaches [3.3] v/o.
When the primary containment is inerted (oxygen concentration, 3.5 v/o), the crimary containment hydrogen i
recombiner will only function until the oxygen is used up (2.0 v/o hydrogen combines with 1.0 v/o oxygen). Two recombiners are provided to meet the requirement for redundancy and independence.
Each recombiner is powered from a separate Engineered Safety Feature bus and is provided with separate power panel and control panel, f
(continued)
ABWR TS B 3.6-1 P&R, 08/06/93 8:58am
s l
G G
t Primary Containment Hydrogen Recombiners B 3.6.3.1 i
i BASES (continued) i APPLICABLE The process gas circulating through the heater, the reaction SAFETY ANALYSES chambg/hr (150 scfm) by the use of an orifice plate r, and the cooler is automatically regulated to 255 m installed in the cooler. The process gas is heated to 718'C (1325'F). The hydrogen and oxygen gases are recombined into water vapor, which is then cor.densed in the water spray gas i
cooler by the associated residual heat removal subsystem and l
discharged with some of the effluent process gas to the i
suppression chamber. The majority of the cooled, effluent process gas is mixed with the incoming process gas to dilute the incoming gas prior to the mixture entering the heater section.
i The primary containment hydrogen recombiner provides the capability of controlling the bulk hydrogen concentration in primary containment to less than the lower flammable concentration of 4.0 v/o following a DBA. This control would prevent a primary containment wide hydrogen burn, thus ensuring that pressure and temperature conditions assumed in the analysis are not exceeded. The limiting DBA relative to hydrogen generation is a LOCA.
Hydrogen may accumulate in primary containment following a LOCA as a result of:
a.
A metal steam reaction between the zirconium fuel rod cladding and the reactor coolant; or b.
Radiolytic decomposition of water in the Reactor Coolant System.
To evaluate the potential for hydrogen accumulation in l
primary containment following a LOCA, the hydrogen generation is calculated as a function of time following the i
initiation of the accident. Assumptions recommended by Reference 3 are used to maximize the amount of hydrogen calculated.
t The calculation confirms that when the mitigating systems are actuated in accordance with emergency procedures, the peak hydrogen concentration in the primary containment is,
j 4.0 v/o (Ref. 4).
The primary containment hydrogen recombiners satisfy l
Criterion 3 of the NRC Policy Statement.
F (continued)
ABWR TS B 3.6-2 P&R, 08/06/93 8:58am J
i 6
l d
I
l Primary Containment Hydrogen Recombiners B 3.6.3.1 BASES (continued)
LC0 Two primary containment hydrogen recombiners must be OPERABLE.
This ensures operation of at least one primary containment hydrogen recombiner subsystem in the event of a worst case single active failure.
Operation with at least one primary containment hydrogen recombiner subsystem ensures that the post LOCA hydrogen concentration can be prevented from exceeding the flammability limit.
APPLICABILITY In MODES 1 and 2, the two primary containment hydrogen i
recombiners are required to control the hydrogen concentration within primary containment below its flammability limit of 4.0 v/o following a LOCA, assuming a worst case single failure.
l In MODE 3, both the hydrogen production rate and the total hydrogen produced after a LOCA would be less than that i
calculated for the DBA LOCA. Also, because of the limited time in this MODE, the probability of an accident requiring i
the primary containment hydrogen recombiner is low.
Therefore, the primary containment hydrogen recombiner is not required in MODE 3.
In MODES 4 and 5, the probability and consequences of a LOCA are low due to the pressure and temperature limitations in these MODES. Therefore, the primary containment hydrogen recombiner is not required in these MODES.
ACTIONS A.1 With one primary containment hydrogen recombiner inoperable, I
the inoperable recombiner must be restored to OPERABLE status within 30 days.
In this Condition, the remaining OPERABLE recombiner is adequate to perform the hydrogen control function. However, the overall reliability is reduced because a single failure in the OPERABLE recombiner could result in reduced hydrogen control capability. The 30 day Completion Time is based on the low probability of the occurrence of a LOCA that would generate hydrogen in amounts capable of exceeding the flammability limit, the amount of time available after the event for operator action (continued)
ABWR TS B 3.6-3 P&R, 08/06/93 8:58am l
l
^
(
Primary Containment Hydrogen Recombiners B 3.6.3.1 BASES ACTIONS
.A_d (continued) to prevent exceeding this limit, and the low probability of failure of the OPERABLE primary containment hydrogen r
recombiner.
Required Action A.1 has been modified by a Note indicating that the provisions of LC0 3.0.4 are not applicable.
As a result, a MODE change is allowed when one recombiner is inoperable. This allowance is provided because of the low probability of the occurrence of a LOCA that would generate hydrogen in amounts capable of exceeding the flammability limit, the low probability of the failure of the OPERABLE subsystem, and the amount of time available after a postulated LOCA for operator action to prevent exceeding the flammability limit.
B.1 and B.2 With two primary containment hydrogen recombiners inoperable, the ability to perform the hydrogen control function via alternate capabilities must be verified by administrative means within I hour.
The alternate hydrogen control capabilities are provided by the atmospheric control system (ACS). The 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Completion Time allows a reasonable period of time to verify that a loss of hydrogen l
control function does not exist.
Both the initial verification and all subsequent verifications may be performed as an administrative check by examining logs or i
other information to determine the availability of the i
alternate hydrogen control system.
It does not mean to t
perform the Surveillances needed to demonstrate OPERABILITY of the alternate hydrogen control system.
If the ability to perform the hydrogen control function is maintained, continued operation is permitted with two hydrogen recombiners inoperable for up to 7 days.
Seven days is a reasonable time to allow two hydrogen recombiners to be inoperable because the hydrogen control function is maintained and because of the low probability of the occurrence of a LOCA that would generate hydrogen in the amounts capable of exceeding the flammability limit.
F i
(continued)
ABWR TS B 3.6-4 P&R, 08/06/93 8:58am O
I e
s Primary Containment Hydrogen Recombiners B 3.6.3.1 BASES t
ACTIONS C.1 (continued)
If any Required Action and associated Completion Time cannot i
be met, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The allowed Completion Time of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-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.
SURVEILLANCE SR 3.6.3.1.1 REQUIREMENTS Performance of a system functional test for each primary containment hydrogen recombiner ensures that the recombiners are OPERABLE and can attain and sustain the temperature necessary for hydrogen recombination.
In particular, this SR verifies that the minimum heater sheath temperature increases to 2 718'C (1325'F) in 5 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> s and that it is maintained > 690*C (1275'F) and < 746' (1375'F) for 2 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> thereafter to check the ability of the recombiner to function properly (and to make sure that significant i
heater elements are not burned out).
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.
SR 3.6.3.1.2 This SR ensures there are no physical problems that could affect recombiner operation.
Since the recombiners are mechanically passive, except for the blower assemblies, they are subject to only minimal mechanical failure. The only credible failures involve loss of power or blower function, blockage of the internal flow path, missile impact, etc. A visual inspection is sufficient to determine abnormal conditions that could cause such failures.
Operating experience has shown that these components usually pass the Surveillance when performed at the 18 month 4
(continued)
ABWR TS B 3.6-5 P&R, 08/06/93 10:48am
Primary Containment Hydrogen Recombiners i
B 3.6.3.1 BASES I
SURVEILLANCE SR 3.6.3.1.2 (continued)
REQUIREMENTS Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.
This SR requires performance of a resistance to ground test of each heater phase to make sure that there are no detectable grounds in any heater phase. This is accomplished by verifying that the resistance to ground for any heater phase is 2 [10,000] ohms.
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.
~
i REFERENCES 1.
2.
10 CFR 50, Appendix A, GDC 41.
3.
Regulatory Guide 1.7, Revision 1.
4.
i i
[
I ABWR TS B 3.6-6 P&R, 08/06/93 8:58am
e 4
t Containment Systems B 3.6.3.2 l
B 3.6 CONTAINMENT SYSTEMS B 3.6.3.2 Primary Containment and Oxygen Concentration BASES BACKGROUND All nuclear reactors must be designed to withstand events that generate hydrogen either due to the zirconium metal water reaction in the core or due to radiolysis.
t APPLICABLE The primary method to control hydrogen is to inert the SAFETY ANSLYSES primary containment. With the primary containment inert, that is, oxygen concentration < 3.5 volume percent (v/o), a combustible mixture cannot be present in the primary containment for any hydrogen concentration.
The capability to inert the primary containment and maintain oxygen < 3.5 v/o works together with the hydrogen recombiners (LCO 3.6.3.1, " Primary Containment Hydrogen Recombiners") to provide redundant and diverse methods to mitigate events i
that produce hydrogen.
For example, an event that rapidly generates hydrogen from zirconium metal water reaction will result in excessive hydrogen in primary containment, but oxygen concentration will remain < 3.5 v/o and no combustion i
can occur.
Long term generation of both hydrogen and oxygen from radiolytic decomposition of water may eventually result in a combustible mixture in primary containment, except that the hydrogen recombiners remove hydrogen and oxygen gases faster than they can be produced from radiolysis and again 7
no combustion can occur. This LC0 ensures that oxygen concentration does not exceed 3.5 v/o during operation in the applicable conditions.
The Reference I calculations assume that the primary containment is inerted when a Design Basis Accident loss of coolant accident occurs. Thus, the hydrogen assumed to be released to the primary containment as a result of metal water reaction in the reactor core will not produce combustible gas mixtures in the primary containment.
0xygen, which is subsequently generated by radiolytic decomposition of water, is recombined by the hydrogen recombiners (LC0 3.6.3.1) more rapidly than it is produced.
Primary containment oxygen concentration satisfies Criterion 2 of the NRC Policy Statement.
(continued) l l
ABWR TS B 3.6-1 P&R, 08/06/93 10:57am l
i i
Containment Systems B 3.6.3.2 BASES i
LCO The primary containment oxygen concentration is maintained
< 3.5 v/o to ensure that an event that produces any amount i
of hydrogen does not result in a combustible mixture inside primary containment.
1 APPLICABILITY The primary containment oxygen concentration must be within the specified limit when primary containment is inerted, except as allowed by the relaxations during startup and shutdown addressed below. The primary containment must be inert in MODE 1, since this is the condition with the highest probability of an event that could produce hydrogen.
Inerting the primary containment is an operational problem because it prevents containment access without an appropriate breathing apparatus. Therefore, the primary containment is inerted as late as possible in the plant startup and de-inerted as soon as possible in the plant shutdown. As long as reactor power is < 15% RTP, the i
potential for an event that generates significant hydrogen is low and the primary containment need not be inert.
Furthermore, the probability of an event that generates hydrogen ucurring within the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of a startup, or within the last 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> before a shutdown, is low enough that these " windows," when the primary containment is not inerted, are also justified. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> time period is a reasonable amount of time to allow plant personnel to perform inerting or de-inerting.
ACTIONS A.1 If oxygen concentration is 2 3.5 v/o at any time while operating in MODE 1, with the exception of the relaxations allowed during startup and shutdown, oxygen concentration must be restored to < 3.5 v/o within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Completion Time is allowed when oxygen concentration is 2 3.5 v/o because of the availability of other hydrogen mitigating systems (e.g., hydrogen recombiners) and the low probability and long duration of an event that would generate significant amounts of hydrogen occurring during this period.
(continued)
ABWR TS B 3.6-2 P&R, 08/06/93 10:56am i
a o
Containment Systems i
B 3.6.3.2 I
BASES ACTIONS Bl i
(continued)
If oxygen concentration cannot be restored to within limits I
within the required Completion Time, the plant must be brought to a MODE in which the LC0 does not apply. To achieve this status, power must be reduced to s 15% RPT within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. The 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> Completion Time is reasonable, based on operating experience, to reduce reactor power from f
full power conditions in an orderly manner and without challenging plant systems.
SURVEILLANCE SR 3.6.3.2.1 REQUIREMENTS The primary containment must be determined to be inert by verifying that oxygen concentration is < 3.5 v/o. The 7 day Frequency is based on the slow rate at which oxygen concentration can change and on other indications of abnormal conditions (which would lead to more frequent checking by operators in accordance with plant procedures).
Also, this Frequency has been shown to be acceptable through operating experience.
REFERENCES 1.
l i
l l
~
l t
ABWR TS B 3.6-3 P&R, 08/06/93 10:21am i
Containment Systems I
B 3.6.4.1 i
B 3.6 CONTAINMENT SYSTEMS B 3.6.4.1 Secondary Containment BASES BACKGROUND The function of the [ secondary containment) is to contain, dilute, and hold up fission products that may leak from
{
primary containment following a Design Basis Accident (DBA).
In conjunction with operation of the Standby Gas Treatment (SGT) System and closure of certain valves whose lines penetrate the [ secondary containment], the [ secondary containment] is designed to reduce the activity level of the fission products prior to release to the environment and to isolate and contain fission products that are released during certain operations that take place inside primary containment, when primary containment is not required to be OPERABLE, or that take place outside primary containment.
The [ secondary containment] is a structure that completely encloses the primary containment and those components that i
may be postulated to contain primary system fluid. This structure forms a control volume that serves to hold up and dilute the fission products.
It is possible for the pressure in the control volume to rise relative to the environmental pressure (e.g., due to pump and motor heat load additions). To prevent ground level exfiltration while allowing the [ secondary containment] to be designed as a t
conventional structure, the [ secondary containment] requires support systems to maintain the control volume pressure at less than the external pressure.
Requirements for these systems are specified separately in LC0 3.6.4.2, " Secondary Containment Isolation Valves (SCIVs)," and LCO 3.6.4.3,
" Standby Gas Treatment (SGT) System."
i APPLICABLE There are two principal accidents for which credit is SAFETY ANALYSES taken for [ secondary containment] OPERABILITY These are a loss of coolant accident (LOCA) (Ref.1) and a fuel handling accident inside secondary containment (Ref. 2). The
[ secondary containment) performs no active function in response to each of these limiting events; however, its leak
+
tightness is required to ensure that the release of radioactive materials from the primary containment is restricted to those leakage paths and associated leakage rates assumed in the accident analysis, and that fission i
(continued) r ABWR TS B 3.6-1 P&R, 08/06/93 12:18pm
a Containment Systems B 3.6.4.1 BASES APPLICABLE products entrapped within the [ secondary containment]
SAFETY ANALYSES structure will be treated by the SGT System prior to (continued) discharge to the environment.
[ Secondary containment] satisfies Criterion 3 of the NRC Policy Statement.
LCO An OPERABLE [ secondary containment] provides a control volume into which fission products that bypass or leak from primary containment, or are released from the reactor coolant pressure boundary components located in [ secondary containment], can be diluted and processed prior to release to the environment.
For the [ secondary containment] to be considered OPERABLE, it must have adequate leak tightness to ensure that the required vacuum can be established and maintained.
APPLICABILITY In MODES 1, 2, and 3, a LOCA could lead to a fission product release to primary containment that leaks to [ secondary containment]. Therefore, [ secondary containment]
l OPERABILITY is required during the same operating conditions that require primary containment OPERABILITY.
In MODES 4 and 5, the probability and consequences of the LOCA are reduced due to the pressure and temperature limitations in these MODES. Therefore, maintaining
[ secondary containment] OPERABLE is not required in MODE 4 or 5 to ensure a control volume, except for other situations for which significant releases of radioactive material can be postulated, such as during operations with a potential for draining the reactor vessel (0PDRVs), during CORE ALTERATIONS, or during movement of irradiated fuel assemblies in the [ primary or secondary containment].
ACTIONS 621 If [ secondary containment] is inoperable, it must be restored to OPERABLE status 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 provides a period of time to correct the problem that is commensurate with the importance of (continued)
ABWR TS B 3.6-2 P&R, 08/05/93 12:18pm
Containment Systems B 3.6.4.1 BASES r
ACTIONS A,1 (continued) maintaining [ secondary containment] during MODES 1, 2, and 3.
This time period also ensures that the probability of an accident (requiring [ secondary containment]
l OPERABILITY) occurring during periods where [ secondary containment] is inoperable is minimal.
T t
B.1 and B.2 If [ secondary containment] cannot be restored to OPERABLE status within the required Completion Time, the plant must be brought to a MODE in which the LC0 does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and to MODE 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without t
challenging plant systems.
C.I. C.2. and C.3 Movement of irradiated fuel assemblies in the [ secondary containment], CORE ALTERATIONS, and OPDRVs can be postulated to cause fission product release to the [ secondary containment].
In such cases, the [ secondary containment] is the only barrier to release of fission products to the environment. CORE ALTERATIONS and movement of irradiated 2
fuel assemblies must be immediately suspended if the i
[ secondary containment] is inoperable.
1 Suspension of these activities shall not preclude completing an action that involves moving a component to a safe position. Also, action must be immediately initiated to suspend OPDRVs to minimize the probability of a vessel draindown and subsequent potential for fission product release. Actions must continue until OPDRVs are suspended.
Required Action C.1 has been modified by a Note stating that LC0 3.0.3 is not applicable.
If moving irradiated fuel assemblies while in MODE 4 or 5, LC0 3.0.3 would not specify l
any action.
If moving irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations.
Therefore, in either case, inability to suspend (continued)
ABWR TS B 3.6-3 P&R, 08/06/93 12:18pm 3
l i
n Containment Systems l
B 3.6.4.1 BASES ACTIONS C.). C.2. and C.3 (continued)
I movement of irradiated fuel assemblies would not be a sufficient reason to require a reactor shutdown.
l SURVEILLANCE SR 3.6.4.1.1 REQUIREMENTS This SR ensures that the [ secondary containment] boundary is sufficiently leak tight to preclude exfiltration under expected wind conditions. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Frequency of this SR was developed based on operating experience related to
[ secondary containment] vacuum variations during the applicable MODES and the low probability of a DBA occurring between surveillances.
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
[ secondary containment] vacuum condition.
SR 3.6.4.1.2 and SR 3.6.4.1.3 Verifying that [ secondary containment] equipment hatches and access door: are closed ensures that the infiltration of outside air of such a magnitude as to prevent maintaining the desired negative pressure does not occur. Verifying that all such openings are closed provides adequate assurance that exfiltration from the [ secondary containment]
will not occur. Maintaining [ secondary containment]
OPERABILITY requires verifying each door in the access opening is closed, except when the access opening is being used for normal transient entry and exit (then, at least one door must remain closed). The 31 day Frequency for these SRs has been shown to be adequate, based on operating experience, and is considered adequate in view of the cther indications of door and hatch status that are available to the operator.
SR 3.6.4.1.4 and SR 3.6.4.1.5 The SGT System exhausts the secondary containment atmosphere to the environment through appropriate treatment equipment.
(continued)
ABWR TS B 3.6-4 P&R, 08/06/93 5:08pm l
r
O r
Containment Systems B 3.6.4.1 BASES SURVEILLANCE SR 3.6.4.1.4 and SR 3.6.4.1.5 (continued)
REQUIREMENTS To enstre that all fission products are treated, SR 3.6.4.1.4 verifies that the SGT System will rapidly establish and maintain a pressure in the [ secondary containment] that is less than the lowest postulated pressure external to the [ secondary containment] boundary.
This is confirmed by demonstrating that one SGT subsystem 3
will draw down the [ secondary containment] to 2 6.34 Kg/m (0.25 inches) incher of vacuum water gauge in s [120] seconds.
This cannot be accomplished if the
[ secondary containment] boundary is not intact.
SR3.6.4.1.5demonstpatesthatoneSGTsubsystemcan maintain 2 6.34 Kg/m (0.25 inches of vacuum water gauge for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> at a flow rate s 6800 m}/h (4000 cfm).
The I hour test period allows [ secondary containment] to be in thermal equilibrium at steady state conditions.
Therefore, these two tests are used to ensure [ secondary containment]
boundary integrity.
Since these SRs are [ secondary containment] tests, they need not be performed with each SGT subsystem.
The SGT subsystems are tested on a STAGGERED TEST BASIS, however, to ensure that in addition to the requirements of LC0 3.6.4.3, either SGT subsystem will perform this test.
The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outaga and the potential for an unplanned transient if the Surveillance were performed with 5
the reactor at power. Operating experience has shown 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.
2.
i ABWR TS B 3.6-5 P&R, 08/06/93 5:09pm
e Containment Systems B 3.6.4.2 B 3.6 CONTAINMENT SYSTEtiS B 3.6.4.2 Secondary Containment Isolation Valves (SCIVs)
BASES BACKGROUND The function of tSe SCIVs, in combination with other accident mitigation systems, is to limit fission product release during and following postulated Design Basis Accidents (DBAs) (Ref. 1).
Secondary containment isolation within the time limits specified for those isolation valves designed to close automatically ensures that fission products that leak from primary containment following a DBA, or that are released during certain operations when primi.ry containment is not required to be OPERABLE, or take place outside primary containment, are maintained within applicable limits.
The OPEPABILITY requirements for SCIVs help ensure that adequate secondary containment leak tightness is maintained during and after an accident by minimizing potential paths to the environment. These isolation devices consist of either passive devices or active (automatic) devices.
Manual valves, de-activated automatic valves secured in their closed position (including check valves with flow through the valve secured), and blind flanges are considered passive devices.
Automatic SCIVs close on a secondary containment isolation signal to prevent leakage of untreated radioactive material from secondary containment following a DBA or other accidents.
Other penetrations are isolated by the use of valves in the closed position or blind flanges.
APPLICABLE The SCIVs must be OPERABLE to ensure that secondary SAFETY ANALYSES containment is a leak tight barrier to fission product i
releases. The principal accidents for which secondary l
containment leak tightness is required are a loss of coolant accident (Ref. 1), and a fuel handling accident inside secondary containment (Rr#. 2). The secondary containment performs no active function in y,;ponse to either of these limiting events, but its leaA tie tness is required to (continued)
ABWR TS B 3.6-1 P&R, 08/06/93 5:15pm i
a r
Containment Systems i
B 3.6.4.2 BASES i
APPLICABLE ensure that leakage from the primary containment is SAFETY ANALYSES processed by the Standby Gas Treatment (SGT) System before s
(contiued) being released to the environment.
j Maintaining SCIVs OPERABLE with isolation times within I
limits ensures that fission products will remain trapped inside secondary containment so that they can be treated by the SGT System prior to discharge to the environment.
l SCIVs satisfy Criterion 3 of the NRC Policy Statement.
l LCD SCIVs form a part of the secondary containment boundary. The l
SCIV safety function is related to control of offsite radiation releases resulting from DBAs.
The automatic power operated isolatian valves are considered OPERABLE when their isolation times are within limits and the valves actuate on an autonatic isolation signal. The i
valves covered by this LCO, along with their associated stroke times, are listed in Reference 3.
The normally closed isolation valves or blind flanges are considered OPERABLE when manual valves are closed, automatic i
valves are de-activated and secured in their closed position, and blind flanges are in place.
These passive isolation valves or devices are listed in Reference 3.
l APPLICABILITY In MODES I, 2, and 3, a DBA could lead to a fission product release to the primary containment that leaks to the secondary containment. Therefore, the OPERABILITY of SCIVs is required.
In MODES 4 and 5, the probability and consequences of these events are reduced due to pressure and temperature limitations in these MODES. Therefore, maintaining SCIVs OPERABLE is not required in MODE 4 or 5, except for other situations under which significant radioactive releases can be postulated, such as during operations with a potential for draining the reactor vessel (OPDRVs), during I
t t
(continued)
ABWR TS B 3.6-2 P&R, 08/06/93 5:15pm
r e
Containment Systems B 3.6.4.2 BASES APPLICABILITY CORE ALTERATIONS, or during movement of irradiated fuel (continued) assemblies in the secondary containment. Moving irradiated fuel assemblies in the secondary containment may also occur in MODES 1, 2, and 3.
r ACTIONS The ACTIONS are modified by three Notes. The first Note allows penetration flow paths to be unisolated intermittently under administrative controls. These controls consist of stationing a dedicated operator, who is in continuous communication with the control room, at the controls of the valve.
In this way, the penetration can be rapidly isolated when a need for secondary containment isolation is indicated.
The second Note provides clarification that for the purpose of this LCO separate Condition entry is allowed for each penetration flow path.
The third Note ensures appropriate remedial actions are f
taken, if neces.ary, if the affected system (s) are rendered j
inoperable by an inoperable SCIV.
i r
1 A.I and A.2
^
In the event that there are one or more penetration flow paths with one SCIV inoperable, the affected penetration l
flow path (s) must be isolated. The method of isolation must i
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 SCIV, a closed manual valve, and a blind flange.
For penetrations isolated in accordance with l
Required Action A.1, the valve used to isolate the penetration should be the closest available valve to secondary containment.
The Required Action must be completed within the 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> Completion Time. The specified time period is reasonable considering the time required to isolate the penetration, and the probability of a DBA, which requires the SCIVs to close, occurring during this short time, is very low.
For affected penetrations that have been isolated in accordance with Required Action A.1, the affected (continued) i ABWR TS B 3.6-3 P&R, 08/06/93 5:15pm I
~.
l
~
r Containment Systems B 3.6.4.2 BASES ACTIONS A.1 and A.2 (continued) penetration must be verified to be isolated on a periodic basis. This is necessary to ensure that secondary containment penetrations required to be isolated following an accident, but no longer capable of being automatically isolated, will be in the isolation position should an event occur. The Completion Time of once per 31 days is appropriate because the valves are operated under administrative controls and the probability of their misalignment is low. This Required Action does not require any testing or valve manipulation.
Rather, it involves verification that the affected penetration remains isolated.
Required Action A.2 is modified by a Note that applies to valves and blind flanges located in high radiation areas and allows them to be verified closed by use of administrative controls. Allowing verification by administrative controls is considered acceptable, since access to these areas is typically restricted.
Therefore, the probability of misalignment of these valves, once they have been verified to be in the proper position, is low.
Bil 1
With two SCIVs in one or more penetration flow paths inoperable, the affected penetration flow path must be isolated within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The method of isolation must i
d include the use of at least one isolation barriar that l
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 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 probability of a DBA, which requires the SCIVs to close, occurring during this short time, is very low.
The Condition has been modified by a Note stating that Condition B is only applicable to penetration flow paths with two isolation valves.
This clarifies that only Condition A is entered if one SCIV is inoperable in each of two penetrations.
I 1
(continued)
ABWR TS B 3.6-4 P&R, 08/06/93 5:15pm 4
c.
Containment Systems i
B 3.6.4.2 BASES i
ACTIONS C.1 and C.2 (continued)
If any Required Action and associated Completion Time cannot be cet, the plant must be brought to a MODE in which the LC0 l
does net apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and to MODE 4 l
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 3
orderly manner and without challenging plant systems.
D l. D.2. and D.3 If any Required Action and associated Completion Time are not be met, the plant must be placed in a condition in which the LC0 does not apply.
If applicable, CORE ALTERATIONS and the movement of irradiated fuel assemblies in the secondary i
containment must be immediately suspended. Suspension of these activities shall not preclude completion of movement of a component to a safe position. Also, if applicable, C
i actions must be immediately initiated to suspend OPDRVs in j
order to minimize the probability of a vessel draindown and the subsequent potential for fission product release.
j Actions must continue until OPDRVs are suspended.
t I
Required Action D.1 has been modified by a Note stating that LCO 3.0 't is not applicable.
If moving irradiated fuel assemblies while in MODE 4 or 5, LCO 3.0.3 would not specify any action.
If moving fuel while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations.
Therefore, in either case, inability to suspend movement of irradiated fuel assemblies would not be a sufficient reason I
to require a reactor shutdown.
SURVEILLANCE SR 3.6.4.2.1 l
REQUIREMENTS This SR verifies that each secondary containment manual isolation valve and blind flange that is required to be closed during accident conditions is closed.
The SR helps i
to ensure that post accident leakage of radioactive fluids or gases outside of the secondary containment boundary is within design limits. This SR does not require any testing i
or valve manipulation. Rather, it involves verification (continued)
ABWR TS B 3.6-5 P&R, 08/06/93 5:17pm
r Containment Systems B 3.6.4.2 j
BASES SURVEILLANCE SR 3.6.4.2.1 (continued)
REQUIREMENTS that those valves in secondary containment that are capable of being mispositioned are in the correct position.
Since these valves are readily accessible to personnel during i
normal unit operation and verification of their position is relatively easy, the 31 day Frequency was chosen to provide 3
added assurance that the valves are in the correct positions.
Two Notes have been added to this SR. The first Note applies to valves and blind flanges located in high radiation areas and allows them to be verified by use of administrative controls. Allowing verification by administrative controls is considered acceptable, since access to these areas is typically restricted during MODES 1, 2, and 3 for ALARA reasons. Therefore, the probability of misalignment of these valves, once they have l
been verified to be in the proper position, is low.
A second Note has been included to clarify that SCIVs that are open under administrative centrols are not required to meet the SR during the time the valves ara apen.
SR 3.6.4.2.2 Verifying that the isolation time of each power operated and each automatic SCIV is within limits is required to demonstrate OPERABILITY. The isolation time test ensures j
that 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.
SR 3.6.4.2.3 Verifying that each automatic SCIV closes on a secondary I
i containment isolation signal is required to prevent leakage of radioactive material from secondary conta1riment following a DBA or other accidents. This SR ensures that each automatic SCIV will actuate to the isolation position on a secondary containment isolation signal. The LOGIC SYSTEM FUNCTIONAL TEST in SR 3.3.6.2.5 overlaps this SR to provide complete testing of the safety function. The i
(continued) 1 ABWR TS B 3.6-6 P&R, 08/06/93 5:15pm a
I r
Containment Systems B 3.6.4.2 BASES J
SURVEILLANCE SR 3.6.4.2.3 (continued)
REQUIREMENTS
[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 shcwn 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.
2.
3.
r l
l i
i i
a
?
1 i
i j
i
.l l
l i
ABWR TS B 3.6-7 P&R, 08/06/93 5:15pm
l r
I Containment Systems B 3.6.4.3 t
B 3.6 CONTAINMENT SYSTEMS B 3.6.4.3 Standby Gas Treatment (SGT) System
)
i BASES BACKGROUND The SGT System is required by 10 CFR 50, Appendix A, GDC 41,
" Containment Atmosphere Cleanup" (Ref. 1). The function of the SGT System is to ensure that radioactive materials that leak from the primary containment into the secondary I
containment following a Design Basis Accident (DBA) are filtered and adsorbed prior to exhausting to the i
environment.
The SGT System consists the following components:
a.
Two 100 percent capacity charcoal filter trains, each i
consisting of (components listed in order of air flow
{
direction)-
1.
a moisture separator; j
l 2.
an electric heaters; 3.
a prefilter; 4.
a pre-high efficiency particulate air (HEPA) t filter; 5.
a space heater; 6.
a charcoal adsorber; 7.
a spare heater; 8.
a post-HEPA filter; and b.
Two fully redundant subsystems, each with its own ductwork, flow element, dampers, and instrumentation
)
controls, consisting of i
1.
a process fan and i
2.
a cooling fan.
I 1
(continued)
ABWR TS B 3.6-1 P&R, 08/06/93 2:21pm
i
~
r i
Containment System B 3.6.4.3 i
BASES BACKGROUND The sizing of the SGT System equipment and components is (continued) based on the results of an infiltration analysis, as well as an exfiltration analysis of the secondary containment. The internal pressure of the SGT System boundary region is maintained at a negative pressure of 6.35 mm (0.25 inches) water gauge relative to the surrounding spaces when the system is in operation, which represents the internal pressure required to ensure zero exfiltration of air from the building when exposed to an 8.9 m/s (20 mph) wind blowing at an angle of 45* to the building. The continuous negative differential pressure is established within 10 r
minutes after SGT System initiation.
The moisture separator is provided to remove entrained water in the air, while the electric heater reduces the relative humidity of the influent airstream to the adsorber section of the filter train to less than 70% whenever SGT System is in operation (Ref. 2). The prefilter removes large particulate matter, while the HEPA filter removes fine j
particulate matter and protects the charcoal from fouling.
The charcoal adsorber removes gaseous elemental iodine and l
organic iodides, and the final HEPA filter collects any
[
carbon fines exhausted from the charcoal adsorber.
The SGT System automatically starts and operates in response to actuation signals indicativa of conditions or an accident that could require operation of the system.
Following initiation, both SGT System train process fans start.
Upon verification that both trains are operating, one of the redundant trains is normally shut 17wn.
APPLICABLE The design basis for the SGT System is to mitigate the SAFETY ANALYSES consequences of a loss of coolant accident and fuel handling 4
accidents (Ref. 2).
For all events analyzed, the SGT System i
is shown to be automatically initiated to reduce, via I
filtration and adsorption, the radioactive material released to the environment.
The SGT System satisfies Criterion 3 of the NRC Policy Statement.
(continued)
ABWR TS B 3.6-2 P&R, 08/06/93 2:21pm
a e
Containment System B 3.6.4.3 BASES LCO Following a DBA, a minimum of one SGT System train is required to maintain the secondary containment at the required negative pressure with respect to the surrounding spaces within 10 minutes of its initiation, and to process l
gaseous releases. Meeting the LC0 requirements for two OPERABLE trains ensures operation of at least one SGT System train in the event of a single active failure.
l APPLICABILITY In MODES 1, 2, and 3, a DBA could lead to a fission product i
release to primary containment that leaks to secondary containment.
Therefore, SGT System OPERABILITY is required during these MODES.
In MODES 4 and 5, the probability and consequences of these events are reduced due to the pressure and temperature limitations in these MODES. Therefore, maintaining the SGT l
System in OPERABLE status is not required in MODE 4 or 5,
[
except for other situations under which significant releases of radioactive material can be postulated, such as during operations with a potential for draining the reactor vessel (0PDRVs), during CORE ALTERATIONS, or during movement of irradiated fuel assemblies in the secondary containment.
l ACTIONS A.]
2 With one SGT train inoperable, the inoperable train must be restored to OPERABLE status in 7 days.
In this condition, the remaining OPERABLE SGT train is adequate to perform the l
required radioactivity release control function. However, the overall system reliability is reduced because a single active failure in the OPERABLE train could result in the radioactivity release control function not being adequately performed.
The 7 day Completion Time is based on consideration of such factors as the availability of the OPERABLE redundant SGT System train and the low probability f
of a DBA occurring during this period.
B.1 and B.2 If the SGT System train cannot be restored to OPERABLE status within the required Completion Time in MODE 1, 2, (continued)
ABWR TS B 3.6-3 P&R, 08/06/93 5:19pm
l O
r Containment System B 3.6.4.3 BASES i
ACTIONS B.1 and B.2 (continued) or 3, the plant must be brought to a MODE in which the LC0 i
does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and to MODE 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
i t
i C.I. C.2.1. C.2.2. and C.2.3 l
During movement of irradiated fuel assemblies, in the
[ secondary containment], during CORE ALTERATIONS, or during OPDRVs, when Required Action A.1 cannot be completed within the required Completion Time, the OPERABLE SGT System train should immediately be placed in operation. This action ensures that the remaining train is OPERABLE, that no l
failures that could prevent automatic actuation have i
occurred, and that any other failure would be readily detected.
j An alternative to Required Action C.1 is to immediately
)
suspend activities that represent a potential for releasing radioactive materia? to the secondary containment, thus placing the plant in a condition that minimizes risk.
If applicable, CORE ALTERATIONS and movement of irradiated fuel assemblies must iinmediately be suspended.
Suspension of these activities must not preclude completion of movement of a component to a safe position. Also, if applicable, actions must immediately be initiated to suspend OPDRVs in order to minimize the probability of a vessel draindown and subsequent potential for fission product release. Actions must continue until OPDRVs are suspended.
The Required Actions of Condition C have been modified by a Note stating that LC0 3.0.3 is not applicable.
If moving irradiated fuel assemblies while in MODE 4 or 5, LCO 3.0.3 would not specify any action.
If moving irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations.
Therefore, in either i
i 1
(continued)
ABWR TS B 3.6-4 P&R, 08/06/93 2:21pm W
s' e-Containment System B 3.6.4.3 BASES r
SURVEILLANCE SR C.l. C.2.1. C.2.2. and C.2.3 (continued) t REQUIREMENTS case, inability to suspend movement of irradiated fuel assemblies would not be a sufficient reason to require a reactor shutdown.
D.I. D.2. and 0.3 When both SGT System trains are inoperable, if applicable, CORE ALTERATIONS and movement of irradiated fuel assemblies in [ secondary containment] must be immediately suspended.
Suspension of these activities shall not preclude completion of movement of a component to a safe position. Also, if applicable, actions must be immediately be initiated to suspend CPDRVs in order to minimize the probability of a vessel draindown and subsequent potential for fission product release. Actions must continue until OPDRVs are suspended.
Required Action D.1 has been modified by a Note stating that LCO 3.0.3 is not applicable.
SURVEILLANCE SR 3.6.4.3.1 REQUIREMENTS Operating each SGT System for 2 [10] continuous hours ensures that both subsystems 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 i
temperature)] for 2 [10] continuous hours every 31 days eliminates moisture on the adsorbers and HEPA filters. The 31 day Frequency was developed in consideration of the known reliability of fan motors and controls and the redundancy available in the system.
SR 3.6.4.3.2 This SR verifies that the required SGT System filter testing is performed in accordance with the Ventilation Filter Testing Program (VFTP). The SGT System filter tests are in accordance with Regulatory Guide 1.52 (Ref. 4). The VFTP (continued)
ABWR TS B 3.6-5 P&R, 08/06/93 2:21pm
\\m Containment System B 3.6.4.3 BASES SURVEILLANCE SR 3.6.4.3.2 (continued)
REQUIREMENTS 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).
Specified test frequencies and additional information are discussed in detail in the VFTP.
SR 3.6.4.3.3 This SR requires verification that each SGT subsystem starts upon receipt of an actual or simulated initiation signal.
The LOGIC SYSTEM FUNCTIONAL TEST in SR 3.3.6.2.5 overlaps this SR to provide complete testing of the safety function.
While this Surveillance can be performed with the reactor at power, operating experience has shown these components usually pass the Surveillance 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.
SR 3.6.4.3.4 This SR requires verification that the SGT System filter cooler bypass damper can be opened and the fan started.
This ensures that the ventilation mode of SGT System operation is available. While this Surveillance can be performed with the reactor at power, operating experience has shown these components usually pass the Surveillance 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.
REFERENCES 1.
10 CFR 50, Appendix A, GDC 41.
2.
(continued)
ABWR TS B 3.6-6 P&R, 08/06/93 2:21pm
\\ c.
I, Containment System B 3.6.4.3 BASES REFERENCES 3.
(continued) 4.
Regulatory Guide 1.52, Rev. [2].
ABWR TS B 3.6-7 P&R, 08/06/93 2:21pm
UHS i
3.7.1
,r 3.7 PLANT SYSTEMS 3.7.1 Ultimate Heat Sink (UHS)
LCO 3.7.1 UHS shall be OPERABLE.
APPLICABILITY:
MODES 1, 2, and 3.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME 4
A. One or more [ UHS A.1 Restore [ UHS active 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> i
active components]
components] to inoperable.
OPERABLE status.
QB l
A.2 Declare associated 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> RCW/RSW subsystems inoperable.
B.
Required Action and B.1 Be in MODE 3.
12 1ours I
associated Completion i
Time of Condition A AND not met.
B.2 Be in MODE 4.
36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> l
Q3 UHS inoperable for reasons other than Condition A.
(continued) i l
ABWR TS 3.7-1 P&R, 7/30/93
r 4
UHS i
3.7.1 L
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.1.1 Verify the water level of each [ UHS 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> dedicated supply] is [ equivalent to 30 days supply volume].
SR 3.7.1.2 Verifytheaveragewatertemperatureofthe 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> UHS is s 35"C (95 F).
SR 3.7.1.3 Operate each [ UHS active component not 31 days normally operating] for 2 [15] minutes.
SR 3.7.1.4 Verify each [ UHS active components not 18 months normally operating] actuates on an actual or simulated initiation signal.
1 4
i ABWR TS 3.7-2 P&R,7/30/93
RCW/RSW System 3.7.2 3.7 PLANT SYSTEMS 3.7.2 Reactor Cooling Water (RCW) System and Reactor Service Water (RSW)
System l'
LCO 3.7.2 Division A, B and C RCW and RSW subsystems shall be OPERABLE.
APPLICABILITY:
MODES I, 2, and 3.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME i
A. One subsystem
NOTES--------
I. Enter applicable Conditions and Required Actions of LCO 3.8.I, "AC Sources-Operating," for diesel generator made inoperable by [RSW).
- 2. Enter applicable Conditions and Required Actions of LC0 3.4.9,
Shutdown Cooling System-Hot Shutdown," for
[RHR shutdown cooling] made inoperable by i
[RCW or RSW].
A.I Restore 7 days subsystem to
{
OPERABLE status.
t i
ABWR TS 3.7-I P&R, 7/30/93 j
RCW/RSW System
~
3.7.2 r
B.
Required Action and B.1 Be in MODE 3.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time of Condition A AND not met.
B.2 Be in. MODE 4.
36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.2.1 Verify the water level [in each RSW pump 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> well of the intake structure] is 2 [
]m
( ft) [mean sea level).
NOTE------------------------
Isolation of flow to individual components does not render RCW or RSW System inoperable.
P Verify each RCW and RSW subsystem manual, 31 days power operated, and automatic valve in the flow pt.th servicing safety related systems or components, that is not locked, sealed, or otherwise secured in position is in the correct position.
t SR 3.7.2.3 Verify each RCW/RSW subsystem actuates on 18 months an actual or simulated initiation signal.
i I
f ABWR TS 3.7-2 P&R, 7/30/93
CRHA HVAC EF System 3.7.3 3.7 PLANT SYSTEM
+
3.7.3 Control Room Habitability Area (CRHA) HVAC System - Emergency i
Filtration (EF) Subsystem LCO 3.7.3 Two EF trains of the CRHA HVAC System shall be OPERABLE.
1 APPLICABILITY:
MODES 1, 2, and 3, During movement of irradiated fuel assemblies in the primary or secondary containment, During CORE ALTERATIONS, During operations with a potential for draining the reactor vessel (0PDRVs).
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
One EF train A.1 Restore EF train to 7 days inoperable.
OPERABLE status.
B.
Required Action and B.1 Be in MODE 3.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Associated Completion Time of Condition A AND not met in MODE 1, 2, or 3.
B.2 Be in MODE 4.
36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> (continued) h I
1 i
t I
t ABWR TS 3.7-1 P&R, 7/30/93
CRHA HVAC EF System 3.7.3 r
ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME I
C.
Required Action and
= ------ N OT E --
- = - - -
associated Completion LC0 3.0.3 is not applicable.
Time of Condition A
-=--
=----
not met during movement of irradiated C.I Place OPERABLE EF train Immediately fuel assemblies in the in isolation mode.
primary or secondary containment, during OR CORE ALTERATIONS, or during OPDRVs.
C.2.1 Suspend movement of Immediately irradiated fuel assemblies in the primary and secondary containment.
AND C.2.2 Suspend CORE Immediately ALTERATIONS.
AND C.2.3 Initiate action to Immediately suspend OPDRVs.
l i
D.
Two EF trains D.I Enter LC0 3.0.3.
Immediately inoperable in MODE I, 2, or 3.
t t
(continued) l d
s 1
ABWR TS 3.7-2 P&R, 7/30/93
CRHA HVAC EF System 3.7.3 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME E.
Two EF trains
NOTE =---
-==-----
inoperable during LC0 3.0.3 is not applicable, movement of irradiated fuel assemblies in the primary or secondary E.1 Suspend movement of Immediately containment, during irradiated fuel CORE ALTERATIONS, or assemblies in the during OPDRVs.
primary and secondary j
containment.
AND E.2 Suspend CORE Immediately f
ALTERATIONS.
AND E.3 Initiate action to Immediately suspend OPDRVs.
l t
SURVEILLANCE REQUIREMENTS a
1 SURVEILLANCE FREQUENCY F
SR 3.7.3.1 Operate each EF train for 2 10 continuous 31 days hours with the heaters operating.
SR 3.7.3.2 Perform required EF filter testing in In accordance accordance with the Ventilation Filter with the VFTP Testing Program (VFTP).
(continued)
SR 3.7.3.3 Verify each EF train actuates on an actual 18 months or simulated initiation signal.
t ABWR TS 3.7-3 P&R, 7/30/93
,_.u__
CRHA HVAC EF System l
3.7.3 j
SURVEILLANCE REQUIREMENTS (continued) l SURVEILLANCE FREQUENCY SR 3.7.3.4 Verify each EF train can maintain 18 months
[
apositivepressureof2[3.17to on a 12.68 kg/m (.125 to.5 inches) water STAGGERED gauge relative to adjacent buildings TEST BASIS during the isolation mode p/h ([ ] cfm).
f operation t
at a flow rate of s [ ]m
==
6 4
i 1
i I
6 I
I 2
3 ABWR TS 3.7-4 P&R, 7/30/93 k
l Control-Room AC System 3.7.4 i
3.7 PLANT SYSTEM 3.7.4 Control Room Habitability Area (CRHA) - Control Room Air Conditioning i
(CRAC) Subsystem LC0 3.7.4 Two CRAC subsystems shall be OPERABLE.
APPLICABILITY:
MODES 1, 2, and 3, During movement of irradiated fuel assemblies in the primary or secondary containment, i
During CORE ALTERATIONS, During operations with a potential for draining the reactor vessel (0PDRVs).
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME l
A.
One CRAC subsystem A.1 Restore CRAC 30 days inoperable.
subsystem to OPERABLE status.
B.
Required Action and B.1 Be in MODE 3.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> I
Associated Completion Time of Condition A AND not met in MODE 1, 2, i
or 3.
B.2 Be in MODE 4.
36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> r
(continued) 1 i
i ABWR TS 3.7-1 P&R, 7/30/93 l
Control Room AC System 3.7.4 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME C.
Required Action and
---NOTE-------------
associated Completion LC0 3.0.3 is not applicable.
Time of Condition A
--- -- -- ---= -
not met during movement of irradiated C.1 Place OPERABLE CRAC Immediately fuel assemblies in the subsystem in operation.
I primary or secondary containment, during QR l
CORE ALTERATIONS, or i
during OPDRVs.
C.2.I Suspend movement of Immediately irradiated fuel assemblies in the primary and secondary containment.
AND 1
C.2.2 Suspend CORE Immediately ALTERATIONS.
1 A.!!D l
C.2.3 Initiate action to Immediately suspend OPDRVs.
I D.
Two CRAC subsystems D.I Enter LC0 3.0.3.
Immediately inoperable in MODE I, 2, or 3.
(continued)
ABWR TS 3.7-2 P&R, 7/30/93 l
l
i Control Room AC System 3.7.4
=
t ACTIONS (continued) i CONDITION REQUIRED ACTION COMPLETION TIME E.
Two CRAC subsystems
NOTE-inoperable during LC0 3.0.3 is not applicable.
movement of irradiated fuel assemblies in the primary or secondary E.1 Suspend movement of Immediately containment, during irradiated fuel CORE ALTERATIONS, or assemblies in the during OPDRVs.
primary and secondary containment.
AND E.2 Suspend CORE Immediately ALTERATIONS.
AND E.3 Initiate action to Immediately suspend OPDRVs.
i i
SURVEILLANCE REQUIREMENTS i
SURVEILLANCE FREQUENCY SR 3.7.4.1 Verify each CRAC subsystem has the 18 months capability to remove the assumed heat load.
f i
SR 3.7.4.2 Verify each CRAC subsystem actuates on an 18 months actual or simulated initiation signal.
l
(
i ABWR TS 3.7-3 P&R, 7/30/93 a
Main Condenser Offgas 3.7.5 3.7 PLANT SYSTEMS 3.7.5 Main Condenser Offgas i
LC0 3.7.5 The gross gamma activity rate of the noble gases measured at
[the offgas recombiner effluent] shall be s [380] mci /second
[after decay of 30 minutes].
t APPLICABILITY:
MODE 1, I
MODES 2 and 3 with any [ main steam line not isolated and]
l steam jet air ejector (SJAE) in operation.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME l
A.
Gross gamma activity A.1 Restore gross gamma 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />
[
rate of the noble activity rate of the gases not within noble gases to within i
limit.
limit.
B.
Required Action and B.1 Isolate all main 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion steam lines.
Time not met.
08 8.2 Isolate SJAE.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> OR B.3.1 Be in MODE 3.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> AND B.3.2 Be in MODE 4.
36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> b
ABWR TS 3.7-1 P&R, 7/30/93
Main Condenser Offgas 3.7.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.5.1
=----------NOTE--------------------
Not required to be performed until 31 days after any [ main steam line not isolated and] SJAE in operation.
Verify the gross gamma activity rate of the 31 days noble gases is s [380] mci /second [after decay of 30 minutes].
AND Once within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after a 2 50% increase in the nominal steady state fission gas release after factoring out increases due to changes in THERMAL POWER level i
ABWR TS 3.7-2 P&R, 7/30/93
i Main Turbine Bypass System 3.7.6 3.7 PLANT SYSTEMS 3.7.6 Main Turbine Bypass System i
LCO 3.7.6 The Main Turbine Bypass System shall be OPERABLE.
OB LCO 3.2.2, " MINIMUM CRITICAL POWER RATIO (MCPR)," limits for i
an inoperable Main Turbine Bypass System, as specified in the Core Operating Limits Report (COLR), are made applicable.
i J
APPLICABILITY:
THERMAL POWER 2 40% RTP.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
Requirements of the A.1 Satisfy the 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> LCO not met or Main requirements of the Turbine Bypass System LCO or restore Main inoperable.
Turbine Bypass System to OPERABLE status.
B.
Required Action and B.1 Reduce THERMAL POWER 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> associated Campletion to < 40% RTP.
Time not met.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.6.1 Verify one complete cycle of each main 31 days turbine bypass valve.
(continued)
ABWR TS 3.7-1 P&R, 7/30/93
Main Turbine Bypass System 3.7.6 SURVEILLANCE REQUIREMENTS (continued) i SURVEILLANCE FREQUENCY i
SR 3.7.6.2 Perform a system functional test.
18 months t
t i
SR 3.7.6.3 Verify the TURBINE BYPASS SYSTEM RESPONSE 18 months l
TIME is within limits.
I r
b o
?
I r
ABWR TS 3.7-2 P&R, 7/30/93 i
Fuel Pool Water Level 3.7.7 j
3.7 PLANT SYSTEMS 3.7.7 Fuel Pool Water Level LCO 3.7.7 The fuel pool water level shall be 2 7.01 m (23 ft) over the i
top of irradiated fuel assemblies seated in the spent fuel storage pool.
i APPLICABILITY:
During movement of irradiated fuel assemblies in the associated fuel storage pool.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
Fuel pool water level A.1
NOTE---------
not within limit.
LC0 3.0.3 is not applicable.
Suspend movement of Immediately irradiated fuel assemblies in the associated fuel storage pool (s).
r SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.7.1 Verify the fuel pool water level is 2 7 days 7.01 m (23 ft) over the top of irradiated fuel assemblies seated in the storage racks.
i i
ABWR TS 3.7-1 P&R, 7/30/93
i l
3.6.1.1 i
i 3.6 CONTAINMENT SYSTEMS
+
3.6.1.1 Primary Containment l
t LCO 3.6.1.1 Primary containment shall be OPERABLE.
i APPLICABILITY:
MODES 1,.2, and 3.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME f
A.
Primary containment A.1 Restore primary I hour inoperable.
containment to OPERABLE status.
i i
B.
Required Action and B.1 Be in MODE 3.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time not met.
AND r
B.2 Be in MODE 4.
36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> l
t i
h e
b i
6 ABWR TS 3.6-1 P&R, 8/13/93 i
Primary Containment 3.6.1.1 y
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY i
SR 3.6.1.1.1 Perform required visual examinations and
NOTE------
leakage rate testing except for primary SR 3.0.2 is not containment air lock testing, in applicable accordance with 10 CFR 50, Appendix J, l
as modified by approved exemptions.
In accordance The maximum allowable leakage rate, L,
with 10 CFR 50, is [0.5]% of primary containment air Appendix J, as weight per day at the calculated peak modified by containment pressure, P,.
approved exemptions 3.6.1.1.2 Verify primary containment structural In accordance
)
integrity in accordance with the Primary with the Containment Tendon Surveillance Program.
Primary Containment Tendon Surveillance Program I
SR 3.6.1.1.3 Verify drywell to suppression chamber 18 months differential pressure does not decrease at a rate > 6 mm (0.25 inch) water gauge AND per minute tested over a [10] minute periodataninitialpifferential
NOTE------
pressure of.07 kg/cm d (lpsid).
Only required after two consecutive tests fail and continues until j
two consecutive tests pass 9 months r
(
I t
ABWR TS 3.6-2 P&R, 8/13/93 f
1
Primary Containment Air Locks 3.6.1.2 3.6 CONTAINMENT SYSTEMS j
3.6.1.2 Primary Containment Air Locks I
LC0 3.6.1.2 Two primary containment air locks shall be OPERABLE.
i APPLICABILITY:
MODES 1, 2, and 3.
t ACTIONS
- =-----------
=
NOTES-----------
= - - - - - -
= - - -
1.
Entry and exit is permissible to perform repairs of the affected air lock i
components.
3 3
2.
Separate Condition entry is allowed for each air lock.
3.
Enter applicable Conditions and Required Actions of LC0 3.6.1.1, " Primary Containment," when air lock leakage results in exceeding overall containment leakage rate acceptance criteria.
_ =- __------------------------------------------------ _ _
CONDITION REQUIRED ACTION COMPLETION TIME I
A.
One or more primary
NOTES
=
t containment air locks 1.
Required Actions A.1, with one primary A.2, and A.3 are not containment air lock applicable if both doors door inoperable.
in the same air lock are inoperable and Condition C is entered.
2.
Entry and exit is permissible for 7 days under administrative controls if both air locks are inoperable.
(continued) i P
ABWR TS 3.6-1 P&R, 8/13/93 l
t e
Primary Containment Air Locks 3.6.1.2 i
ACTIONS i
CONDITION REQUIRED ACTION COMPLETION TIME A.
(continued)
A.1 Verify the OPERABLE 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> door is closed in the affected air lock.
AND A.2 Lock the OPERABLE 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> door closed in the t
affected air lock.
AND A.3
NOTE---------
Air lock doors in high radiation areas or areas with limited access due to inerting may be verified locked closed by administrative means.
Once per 31 days Verify the OPERABLE door is locked closed in the affected air lock.
B.
One or more primary
NOTES------------
containment air locks 1.
Required Actions B.1, with primary B.2, and B.3 are not containment air lock applicable if both doors interlock mechanism in the same air lock are inoperable.
inoperable and Condition C is entered.
2.
Entry into and exit from containment is permissible under the control of a dedicated individual.
(continued)
ABWR TS 3.6-2 P&R, 8/13/93 l
~
i
'I Primary Containment Air Locks 3.6.1.2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B.
(continued)
B.I Verify an OPERABLE-I hour door is closed in the affected air lock.
AND B.2 Lock an OPERABLE door 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> closed in the i
affected uair lock.
AND l
B.3
NOTE---------
Air lock doors in high radiation areas or areas with limited access due to inerting may be i
verified locked closed by administrative means.
Verify an OPERABLE Once per 31 days door is locked closed in the affected air lock.
f p
?
ABk'R TS 3.6-3 P&R, 8/13/93
Primary Containment Air Locks I
3.6.1.2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C.
One or more primary C.1 Initiate action to Immediately containment air locks evaluate primary inoperable for reasons containment overall other than Condition A leakage rate per i
or B.
LC0 3.6.1.1, using current air lock test results.
bi Q l
C.2 Verify a door is I hour closed in the affected air lock.
AND (continued)
C.
(continued)
C.3 Restore air lock to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> OPERABLE status.
D.
Required Action and D.1 Be in MODE 3.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time not met.
AND D.2 Be in MODE 4.
36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> ABk'R TS 3.6-4 P&R, 8/13/93 l
i Primary Containment Air Locks 3.6.1.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY l
NOTES------------------
t 1.
An inoperable air lock door does not invalidate the previous successful performance of the overall air lock leakage test.
t 2.
Results shall be evaluated against acceptance criteria of SR 3.6.1.1.1 in accordance with 10 CFR 50, Appendix J, as modified by approved exemptions.
s l
Perform required primary containment air
-- NOT E---
lock leakage rate testing in accordance SR 3.0.2 is not with 10 CFR 50, Appendix J, as modified applicable by approved exemptions.
The acceptance criteria for air lock In accordance testing are:
with 10 CFR 50, Appendix J, as a.
Overall air lock leakage rate is modified by s 0.05 L, when tested at 2 P,.
approved exemptions b.
For each door, leakage rate is s 0.01 L when the gap between the door sea s is pressurized to a
2.7 kg/cm g (10 psig) for at least 15 minutes.
~~~
SR 3.6.1.2.2 Verify primary containment air lock seal 7 days air flask pressure is 2 [90] psig.
NOTE----
- =-
= = = - - - - -
Only required to be performed upon entry into primary containment when the primary containment is de-inerted.
=-
Verify only one door in the primary 184 days containment air lock can be opened at a time.
ABWR TS 3.6-5 P&R, 8/13/93
PCIVs i
3.6.1.3 3.6 CONTAINMENT SYSTEMS 3.6.1.3 Primary Containment Isolation Valves (PCIVs) i LC0 3.6.1.3 Each PCIV shall be OPERABLE.
APPLICABILITY:
MODES 1, 2, and 3, When associated instrun,.tation is required to be OPERABLE per LC0 3.3.6.1, " Primary Containment Isolation Instrumentation."
ACTIONS t
NOTES------------------------------------
-l 1.
Penetration flow paths [except for purge valve penetration flow paths] may be unisolated intermittently under administrative controls.
l 2.
Separate Condition entry is allowed for each penetration flow path.
3.
Enter applicable Conditions and Required Actions for systems made inoperable by PCIVs.
4.
Enter applicable Conditions and Required' Actions of LCO 3.6.1.1, " Primary l
Containment," when PCIV leakage results in exceeding overall containment leakage rate acceptance criteria.
= _=----------------------------------------------------------
f CONDITION REQUIRED ACTION COMPLETION TIME A.
NOTE---------
A.1 Isolate the affected 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> except Only applicable to penetration flow path for main steam penetration flow paths by use of at least line with two PCIVs.
one closed and de-activated AND automatic valve, One or more closed manual valve, 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> for main penetration flow paths blind flange, or steam line with one PCIV check valve with flow inoperable (except for through the valve i
purge valve or secured.
secondary containment bypass leakage not AND within limit].
(continued)
I ABWR TS 3.6-1 P%R, 8/13/93
[
t
PCIVs l
3.6.I.3 l
~
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
(continued)
A.2
NOTE---------
Valves and blind flanges in high radiation areas may be verified by use of administrative means.
Verify the affected Once per 31 days penetration flow path for isolation is isolated.
devices outside primary containment, drywell, and steam tunnel AND Prior to entering MODE 2 or 3 from MODE 4, if primary containment was de-inerted while in MODE 4, if not performed within the previous 92 days, for isolation devices inside i
primary containment (continued) i ABWR TS 3.6-2 P%R, 8/13/93
i PCIVs 3.6.1.3 ACTIONS (continued)
CONDITI0N REQUIRED ACTION COMPLETION TIME B.
NOTE---------
8.1 Isolate the affected I hour f
Only applicable to penetration flow path penetration flow paths by use of at least with two PCIVs.
one closed and de-activated automatic valve, One or more closed manual valve, penetration flow paths or blind flange.
with two PCIVs inoperable [except for purge valve leakage not within limit].
i C.
NOTE---------
C.1 Isolate the affected 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> except Only applicable to penetration flow path for excess flow penetration flow paths by use of at least check valves with only one PCIV.
one closed and (EFCVs) de-activated automatic valve, AND One or more closed manual valve, penetration flow paths or blind flange.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for with one PCIV EFCVs i
AND C.2
NOTE---------
Valves and blind flanges in high radiation areas may be verified by use of administrative means.
Verify the affected Once per 31 days penetration flow path is isolated.
D.
Secondary containment D.1 Restore leakage rate 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> I
bypass leakage rate to within limit.
not within limit.
l l
(continued)
ABWR TS 3.6-3 P%R, 8/13/93
I PCIVs 3.6.1.3 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME i
E. One or more E.1 Isolate the affected 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> i
penetration flow paths penetration flow path l
with one or more by use of at least containment purge one [ closed and valves not within de-activated i
purge valve leakage automatic valve, limits.
closed manual valve, 1
or blind flange].
l t
AND E.2
NOTE---------
i Valves and blind flanges in high radiation areas may l
be verified by use of administrative means.
5 Verify the affected Once per 31 days penetration flow path for isolation is isolated, devices outside containment AND Prior to entering MODE 4 if not performed within the previous 92 days for isolation i
devices inside containment AND E.3 Perform SR 3.6.1.3.7 Once per for the resilient
[92] days seal purge valves closed to comply with Required Action E.1.
(continued) 1 ABWR TS 3.6-4 P%R, 8/13/93-
PCIVs 3.6.1.3 i
ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME i
F.
Required Action and F.1 Be in MODE 3.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time of Condition A, AND B, C, D, or E not met in MODE 1, 2, or 3.
F.2 Be in MODE 4.
36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> G.
Required Action and G.I
NOTE---------
associated Completion LCO 3.0.3 is not Time of Condition A, applicable.
B, C, D, or E not met r
for PCIV(s) required r
to be OPERABLE during Suspend movement of Immediately movement of irradiated irradiated fuel fuel assemblies in assemblies in primary i
the primary or and secondary secondary containment.
containment.
h I
H.
Required Action and H.1 Suspend CORE Immediately associated Completion ALTERATIONS.
Time of Condition A, B, C, D, or E not met for PCIV(s) required i
to be OPERABLE during 4
CORE ALTERATIONS, j
I.
Required Action and I.I Initiate action to Immediately associated Completion suspend OPDRVs.
Time of Condition A, B, C, D, or E not met OR for PCIV(s) required to be OPERABLE during I.2 Initiate action to Immediately MODE 4 or 5 or during restore valve (s) to operations with a OPERABLE status.
potential for draining the reactor vessel (OPDRVs).
t ABWR TS 3.6-5 P%R, 8/I3/93 r
p
+
_..e.
.. = _.,
I PCIVs 3.6.1.3
)
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY
)
NOTE------------------
I, Only required to be met in MODES 1, 2, and 3.
- - - - - - - - - - - - - - - - - - - - - - - - - _ _ _ _ _ _ = - - - - -
t Verify each 550 mm (22 inch) primary 31 days containment purge valve is sealed closed except for one purge valve in a penetration flow path while in
-t Condition E of this LCO.
NOTES------------------
t 1.
Only required to be met in MODES 1, 2, and 3.
2.
Not required to be met when the 550 mm (22 inch) primary containment purge valves are open for t
pressure control, ALARA or air quality considerations for personnel j
entry, or Surveillances that require the valves to be open.
l i
Verify each 550 mm (22 inch) primary 31 days containment purge valve is closed.
(continued) i f
9 ABWR TS 3.6-6 P%R, 8/13/93 j
t r
PCIVs 3.6.1.3 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.6.1.3.3
NOTES-1.
Valves and blind flanges in high radiation areas may be verified by use of administrative means.
)
2.
Not required to be met for PCIVs that are open under administrative controls.
Verify each primary containment isolation 31 days manual valve and blind flange that is located outside primary containment and is required to be closed during accident conditions is closed.
NOTES------------------
l.
Valves and blind flanges in high radiation areas may be verified by use of administrative means.
l 2.
Not required to be met for PCIVs that l
are open under administrative controls.
Verify each primary containment isolation Prior to manual valve and blind flange that is entering MODE 2 located inside primary containment and is or 3 from required to be closed during accident MODE 4, if conditions is closed.
primary containment was de-inerted while in MODE 4, if not performed within the previous 92 days 4
ABWR TS 3.6-7 P%R, 8/13/93
i i
PCIVs 3.6.1.3 SURVEILLANCE REQUIREMENTS (continued) i SURVEILLANCE FREQUENCY SR 3.6.1.3.5 Verify continuity of the traversing incore 31 days i
prob (TIP) shear isolation valve explosive l
charge.
j i
SR 3.6.1.3.6 Verify the isolation time of each power In l
operated and each automatic PCIV, except accordance MSIVs, is within limits.
~
with the Inservice Testing Program or 92 days SR 3.6.1.3.7
NOTES------------------
l 1.
Only required to be met in MODES 1, 2, and 3.
2.
Results shall be evaluated against l
acceptance criteria of SR 3.6.1.1.1 in accordance with 10 CFR 50, Appendix J, as modified by approved exemptions.
Perform leakage rate testing for each 184 days primary containment purge valve with resilient seals.
AND Once within 92 days after opening the valve r
i ABWR TS 3.6-8 P%R, 8/13/93 m
i
PCIVs 3.6.1.3 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.6.1.3.8 Verify the isolation time of each MSIV is In accordance 2 3 seconds and s 4.5 seconds.
with the Inservice Testing Program or
?
18 months f
i SR 3.6.1.3.9 Verify each automatic PCIV actuates to 18 months the isolation position on an actual or simulated isolation signal.
SR 3.6.1.3.10 Verify each reactor instrumentation line 18 months EFCV actuates on a simulated instrument line break to restrict flow to 5 3.8 L/h (lgph).
i b
h p
I ABWR TS 3.6-9 P%R, 8/13/93
PCIVs 3.6.1.3 k
SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.6.1.3.11 Remove and test the explosive squib from 18 months on each shear isolation valve of the TIP a STAGGERED System.
TEST BASIS SR 3.6.1.3.12
NOTE-------------------
1.
Only required to be met in MODES 1, 2, and 3.
2.
Results shall be evaluated against
-- acceptance criteria of SR 3.6.1.1.1 in accordance with 10 CFR 50, Appendix J, as modified by approved exemptions.
Verify the combined leakage rate of 18 months l
[3.81/m (1 gpm) times the total number of S
PCIVs] through hydrostatically tested lines that penetrate the primary containment is not exceeded when these isolation valves are tested at 2[ ] kg/cm g ([ ] psig).
2 o
6 ABWR TS 3.6-10 P%R, 8/13/93
PCIVs 3.6.1.3 I
SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.6.1.3.13
NOTE--------
NOTE --
Results shall be evaluated against SR 3.0.2 acceptance criteria of SR 3.6.1.1.1 in is not accordance with 10 CFR 50, Appendix J, as applicable modified by approved exemptions.
=
Verify /h (35 scfh) when tested at 21.76 leakage rate through each M In accordance s1m with 2
kg/cm g (25 psig).
10 CFR 50, Appendix J, as modified by approved xemptions i
9 h
5 I
l ABWR TS 3.6-11 P%R, 8/13/93 l
PCIVs 3.6.1.3 i
SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.6.1.3.14
NOTES------------
.i 1.
Only required to be met in MODES 1, 2, and 3.
i r
Verify each [550 mm (22 inch)] primary 18 months containment purge valve is blocked to i
restrict the valve from opening > [50]%.
i i
}
SR 3.6.1.3.15
NOTE-------------------
Results shall be evaluated against i
acceptance criteria of SR 3.6.1.1.1 in i
accordance with 10 CFR 50, Appendix J, as i
modified by approved exemptions.
i Verify the combined leakage rate for all 18 months
(
secondary containment bypass leakage paths is s [a L,] when pressurized to 2[
] kg/cm g ([
] psig).
5 1
i i
i l
i i
t i
ABWR TS 3.6-12 P%R, 8/13/93 j
Drywell Pressure 3.6.1.4 i
3.6 CONTAINMENT SYSTEMS 3.6.1.4 Drywell Pressure t
LC0 3.6.1.4 Drywell pressure shall be s.0527 kg/cm g (0.75 psig).
APPLICABILITY:
MODES 1, 2, and 3.
ACTIONS CONDITION PEQUIRED ACTION COMPLETION TIME A.
Drywell pressure not A.1 Restore drywell I hour within limit.
pressure to within limit.
i B.
Required Action and B.1 Be in MODE 3.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time not met.
AND B.2 Be in MODE 4.
36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> e
i SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.1.4.1 Verify drywell pressure is within limit.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 4
ABWR TS 3.6-1 P&R, 8/13/93
i Drywell Air Temperature l
3.6.1.5 i
3.6 CONTAINMENT SYSTEMS i
l 3.6.1.5 Drywell Air Temperature l
I 8
LC0 3.6.1.5 Drywell average air temperature shall be s 57 C (135'F).
?
APPLICABILITY:
MODES 1, 2, and 3.
ACTIONS i
CONDITION REQUIRED ACTION COMPLETION TIME i
t i
A.
Drywell average air-A.1 Restore drywell 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> l
temperature not within average air limit.
temperature to within limit.
B.
Required Action and B.1 Be in MODE 3.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> I
associated Completion Time not met.
AND f
B.2 Be in MODE 4.
36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> i
}
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.1.5.1 Verify drywell average air temperature is 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> within limit.
I l
)
ABWR TS 3.6-1 P&R, 8/13/93
Suppression-Chamber-to-Drywell Vacuum Breakers 3.6.1.6 6
3.6 CONTAINMENT SYSTEMS 3.6.1.6 Suppression-Chamber-to-Drywell Vacuum Breakers LCO 3.6.1.6 Eight suppression-chamber-to-drywell vacuum breakers shall be OPERABLE.
AND Eight suppression-chamber-to-drywell vacuum breakers shall be closed.
APPLICABILITY:
MODES 1, 2, and 3.
l ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
One suppression-A.1 Restore one vacuum 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> chamber-to-drywell breaker to OPERABLE vacuum breaker status.
inoperable for opening.
B.
One suppression-B.1 Close the open vacuum 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> chamber-to-drywell breaker.
vacuum breaker not closed.
C.
Required Action and C.1 Be in MODE 3.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time not met.
AND 0.2 Be in MODE 4.
36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> ABWR TS 3.6-1 P&R, 8/13/93
i Suppre:,sion-Chamber-to-Drywell Vacuum Breakers l
3.6.1.6 i
SURVEILLANCE P,EQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.1.6.1
==-NOTE
--==
Not required to be met for vacuum i
breakers that are open during -
Surveillances or when performing their intended function.
Verify each vacuum breaker is closed.
14 days AND Within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after any discharge of steam to the suppression chamber from the safety /
t relief valves (S/RVs) or any operation that causes the drywell-suppression chamber differential pressure to be reduced by 1.0352kg/cm d (0.5 psid).*
l (continued) i
- If position indicating instruments indicate that one or more vacuum breakers are not closed, verify by alternate means that each vacuum breaker is closed within the following 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />.
SR 3.6.1.6.2 Perform a functional test of each 18 months required vacuum breaker.
ABWR TS 3.6-2 P&R, 8/13/93
l 9
Suppression-Chamber-to-Drywell Vacuum Breckers 3.6.1.6 i
SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.6.1.6.3-Verify the opening setpoint of each (18] months l
requipedvacuumbreakeriss.0352 i
kg/cm d (0.5 psid).
i r
[
I
)
i i
1 l
ABWR TS 3.6-3 P&R, 8/13/93 i
Suppression Pool Average Temperature 3.6.2.1 4
3.6 CONTAINMENT SYSTEMS 3.6.2.1 Suppression Pool Average Temperature LCO 3.6.2.1 Suppression pool average temperature shall be:
I a.
s 43 C (110*F) when THERMAL POWER is > 1% RTP ano no testing that adds heat to the suppression pool is being performed; b.
s 46 C (115'F) when THERMAL POWER is > 1% RTP and testing that adds heat to the suppression pool is being performed; and c.
s 49 C (120*F) when THERMAL POWER is s 1% RTP.
APPLICABILITY:
MODES 1, 2, and 3.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
Suppression pool A.1 Verify suppression Once per hour average temperature pool average
> 43"C {l10*F) but temperature is s 49"C s 49"C (120*F).
(120*F).
AND AND THERMAL POWER is > 1%
A.2 Restore suppression' 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> RTP.
pool average temperature to s 43 C AND (110*F).
Not performing testing that adds heat to the suppression pool.
[
5 (continued) l ABWR TS 3.6-1 P&R, 8/13/93 l
i 1
Suppression Pool Average Temperature 3.6.2.1 1
ACTIONS (contir.ued)
CONDITION REQUIRED ACTION COMPLETION TIME
=
B.
Required Action and B.1 Reduce THERMAL POWER 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion to < 1% RTP.
Time of Condition A not met.
C.
Suppression pool C.1 Suspend all testing Immediately average temperature that adds heat to the
> 46"C (ll5'F).
suppression pool.
AND THERMAL POWER > 1%
RTP.
AND Performing testing
[
that adds heat to the suppression pool.
D. ' Suppression pool D.1 Verify suppression Once per averggetemperature pool average 30 minutes 0
> 49 C (120*F) but temperature is s 54 C s 54"C (130*F).
(130*F).
l (continued)
E.
Suppression pool E.1 Depressurize the 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> average temperature reactorvessplto t
0
> 54 0 (130*F).
< 14.1 kg/cm g (200 psig).
t AND 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> E.2 Be in H0DE 4.
t i
i ABWR TS 3.6-2 P&R, 8/13/93 i
1 i
Suppression Pool Average Temperature 3.6.2.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY l
SR 3.6.2.1.1 Verify suppression pool average 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> temperature is within the applicable limits.
AND P
5 minutes when l
performing testing that adds heat to the suppression l
pool i
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ABWR TS 3.6-3 P&R, 8/13/93 f
Suppression Pool Water level 3.6.2.2 3.6 CONTAINMENT SYSTEMS 3.6.2.2 Suppression Pool Water Level LCO 3.6.2.2 Suppression pool water level shall be 2 7 meters and s 7.1 meters.
APPLICABILITY:
MODES 1, 2, and 3.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
Suppression pool water A.1 Restore suppression 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> level not within pool water level to
- limits, within limits.
e B.
Required Action and 8.1 Be in MODE 3.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time not met.
AND B.2 Be in MODE 4.
36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> I
SURVEILLANCE REQUIREMENTS i
SURVEILLANCE FREQUENCY l
I SR 3.6.2.2.1 Verify suppression pool water level is 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> within limits.
S i
s ABWR TS 3.6-1 P&R, 8/13/93 Y
RHR Suppression Pool Cooling -
3.6.2.3
+
3.6 CONTAINMENT SYSTEMS 3.6.2.3 Residual Heat Removal (RHR) Suppression Pool Cooling LC0 3.6.2.3 Three RHR suppression pool cooling subsystems shall be OPERABLE.
l l
APPLICABILITY:
MODES 1, 2, and 3.
I ACTIONS i
i CONDITION REQUIRED ACTION COMPLETION TIME t
A.
One RHR suppression A.1 Restore RHR 7 days pool cooling subsystem suppression pool inoperable.
cooling subsystem to OPERABLE status.
{
B. Two RHR suppression B.1 Restore one RHR 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> pool cooling subsystems suppression pool inoperable.
cooling subsystem r
to operable status.
f C.
Required Action and C.1 Be in MODE 3.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion i
Time of Condition A AND i
not met.
C.2 Be in MODE 4.
36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> l
08 j'
Three RHR suppression pool cooling subsystems inoperable.
l I
ABWR TS 3.6-1 P&R, 08/13/93 0
RHR Suppression Pool Cooling 3.6.2.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.2.3.1 Verify each RHR suppression pool cooling 31 days subsystem manual, power operated, and i
automatic valve in the flow path that is not locked, sealed, or otherwise secured in position is in the correct position or can be aligned to the correct position.
t l
SR 3.6.2.3.2 Verify each RHR pump develops a flow rate In accordance
> 265 1/s (4200 gpm) through the with the associated heat exchanger while operating Inservice in the suppression pool cooling mode.
Testing Program or 92 days L
t i
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l
+
t f
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1 ABWR TS 3.6-2 P&R, 08/13/93 I
r
i RHR Wetwell Spray 3.6.2.4 l
3.6 CONTAINMENT SYSTEMS l
3.6.2.4 Residual Heat Removal (RHR) Wetwell Spray 7
i LCO 3.6.2.4 Two RHR wetwell spray subsystems shall be OPERABLE.
APPLICABILITY:
MODES 1, 2, and 3.
ACTIONS i
CONDITION REQUIRED ACTION COMPLETION TIME 1
A.
One RHR wetwell spray A.1 Restore RHR wetwell 7 days subsystem inoperable.
spray subsystem to OPERABLE status.
i B.
Two RHR wetwell spray B.1 Restore one RHR 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> subsystems inoperable.
wetwell spray subsystem to OPERABLE status.
l C.
Required Action and C.1 Be in MODE 3.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Coupletion Time not met.
AND C.2 Be in MODE 4.
36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> l
l j
r i
L l
I i
1 ABWR TS-3.6-1 P&R, 8/13/93 i
RHR Wetwell Spray 3.6.2.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.2.4.1 Verify each RHR wetwell spray subsystem 31 days manual, power operated, and automatic valve in the flow path that is not i
locked, sealed, or otherwise secured in position is in the correct position or can be aligned to the correct position.
SR 3.6.2.4.2 Verify each associate (i.e., in In subsystems B & C) RHR pump develops a accordance flow rate 2 31.51/s (500 gpm) through with the the heat exchanger while operating in the Inservice wetwell spray mode.
Testing Program or 92 days i
I f
f i
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ABWR TS 3.6-2 P&R, 8/13/93
Primary Containment Hydrogen Recombiners-3.6.3.1 3.6 CONTAINMENT SYSTEMS 3.6.3.1 Primary Containment Hydrogen Recombiners LCO 3.6.3.1 Two primary containment hydrogen recombiners shall be OPERABLE APPLICABILITY:
MODES I and 2.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
One primary A.1
NOTE---------
containment hydrogen LC0 3.0.4 is not recombiner inoperable.
applicable.
Restore primary 30 days containment hydrogen recombiner to OPERABLE status.
B.
Two primary B.1 Verify by I hour i
containment hydrogen administrative means recombiners that the hydrogen AND inoperable.
control function is maintained.
Every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter AND B.2 Restore one primary 7 days containment hydrogen recombiner to f
OPERABLE status.
C.
Required Action and C.1 Be in MODE 3.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time not met.
ABWR TS 3.6-1 P&R, 8/13/93
i
~
Primary Containment Hydrogen Recombiners 3.6.3.1 SURVEILLANCE REQUIREMENTS I
SURVEILLANCE FREQUENCY SR 3.6.3.1.1 Perform a system functional test for each 18 months primary containment hydrogen recombiner.
SR 3.6.3.1.2 Visually examine each primary containment 18 months hydrogen recombiner enclosure and verify there is no evidence of abnormal conditions.
i SR 3.6.3.1.3 Perform a resistance to ground test for 18 months each heater phase.
i t
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ABWR TS 3.6-2 P&R, 8/13/93 i
i i
I
Primary Containment Oxygen Concentration 3.6.3.2 3.6 CONTAINMENT SYSTEMS f
3.6.3.2 Primary Containment Oxygen Concentration i
t LC0 3.6.3.2 The primary containment oxygen concentration shall be
< 3.5 volume percent.
I APPLICABILITY:
MODE 1 during the time period:
j From 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after THERMAL POWER is > 15% RTP following a.
startup, to b.
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to reducing THERMAL POWER to < 15% RTP prior to the next scheduled reactor shutdown.
1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
Primary containment A.1 Restore oxygen 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> oxygen concentration concentration to not within limit.
within limit.
f f
B.
Required Action and B.1 Reduce THERMAL POWER 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> associated Completion to s 15% RTP.
t Time not met.
i i
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY r
SR 3.6.3.2.1 Verify primary containment oxygen 7 days concentration is within limits.
i BWR/4 STS 3.6-1 08/10/93 9:38am
t Secondary Containment 3.6.4.1 t
3.6 CONTAINMENT SYSTEMS 3.6.4.1 Secondary Containment 1
LCO 3.6.4.1 The secondary containment shall be OPERABLE.
APPLICABILITY:
MODES 1, 2, and 3, During movement of irradiated fuel assemblies in the secondary containment, During CORE ALTERATIONS, During operations with a potential for draining the reactor vessel (OPDRVs).
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
Secondary containment A.1 Restore secondary 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> inoperable in MODE 1, containment to 2, or 3.
OPERABLE status.
B.
Required Action and B.1 Be in MODE 3.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time of Condition A AND l
not met.
B.2 Be in MODE 4.
36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> i
C.
NOTE---------
inoperable during LC0 3.0.3 is not movement of irradiated applicable.
i fuel assemblies in the
=----
during CORE Suspend movement of Immediately ALTERATIONS, or during irradiated fuel OPDRVs.
assemblies in the secondary containment.
i AND (continued)
ABWR TS 3.6-1 P&R, 8/13/93
i Secondary Containment 3.6.4.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C.
(continued)
C.2 Suspend CORE Immediately ALTERATIONS.
AND C.3 Initiate action to Immediately suspend OPDRVs.
P SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.4.1.1 Verify secondary containment vacuum is 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 2
2 6.3426E-04 Kg/cm g (0.25 inch of vacuum water gauge).
SR 3.6.4.1.2 Verify all secondary containment 31 days equipment hatches are closed and sealed.
i SR 3.6.4.1.3 Verify each secondary containment access 31 days door is closed, except when the access opening is being used for entry and exit, then at least one door shall be closed.
SR 3.6.4.1.4 Verify each standby gas treatment 18 months on (SGT) subsystem will draw down the a STAGGERED secondary containment to TEST BASIS 2
2 6.3426E-04 Kg/cm g (0.25 inch of vacuum water gauge) in s 120 seconds.
6 (continued) l l
l ABWR TS 3.6-2 P&R, 8/13/93 L
l
J Secondary Containment 3.6.4.1 SURVEILLANCE REQUIREMENTS (continued)
SURVEll. LANCE FREQUENCY SR 3.6.4.1.5 Verify each SGT supsystem can maintain 18 months on a 16.3426E-04 Kg/cm g (0.25 inch of vacuum STAGGERED TEST water gauge) in the secondary BASIS containment for I hour at a flow rate s (4000 cfm).
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I ABWR TS 3.6-3 P&R, 8/13/93 e
SCIVs 3.6.4.2 3.6 CONTAINMENT SYSTEMS i
3.6.4.2 Secondary Containment Isolation Valves (SCIVs)
LCO 3.6.4.2 Each SCIV shall be OPERABLE.
i' APPLICABILITY:
MODES I, 2, and 3, During movement of irradiated fuel assemblies in the
+
secondary containment, During CORE ALTERATIONS, During operations with a potential for draining the reactor r
vessel (OPDRVs).
i ACTIONS i
NOTES =
-=
i I.
Penetration flow paths may be unisolated intermittently under administrative controls.
2.
Separate Condition entry is allowed for each penetration flow path.
3.
Enter applicable Conditions and Required Actions for systems made inoperable by SCIVs.
__=- - -------------------------------
CONDITION REQUIRED ACTION COMPLETION TIME A.
One or more A.I Isolate the affected 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> penetration flow paths penetration flow path with one SCIV by use of at least inoperable.
one closed and de-activated automatic valve, closed manual valve, or blind flange.
I AND l
t (continued) f t
ABWR TS 3.6-1 P&R, 8/13/93
SCIVs 3.6.4.2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
(continued)
A.2
-NOTE---------
Valves and blind flanges in high radiation areas may be verified by use of administrative means.
Verify the affected Once per 31 days penetration flow path is isolated.
B.
NOTE---------
B.1 Isolate the affected 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> i
Only applicable to penetration flow path penetration flow paths by use of at least with two isolation one closed and valves.
de-activated automatic valve, closed manual valve, One or more or blind flange.
penetration flow paths with two SCIVs inoperable.
C.
Required Action and C.1 Be in MODE 3.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time of Condition A AND or B not met in MODE 1, 2, or 3.
C.2 Be in MODE 4.
36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> (continued) t ABWR TS 3.6-2 P&R,8/13/93
SCIVs 3.6.4.2 ACTIONS (continued) e CONDITION REQUIRED ACTION COMPLETION TIME i
D.
Required Action and D.I
= NOTE---------
associated Completion LC0 3.0.3 is not Time of Condition A applicable.
or B not met during
=- -
===---
movement of irradiated I
fuel assemblies in the Suspend movement of Immediately secondary containment, irradiated fuel i
during CORE assemblies in the ALTERATIONS, or during secondary OPDRVs.
containment.
AND J
D.2 Suspend CORE Immediately ALTERATIONS.
l AND D.3 Initiate action to Immediately suspend OPDRVs.
l i
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a e
ABWR TS 3.6-3 P&R, 8/I3/93 1
SCIVs 3.6.4.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.4.2.1
NOTES--=
1.
Valves and blind flanges in high radiation areas may be verified by i
use of administrative means.
2.
Not required to be met for SCIVs that i
are open under administrative controls.
_=_ =--_
a L
Verify each secondary containment 31 days isolation manual valve and blind flange that is required to be closed during accident conditions is closed.
SR 3.6.4.2.2 Verify the isolation time of each power In accordance operated and each automatic SCIV is with the t
within limits.
Inservice Testing Program or 92 days SR 3.6.4.2.3 Verify each automatic SCIV actuates to 18 months the isolation position on an actual or j
simulated actuation signal.
1 E
i l
r ABWR TS 3.6-4 P&R, 8/13/93 f
1 l
SGT System 3.6.4.3 3.6 CONTAINMENT SYSTEMS 3.6.4.3 Standby Gas Treatment (SGT) System i
LCO 3.6.4.3 Two SGT trains shall be OPERABLE.
APPLICABILITY:
MODES 1, 2, and 3, During movement of irradiated fuel assemblies in the i
secondary containment, During CORE ALTERATIONS, During operations with a potential for draining the reactor vessel (0PDRVs).
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
One SGT train A.1 Restore SGT train to 7 days inoperable.
OPERABLE status.
B.
Required Action and B.1 Be in MODE 3.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time of Condition A AND not met in MODE 1, 2, or 3.
B.2 Be in MODE 4.
36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> C.
Required Action and
NOTE-------------
associated Completion LCO 3.0.3 is not applicable.
Time of Condition A not met during movement of irradiated C.1 Place OPERABLE SGT Immediately fuel assemblies in the train in operation, secondary containment, during CORE OR ALTERATIONS, or during OPDRVs.
(continued)
I' ABWR TS 3.6-1 P&R, 8/13/93 Y
P m
SGT System 3.6.4.3 l
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C.
(continued)
C.2.1 Suspend movement of Immediately irradiated fuel assemblies in secondary containment.
I AND C.2.2 Suspend CORE Immediately ALTERATIONS.
AND C.2.3 Initiate action to Immediately suspend OPDRVs.
D.
Two SGT subsystems D.I
--- - - N OT E---------
inoperable during LC0 3.0.3 is not movement of irradiated applicable.
fuel assemblies in the
- -- =-~------
secondary containment, during CORE Suspend movement of Immediately ALTERATIONS, or during irradiated fuel
~,
assemblies in secondary containment.
AND D.2 Suspend CORE Immediately ALTERATIONS.
AND D.3 Initiate action to Immediately suspend OPDRVs.
I k
ABWR TS 3.6-2 P&R, 8/I3/93 1
SGT System 3.6.4.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY
}
SR 3.6.4.3.1 Operate each SGT train for 2 [10]
31 days continuous hours with heaters operating.
SR 3.6.4.3.2 Perform required SGT filter testing in In accordance accordance with the Ventilation Filter with the VFTP Testing Program (VFTP).
t SR 3.6.4.3.3 Verify each SGT train actuates on an 18 months actual or simulated initiation signal.
l SR 3.6.4.3.4 Verify each SGT filter cooler bypass 18 months damper can be opened and the fan started.
{
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ABWR TS 3.6-3 P&R, 8/13/93
UHS B 3.7.1
}
i B 3.7 PLANT SYSTEMS i
B 3.7.1 Ultimate Heat Sink (VHS) i BASES BACKGROUND The UHS is designed to provide sufficient cooling water to the Reactor Service Water (RSW) System to permit safe i
shutdown and cooldown of the unit and to maintain the unit in a safe shutdown condition and, in the event of an accider.'., to provide sufficient cooling water to the RSW System to safely dissipate the heat for that accident. The RSW System is described in the Bases for LC0 3.7.2, " Reactor Cooling Water (RCW) System and Reactor Service Water (RSW)
System."
[This section will describe the UHS design which is site specific.]
l I
APPLICABLE The volume of each water source incorporated in a UHS SAFETY ANALYSES complex is sized so that sufficient water inventory is available for all [RSW) System post LOCA cooling requirements for a 30 day period with no additional makeup water source available (Ref. I).
1 The UHS, satisfies Criterion 3 of the NRC Policy Statement.
i S
LCO OPERABILITY of the UHS is based on a maximum water temperature of 35*C (95'F).
[0ther operability requirements for the UHS are site specific.]
i APPLICABILITY In MODES 1, 2, and 3, the UHS is required to be OPERABLE to i
support OPERABILITY of the equipment serviced by the VHS, and is required to be OPERABLE in these MODES.
In MODES 4 and 5, the OPERABILITY requirements of the UHS is determined by the systems they support.
l (continued)
ABWR TS B 3.7-I P&R, 08/02/93 I:26pm w
}
UHS B 3.7.1 r
t BASES i
ACTIONS
.A.L1 l
If one or more [ UHS active components] are inoperable, action must be taken to restore the inoperable [ components]
to OPERABLE status within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Completion Time is reasonable, based on the low i
probability of an accident occurring during the 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 4
that a [ UHS active component] is inoperable, the number of l
available systems, and the time required to complete the Required Action.
i 2
B.1 and B.2 If the [ UHS active component (s)] cannot be restored to i
OPERABLE status within the associated Completion Time, or the UHS is determined inoperable for reasons other than Condition A, the unit must be placed in a MODE in which the 1
LC0 does not apply. To achieve this status, the unit must l
be placed in at least MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in MODE 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.
t SURVEILLANCE SR 3.7.1.1 i
REQUIREMENTS This SR ensures adequate long term (30 days) cooling can be i
maintained. With the [ UHS dedicated water supply] below the minimum level, the affected [RSW] subsystem must be declared i
inoperable, The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Frequency is based on operating experience related to trending of the parameter variations l
during the applicable MODES.
i j
SR 3.7.1.2 Verification of the UHS temperature ensures that the heat removal capability of the [RSW) System is within the assumptions of the DBA analysis. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Frequency is based on operating experience related to trending of the parameter variations during the applicable MODES.
l (continued) i ABWR TS B 3.7-2 P&R, 08/02/93 1:26pm j
I
i e
B 3.7.1 UHS BASES SURVEILLANCE SR 3.7.1.3 REQUIREMENTS i
(continued)
Operating each [ UHS active component not normally operating]
for 215 minutes ensures that all [the active components]
are OPERABLE and that all associated controls are functioning properly. The 31 day Frequency is based on l
1 operating experience, the known reliability of the [ active components], the redundancy available, and the low probability of significant degradation of the [ active components] occurring between Surveillances.
SR 3.7.1.4 This SR verifies that [each UHS active component not normally operating] will automatically switch to the safety or emergency position to provide cooling water exclusively to the safety related equipment during an accident event.
i This is demonstrated by use of an actual or simulated initiation signal.
The LOGIC SYSTEM FUNCTIONAL TEST in SR 3.3.5.1.6 overlaps this SR to provide complete testing of the safety function.
Operating experience has shown that these components usually pass the SR when performed on the 18 month Frequency.
Therefore, this Frequency is concluded to be acceptable from i
1 a reliability standpoint.
+
i J
REFERENCES 1.
Regulatory Guide 1.27, Revision 2, January 1976.
i 1
i 1
l i
)
j (continued)
ABWR TS B 3.7-3 P&R, 08/02/93 1:26pm
{
RCW/RSW B 3.7.2 9
83.7 PLANT SYSTEMS B 3.7.2
[ Reactor Building Cooling Water (RCW)] System and [ Reactor Service Water (RSW) System]
BASES BACKGROUND The [RCW and RSW] Systems together are designed to provide cooling water for the removal of heat from equipment, such as the diesel generators (DGs), residual heat removal (RHR) heat exchangers, and room coolers for Emergency Core Cooling System equipment, required for a safe reactor shutdown following a Design Basis Accident (DBA) or transient. The combined RCW/RSW System also provides cooling to unit components, as required, during normal operation, shutdown, and reactor isolation modes.
Upon receipt of a loss of offsite power or loss of coolant accident (LOCA) signal, most but not all nonessential loads are automatically isolated, selected nonessential equipment such as control rod drive (CRD) pump oil coolers, instrument and service air compressor coolers, reactor water cleanup (RWCU) pump coolers and reactor internal pump (RIP) MG set coolers and the essential loads are automatically divided between
[RCW/RSW] Divisions A, B, and C.
All nonessential equipment can be manually isolated if required.
During all plant operating modes, all RCW/RSW divisions have at least one pump operating and, therefore, if a LOCA occurs, the RCW/RSW Systems will already be in operation.
The combined RCW/RSW System includes three separate subsystems (A, B, and C).
Each subsystem consists of the ultimate heat sink (UHS), and independent cooling water header, an independent service water loop, and the associated pumps, heat exchangers, piping, valves and instrumentation.
Each subsystem includes two RCW pumps, two RSW pumps, and three RCW to RSW heat exchangers.
Each subsystem is sized to provide sufficient cooling capacity to support the required safety related systems in its respective division during safe shutdown of the unit following a loss-of-coolant accident (LOCA).
Cooling water is pumped from the UHS by the RSW pump (s) in each subsystem to the supply header serving the respective RCW/RSW heat exchangers. After removing heat from the respective RCW subsystem the water is pumped back to the UHS.
In a separate closed loop, cooling water is circulated I
(continued)
ABWR TS B 3.7-1 P&R, 08/03/93 11:49am
i RCW/RSW B 3.7.2 BASES BACKGROUND by the pump (s) in each RCW subsystem through the essential (continued) components to be cooled and back through the RCW/RSW heat exchangers. Thus, the heat removed from the components by the RCW is transferred to the RSW, and then ultimately rejected to the UHS.
Subsystems A, B, and C supply cooling water to redundant equipment required for a safe reactor shutdown. Additional information on the design and operation of the RCW and RSW l
systems along with the specific equipment for which the combined RCW/RSW System supplies cooling water is provided in SSAR, Chapter 9, Ref.1). The combined three division RCW/RSW System is designed to withstand a single active or passive failure coincident with a loss-of-offsite power without losing the capability to supply adequate cooling water to equipment required for safe reactor shutdown.
Following a DBA or transient, the RCW/RSW System will operate automatically with operator action. Manual initiation of supported systems is, however, performed for some cooling operations (e.g., shutdown cooling).
APPLICABLE Sufficient water inventory is available for all [RCW/RSW]
SAFETY ANALYSES System post LOCA cooling requirements for a 30 day period with no additional makeup water source available.
The ability of the [RCW/RSW] System to support long term cooling of the reactor containment is assumed in evaluations of the equipment required for safe reactor shutdown presented in the SSAR, Chapters [9] and [15] (Refs. I and 2, respectively). These analyses include the evaluation of the long term primary containment response after a design basis LOCA.
The ability of the [RCW/RSW) System to provide adequate cooling to the identified safety equipment is an implicit assumption for the safety analyses evaluated in References 1 and 2.
The ability to provide onsite emergency AC power is dependent on the ability of the [RCW/RSW) System to cool the DGs. The long term cooling capability of the RHR, core spray, and RHR service water pumps is also dependent on the cooling provided by the [RCW/RSW) System.
The [ combined RCW/RSW) System, together with the UHS, satisfy Criterion 3 of the NRC Policy Statement.
(continued)
ABWR TS B 3.7-2 P&R, 08/02/93 4:llpm
RCU/RSW I
B 3.7.2 P
BASES LC0 The [RCW/RSW) subsystems are independent of each other to the degree that each has separate controls, power supplies, and the operation of one does not depend on the other.
In the event of a DBA, one subsystem of [RCW/RSW) is required to provide the minimum heat removal capability assumed in the safety analysis for the system to which it supplies cooling water. To ensure this requirement is met, three subsystems of [RCW/RSW) must be OPERABLE. At least two subsystems will operate, if the worst single active failure 1
occurs coincident with the loss of offsite power.
A subsystem is considered OPERABLE when both associated RCW and associated RSW pumps are OPERABLE, all three RCW/RSW heat exchangers are OPERABLE, the UHS is OPERABLE and the associate piping, valves, instrumentation and controls required to perform the safety-related functions are OPERABLE.
i The isolation of the [RCW/RSW] System to components or systems may render those components or systems inoperable, but does not affect the OPERABILITY of the [RCW/RSW) System.
4 2
APPLICABILITY In MODES 1, 2, and 3, the RCW and RSW Systems are required to be OPERABLE to support OPERABILITY of the equipment serviced by the combined Systems. Therefore, the [RCW/RSW]
System are required to be OPERABLE in these MODES.
In MODES 4 and 5, the OPERABILITY requirements of the
[RCW/RSW) System are determined by the systems it supports.
4 ACTIONS A1 4
If one RCW pump and/or one RSW pump and/or one RCW/RSW heat exchanger in the same subsystem is inoperable (i.e., if less than a minimum complement of one RCW pump, one RSW pump and two RCW/RSW heat exchanger are OPERABLE in one subsystem, action must be taken to restore inoperable component (s), and thus the subsystem affected, to OPERABLE status within 30 days.
In this condition sufficient redundant equipment is
~;
still available to provide cooling water to the required safety related components and sufficient heat removal (continued)
ABWR TS B 3.7-3 P&R, 08/02/93 4:llpm
i RCB/RSS B 3.7.2 BASES ACTIONS A.]
(continued) capacity is still available to adequately cool safety related loads, even assuming the worst case single failure.
However, in the degraded mode of this condition, overall reliability is reduced and a subsystem may not be capable of removing heat from the respective RHR heat exchanger at a rate consistent with design basis assumptions and modeling in the analysis for long term containment cooling (depending on other factors such actual UHS temperature).
With a minimum complement of one RCW pump, one RSW pump, and two RCW/RSW heat exchangers, a subsystem is capable of performing its safety related cooling function, consistent with design basis assumptions, for all required modes with the exception of containment cooling. However, beyond design basis calculations performed to support PRA success criteria (Ref. 3) demonstrate that successful operation of only one of three RHR subsystems (in the suppression pool cooling mode) is needed to prevent conditions inside the containment from exceeding its ultimate capacity (see B 3.6.2.3).
Thus, should a DBA occur while in this slightly degraded Condition, even considering a coincident worst case single failure, the combined RCW/RSW and RHR system would retain the capability to ultimately protect containment integrity.
The 30-day Completion Time is reasonable, based on the low probability of an accident occurring during the 30 days that t
a component is inoperable in one or more subsystems, the number of available redundant subsystems, the substantial cooling capability still remaining in a subsystem (s) in this i
Condition, and the expected high subsystem availability afforded by a system where most of the equipment, including the minimum required for most functions, is normally operating.
The Required Action is modified by a Note indicating that the provisions of LCO 3.0.4 are not applicable.
This is acceptable given the substantial degree of redundancy provided by the RCW/RSW and supported systems and the significant operational capability that still exists, in this marginally degraded condition.
(continued)
ABWR TS B 3.7-4 PAR, 08/02/93 4:llpm
RCW/RSW B 3.7.2 BASES ACTIONS H1 (continued)
With one [RCW/RSW) subsystem inoperable for reasons other than Condition A, the [RCW/RSW) subsystem must be restored to OPERABLE status within 7 days.
With the unit in this condition, the remaining OPERABLE [RCW/RSW] subsystems are adequate to perform the heat removal function.
However, the overall reliability is reduced because a single failure in one of the OPERABLE [RCW/RSW) subsystems could result in a potentially significant reduction in the overall heat removal capability.
The 7 day Completion Time is based on the redundant
[RCW/RSW] System capabilities afforded by the OPERABLE subsystems, and the low probability of an accident occurring during this time period.
Required Action B.1 is modified by two Notes indicating that the applicable Conditions of LC0 3.8.1, "AC Sources--
Operating," LC0 3.4.7, " Residual Heat Removal (RHR) Shutdown Cooling System--Hot Shutdown," be entered and Required Actions taken if the inoperable [RCW/RSW] subsystem results in an inoperable DG or RHR shutdown cooling, respectively.
This is in accordance with LC0 3.0.6 and ensures the proper actions are taken for these components.
t C.1 If one RCW pump and/or one RSW pump and/or one RCW/RSW heat exchanger in the same subsystem is inoperable in each of two separate subsystem, one RCW/RSW subsystem must be restored to OPERABLE status within 7 days.
In this condition, sufficient redundant equipment is still available to provide cooling water to the required safety related components and sufficient heat removal capacity is still available to adequately cool safety related loads.
However, a subsystem may not be capable of removing heat from the respective RHR heat exchanger at a rate consistent with design basis assumptions and modeling in the analysis for long term containment cooling. Nonetheless, with a minimum complement 3
of one RCW exchangers, pump, one RSW pump, and two RCW/RSW heat a subsystem is still capable to performing its safety related cooling function, consistent with design basis assumptions, for all other modes.
Furthermore, beyond design basis calculations performed to support PRA success (continued)
ABWR TS B 3.7-5 P&R, 08/02/93 4:llpm
RCW/RSW B 3.7.2 BASES ACTIONS C.)
1 criteria (Ref. 3) demonstrate that only one of three RHR subsystems (in the suppression pool cooling mode) is needed to ultimately protect containment integrity (see B 3.6.2.3).
Therefore, continued operations for a limited time is justified. However, in the degraded mode of this Condition, overall reliability and heat removal capability is reduced from that of Condition A, and thus a more restrictive Completion Time is imposed.
The 7 day Completion Time is reasonable, based on the low probability of an accident occurring during the 7 days that one or more redundant components are inoperable in each of two subsystems, the number of available redundant subsystems, the substantial cooling capability still remaining in subsystems in this Condition, and the expected high subsystem availability afforded by a system where most of the equipment, including the minimum required for most functions, is normally operating.
The Required Action is modified by a Note indicating that l
the applicable Conditions of LCO 3.4.7, " Residual Heat Removal (RHR) Shutdown Cooling - Hot Shutdown" be entered and Required Actions taken if the inoperable RCW/RSW subsystem results in an inoperable required RHR-Shutdown Cooling subsystem. This is in accordance with LC0 3.0.6 and l
ensures the proper actions are taken for these components.
O.1 and 0.2 i
i If the [RCW/RSW] subsystems cannot be restored to OPERABLE i
status within the associated Completion Times of Conditions A, B, or C, or two [RCW/RSWJ subsystems are inoperable for reasons other than Condition C, or all three RCW/RSW subsystems are inoperable, the unit must be placed in a MODE in which the LCD does not apply. To achieve this status, the unit must be placed in at least MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in MODE 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.
(continued)
ABWR TS B 3.7-6 P&R, 08/02/93 j:11pm
RCW/RSW B 3.7.2 i
i BASES i
SURVEILLANCE SR 3.7.2.1 REQUIREMENTS 3
This SR verifies the water level [in each RSW pump well of the intake structure] to be sufficient for the proper i
operation of the [RSW] pumps (net positive suction head and J
pump vortexing are considered in determining this limit).
I The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Frequency is based on operating experience related to trending of the parameter variations during the applicable MODES.
i SR 3.7.2.2 Verifying the correct alignment for each manual, power operated, and automatic valve in each [RCW/RSW) subsystem flow path provides assurance that the proper flow paths will exist for [RCW/RSW) operation.
This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since these valves were verified to be in the correct position prior to locking, sealing, or securing. A 4
valve is also allowed to be in the nonaccident position, and yet considered in the correct position, provided it can be automatically realigned to its accident position within the required time.
This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of being mispositioned are in the correct position. This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves.
This SR is modified by a Note indicating that isolation of the [RCW/RSW System to components or systems may render those compone)nts or systems inoperable, but does not affect the OPERABILITY of the [RCW/RSW] System. As such, when all
[RCW/RSW) pumps, valves, and piping are OPERABLE, but a branch connection off the main header is isolated, the
[RCW/RSW) System is still OPERABLE.
The 31 day Frequency is based on engineering judgment, is 2
consistent with the procedural controls governing valve i
2 operation, and ensures correct valve positions.
t SR 3.7.2.3 This SR verifies that the automatic isolation valves of the
[RSCW/RSW) System will automatically switch to the safety or (continued)
AL.R TS B 3.7-7 P&R, 08/02/93 4:11pm
RCW/RSW B 3.7.2 BASES SURVEILLANCE emergency position to provide cooling water exclusively to REQUIREMENTS the safety related equipment, and limited nonsafety related (continued) equipment, during an accident event. This is demonstrated by the use of an actual or simulated initiation signal.
This SR also verifies the automatic start capability of the RCW and RSW pumps that are in standby and automatic valuing in each of the standby RCW/RSW heat exchangers [and automatic start capability of required UHS active components] in each subsystem.
Operating experience has shown that these components usually pass the SR when performed at the 18 month Frequency.
Therefore, this Frequency is concluded to be acceptable from a reliability standpoint.
REFERENCES 1.
2.
3.
[later]
GF ABWR TS B 3.7-8 P&R, 08/02/93 4:11pm
CRHA HVAC EF System B 3.7.3 B 3.7 PLANT SYSTEMS B 3.7.3 Control Room Habitability Area (CRHA) HVAC System - Emergency Filtration (EF) Subsystem BASES BACKGROUND The Emergency Filtration subsystem of the CRHA HVAC System, provides a radiologically controlled environment from which the unit can be safely operated following a Design Basis Accident (DBA).
The safety related function of the Emergency Filtration subsystem used to control radiation exposure consists of two independent and redundant high efficiency air filtration trains for treatment of recirculated air or outside supply air.
Each subsystem consists of a demister, an electric heater, a prefilter, a high efficiency particulate air (HEPA) filter, an activated charcoal adsorber section, a second HEPA filter, a fan, and the associated ductwork and dampers. Demisters remove water droplets from the airstream.
Prefilters and HEPA filters remove particulate matter that may be radioactive.
The charcoal adsorbers provide a holdup period for gaseous iodine, allowing time for decay.
In addition to the s'afety related standby emergency filtration function, parts of the Emergency Filtration subsystem are operated to maintain the control room environment during normal operation. Upon receipt of the initiation signal (s) (indicative of conditions that could result in radiation exposure to control room personnel), the Emergency Filtration subsystem automatically switches to the isolation mode of operation to prevent infiltration of contaminated air into the control room. A system of dampers isolates the control room, and control room air flow is recirculated and processed through either of the two filter trains.
The Emergency Filtration subsystem is designed to maintain the control room environment for a 30 day continuous occupcncy after a DBA, without exceeding a 5 rem whole body dose.
Emergency Filtration subsystem operation in maintaining the control room habitability is discussed in the SSAR, Sections 6.4.1 and 9.4.1 (Refs. I and 2, respectively).
(continued)
ABWR TS B 3.7-1 P&R, 08/03/93 10:25am
^
r CRHA HVAC EF System B 3.7.3 l
d BASES APPLICABLE The ability of the Emergency Filtration subsystem to j
SAFETY ANALYSES maintain the habitability of the control room is an explicit assumption for the safety analyses presented in the SSAR, 4
1 Chapters 6 and 15 (Refs. 3 and 4, respectively). The isolation mode of the Emergency Filtration subsystem is assumed to operate following a loss of coolant accident, main steam line break, and fuel handling accident.
The radiological doses to control room personnel es a result of the various DBAs are summarized in Reference 4.
No single active or passive failure will cause the loss of outside or recirculated air from the control room.
The Emergency Filtration subsystem satisfies Criterion 3 of the NRC Policy Statement.
LC0 Two redundant trains of the Emergency Filtration subsystem are required to be OPERABLE to ensure that at least one is available, assuming a single failure disables the other train. Total system failure could result in exceeding a dose of 5 rem to the control room operators in the event of a DBA:
The Emergency Filtration subsystem is considered OPERABLE when the individual components necessary to control operator i
exposure are OPERABLE in both trains. A train is considered OPERABLE when its associated:
E 1
a.
Fan is OPERABLE; b.
HEPA filter and charcoal adsorber are not excessively i
restricting flow and are capable of performing their filtration functions; and i
c.
Heater, demister, ductwork, valves, and dampers are OPERABLE, and air circulation can be maintained.
t In addition, the control room boundary must be maintained, i
including the integrity of the walls, floors, ceilings, ductwork, and access doors.
t l
(continued)
ABWR TS B 3.7-2 P&R, 08/03/93 10:30am i
CRHA HVAC EF System B 3.7.3 1
BASES APPLICABILITY In MODES 1, 2, and 3, the Emergency Filtration subsystem must be OPERABLE to control operator exposure during and following a DBA, since the DBA could lead to a fission product release.
In MODES 4 and 5, the probability and consequences of a DBA l
are reduced due to the pressure and temperature limitations in these MODES. Therefore, maintaining the Emergency Filtration subsystem OPERABLE is not required in MODE 4 or 5, except for the following situations under which significant radioactive releases can be postulated:
a.
During operations with a potential for draining the reactor vessel (OPDRVs);
b.
During CORE ALTERATIONS; and c.
During movement of irradiated fuel assemblies in the primary or secondary containment.
l ACTIONS A.1 With one Emergency Filtration train inoperable, the inoperable Emergency Filtration train must be restored to 3
OPERABLE status within 7 days. With the unit in this 3
condition, the remaining OPERABLE Emergency Filtration train is adequate to perform control room radiation protection.
However, the overall reliability is reduced because a single l
failure in the OPERABLE train could result in loss of Emergency Filtration subsystem function.
The 7 day Completion Time is based on the low probability of a DBA occurring during this time period, and that the remaining 4
subsystem can provide the required capabilities.
i B.1 and B.2 In MODE 1, 2, or 3, if the inoperable Emergency Filtration f
train cannot be restored to OPERABLE status within the 4
associated Completion Time, the unit must be placed in a MODE that minimizes risk. To achieve this status, the unit must be placed in at least MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in MODE 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are (continued)
ABWR TS B 3.7-3 P&R, 08/03/93 10:25am l
1 L
CRHA HVAC EF System l'
B 3.7.3 BASES ACTIONS B.1 and B.Z (continued) reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.
t C.I. C.2.1. C.2.2. and C.2.3 The Required Actions of Condition C are modified by a Note l
indicating that LCO 3.0.3 does not apply.
If moving irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations.
Therefore, inability to suspend movement of irradiated fuel assemblies is not sufficient reason to require a reactor shutdown.
During movement of irradiated fuel assemblies in the primary or secondary containment, during CORE ALTERATIONS, or during OPDRVs, if the inoperable Emergency Filtration train cannot be restored to OPERABLE status within the required Conipletion Time, the OPERABLE Emergency Filtration train may be placed in the isolation mode. This action ensures that the remaining train is OPERABLE, that no failures that would prevent automatic actuation will occur, and that any active a
failure will be readily detected.
An alternative to Required Action C.1 is to immediately suspend activities that present a potential for releasing radioactivity that might require isolation of the control room. This places the unit in a condition that minimizes risk.
If applicable, CORE ALTERATIONS and movement of irradiated fuel assemblies in the primary and secondary containment must be suspended immediately. Suspension of these activities shall not preclude c mpletion of movement of a component to a safe positio1. Also, if applicable, actions must be initiated immediately to suspend OPDRVs to minimize the probability of a vessel draindown and subsequent potential for fission product release. Actions must continue until the OPDRVs are suspended.
(continued)
ABWR TS B 3.7-4 P&R, 08/03/93 10:25am I
1 CRHA HVAC EF System B 3.7.3 BASES ACTIONS RJ (continued) i If both Emergency Filtration trains are inoperable in MODE 1, 2, or 3, the Emergency Filtration subsystem may not i
be capable of performing the intended function and the unit is in a condition outside of the accident analyses.
Therefore, LC0 3.0.3 must be entered immediately.
t E.1. E.2. and E.3 The Required Actions of Condition E are modified by a Note i
indicating that LC0 3.0.3 does not apply.
If moving irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations.
f Therefore, inability to suspend movement of irradiated fuel assemblies is not sufficient reason to require a reactor shutdown.
During movement of irradiated fuel assemblies in the primary or secondary containment, during CORE ALTERATIONS, or during OPDRVs, with two Emergency Filtration trains inoperable, action must be taken immediately to suspend activities that present a potential for releasing radioactivity that might require isolation of the control room. This places the unit in a condition that minimizes risk.
If applicable, CORE ALTERATIONS and movement of irradiated fuel assemblies in the primary and secondary containment must be suspended immediately. Suspension of these activities shall not preclude completion of movement of a component to a safe position.
If applicable, actions must be initiated immediately to suspend OPDRVs to minimize the
{
probability of a vessel draindown and subsequent potential l
for fission product release. Actions must continue until the OPDRVs are suspended.
SURVEILLANCE SR 3.7.3.1 REQUIREMENTS This SR verifies that a train in a standby mode starts on demand and continues to operate.
Standby systems should be checked periodically to ensure that they start and function properly. As the environmental and normal operating conditions of this subsystem are not severe, testing each (continued)
ABWR TS B 3.7-5 P&R, 08/03/93 10:28am f
t
CRHA HVAC EF System B 3.7.3 BASES SURVEILLANCE SR 3.7.3.1 (continued)
REQUIREMENTS train once every month provides an adequate check on this subsystem.
Monthly heater operation dries out any moisture accumulated in the charcoal from humidity in the ambient air.
Systems with heaters must be operated for 2 10 continuous hours with the heaters energized.
Furthermore, the 31 day Frequency is based on the known reliability of the equipment and the two subsystem redundancy available.
SR 3.7.3m2 This SR verifies that the required Emergency Filtration testing is performed in accordance with the Ventilation Filter Testing Program (VFTP).
The Emergency Filtration filter tests are in accordance with Regulatory Guide 1.52 (Ref. 5).
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).
Specific test frequencies and additional information are discussed in detail in the VFTP.
This SR verifies that each Emergency Filtration train subsystem starts and operates on an actual or simulated initiation signal.
The LOGIC SYSTEM FUNCTIONAL TEST in SR 3.3.7.1.5 overlaps this SR to provide complete testing of the safety function.
The 18 month Frequency is specified in Reference 5.
SR 3.7.3.4 This SR verifies the integrity of the control room enclosure and the assumed inleakage rates of potentially contaminated air.
The control room positive pressure, with respect to potentially contaminated adjacent areas, is periodically tested to verify proper function of the Emergency Filtration subsystem.
During the emergency mode of operation, the EmergencyFiltrationscbsystemisdesignedtoslightly pressurize the control room to 3.17 to 12.68 Kg/m (.125 to
.5 inches) water gauge positive pressure with respect to (continued)
ABWR TS B 3.7-6 P&R, 08/03/93 10:28am
CRHA HVAC EF System B 3.7.3 i
BASES SURVEILLANCE SR 3.7.3.4 (continued)
REQUIREMENTS adjacent areas to prevent unfiltered inleakage. The i
Emergency Filtration subsystem is designed to maintain this positive pressure at a flow rate of [
] cfm to the control room in the isolation mode.
The Frequency of 18 months on a t
STAGGERED TEST BASIS is consistent with industry practice i
and other filtration system SRs.
1 i
REFERENCES 1.
2.
3.
4.
5.
Regulatory Guide 1.52, Revision 2, March 1978.
=.
1 4
t l
ABWR TS B 3.7-7 P&R, 08/03/93 10:25am
6 Control Room AC System B 3.7.4 B 3.7 PLANT SYSTEMS j
B 3.7.4 Control Room Habitability Area HVAC System - Control Room Air Conditioning (AC) subsystem BASES BACKGROUND The Control Room AC subsystem provides temperature control for the control room following isolation of the control room.
t The Control Room AC subsystem consists of two independent, redundant trains that provide cooling and heating of recirculated control room air.
Each train consists of heating coils, cooling coils, fans, chillers, compressors, ductwork, dampers, and instrumentation and controls to provide for control room temperature control.
The Control Room AC subsystem is designed to provide a controlled environment under both normal and accident conditions. A single train provides the required temperature control to maintain a suitable control room environment for a sustained occupancy of 12 persons. The design conditions for the control room environment are 23.9'C (72*F) and 50% relative humidity. The Control Room AC subsystem operation in maintaining the control room temperature is discussed in the SSAR, Sections 6.4 and 9.4.1 (Refs. I and 2, respectively).
APPLICABLE The design basis of the Control Room AC subsystem is to SAFETY ANALYSES maintain the control room temperature for a 30 day continuous occupancy.
The Control Room AC subsystem components are arranged in redundant safety related trains. During emergency operation, the Control Room AC subsystem maintains a habitable environment and ensures the OPERABILITY of components in the control room. A single active failure of a component of the Control Room AC subsystem, assuming a t
loss of offsite power, does not impair the ability of the system to perform its design function.
Redundant detectors and controls are provided for control room temperature control. The Control Room AC subsystem is designed in accordance with Seismic Category I requirements. The Control Room AC subsystem is capable of removing sensible (continued)
ABWR TS B 3.7-1 P&R, 08/03/93 11:42am
i Control Room AC System B 3.7.4 BASES APPLICABLE and latent heat loads from the control room, including SAFETY ANALYSES consideration of equipment heat loads and personnel (continued) occupancy requirements to ensure equipment OPERABILITY.
The Control Room AC subsystem satisfies Criterion 3 of the NRC Policy Statement.
LCO Two independent and redundant trains of the Control Room AC subsystem are required to be OPERABLE to ensure that at least one is available, assuming a single failure disables the other train. Total system failure could result in the equipment operating temperature exceeding limits.
The Control Room AC subsystem is considered OPERABLE when l
the individual components necessary to maintain the control room temperature are OPERABLE in both trains. These components include the cooling coils, fans, chillers, compressors, ductwork, dampers, and associated instrumentation and controls.
i APPLICABILITY In MODE 1, 2, or 3, the Control Room AC subsystem must be OPERABLE to ensure that the control room temperature will not exceed equipment OPERABILITY limits following control room isolation.
In MODES 4 and 5, the probability and consequences of a Design Basis Accident are reduced due to the pressure and temperature limitations in these MODES. Therefore, maintaining the Control Room AC subsystem OPERABLE is not required in MODE 4 or 5, except for the following situations under which significant radioactive releases can be postulated:
a.
During operations with a potential for draining the reactor vessel (OPDRVs);
b.
During CORE ALTERATIONS; and c.
During movement of irradiated fuel assemblies in the primary or secondary containment.
(continued) 1 ABWR TS B 3.7-2 P&R, 08/03/93 11:42am
Control Room AC System B 3.7.4 d
BASES (continued)
ACTIONS A.1 With one control room AC train inoperable, the inoperable control room AC train must be restored to OPERABLE status within 30 days.
With the unit in this condition, the remaining OPERABLE control room AC train is adequate to perform the control room air conditioning function.
However, the overall reliability is reduced because a single failure in the OPERABLE train could result in loss of the control room air conditioning function. The 30 day Completion Time is based on the low probability of an event occurring requiring control room isolation, the t
consideration that the remaining train can provide the required protection, and the availability of alternate cooling methods.
l B.1 and B.2 1
In MODE 2, or 3, if the inoperable control room AC train cannot be restored to OPERABLE status within the associated i
Completion Time, the unit must be placed in a MODE that minimizes risk.
To achieve this status the unit must be placed in at least MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in MODE 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the r
required unit conditions from full power conditions in an orderly manner and without challenging unit systems.
C.1. C.2.1. C.2.2. and C.2.3 The Required Actions of Condition C are modified by a Note indicating that LCO 3.0.3 does not apply.
If moving irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations. Therefore, inability to suspend movement of irradiated fuel assemblies is not sufficient reason to require a reactor shutdown.
During movement of irradiated fuel assemblies in the primary or secondary containment, during CORE ALTERATIONS, or during l
OPDRVs, if Required Action A.1 cannot be completed within the required Completion Time, the OPERABLE control room AC
)
train may be placed immediately in operation.
i (continued) 1 ABWR TS B 3.7-3 P&R, 08/03/93 11:42am
Control Room AC System B 3.7.4 BASES ACTIONS C.1. C.2.1. C.2.2. and C.2.3 (continued)
This action ensures that the remaining train is OPERABLE, that no failures that would prevent actuation will occur, and that any active failure will be readily detected.
An alternative to Required Action C.1 is to immediately suspend activities that present a potential for releasing 4
radioactivity that might require isolation of the control room. This places the unit in a condition that minimizes risk.
If applicable, CORE ALTERATIONS and movement of irradiated fuel assemblies in the primary and secondary containment must be suspended immediately. Suspension of these i
activities shall not preclude completion of movement of a component to a safe position. Also, if applicable, actions must be initiated immediately to suspend OPDRVs to minimize the probability of a vessel draindown and subsequent potential for fission product release. Actions must continue until the OPDRVs are suspended.
RJ If both control room AC trains are inoperable in MODE 1, 2, 4
or 3, the Control Room AC subsystem may not be capable of performing the intended function. Therefore, LC0 3.0.3 must be entered immediately.
E.1. E.2. and E.3 The Required Actions of Condition E.1 are modified by a Note indicating that LCO 3.0.3 does not apply.
If moving irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations.
Therefore, inability to suspend movement of irradiated fuel assemblies is not sufficient reason to require a reactor shutdown.
During movement of irradiated fuel assemblies in the primary or secondary containment, during CORE ALTERATIONS, or during OPDRVs with two control room AC trains inoperable, action must be taken to immediately suspend activities that present (continued)
ABWR TS B 3.7-4 P&R, 08/03/93 11:45am
Control Room AC System B 3.7.4 BASES ACTIONS E.1. E.2. and E.3 (continued) a potential for releasing radioactivity that might require isolation of the control room. This places the unit in a condition that minimizes risk.
If applicable, CORE ALTERATIONS and handling of irradiated fuel in the primary or secondary containment must be suspended immediately.
Suspension of these activities shall not preclude completion of movement of a component to a safe position. Also, if applicable, actions must be initiated i
immediately to suspend OPDRVs to minimize the probability of a vessel draindown and subsequent potential for fission i
product release. Actions must continue until the OPDRVs are suspended.
SURVEILLANCE SR 3.7.4.1 REQUIREMENTS This SR verifies that the heat removal capability of the system is sufficient to remove the assumed heat load in the control room. The SR consists of a combination of testing and calculation. The 18 month Frequency is appropriate i
since significant degradation of the Control Room AC subsystem is not expected over this time period.
REFERENCES 1.
2.
I i
i j
M ABWR TS B 3.7-5 P&R, 08/03/93 11:46am
Main Turbine Bypass System B 3.7.6 B 3.7 PLANT SYSTEMS B 3.7.6 Main Turbine Bypass System BASES BACKGROUND The Main Turbine Bypass System is designed to control steam pressure when reactor steam generation exceeds turbine requirements during unit startup, sudden load reduction, and cooldown.
It allows excess steam flow from the reactor to the condenser without going through the turbine. The bypass capacity of the system is 33% of the Nuclear Steam Supply System rated steam flow.
Sudden load reductions within the capacity of the steam bypass can be accommodated without reactor scram. The Main Turbine Bypass System consists of a three valve chest connected to the main steam lines between the main steam isolation valves and the turbine stop valves.
Each of these valves is sequentially operated by hydraulic cylinders. The bypass valves are controlled by the pressure regulation function of the Steam Bypass and Pressure Control System, as discussed in the SSAR, Section 7.7.1.8 (Ref. 1).
The bypass valves are normally closed, and the pressure regulator controls the turbine control valves, directing all steam flow to the turbine.
If the speed governor or the load limiter restricts steam flow to the turbine, the pressure regulator controls the system pressure by opening the bypass valves. Additionally, for the turbine trip and load rejection events only (Ref. 2) there is a Fast Opening Mode of turbine bypass operation.
In the Fast Opening Mode, the turbine bypass will open rapidly in response to a signal generated by the turbine trip or load rejection, independent of steam pressure. When the bypass valves open, the steam flows from the bypass chest, through connecting piping, to the pressure breakdown assemblies, where a series of orifices are used to further reduce the steam pressure before the steam enters the condenser.
APPLICABLE The Main Turbine Bypass System is assumed to function during SAFETY ANALYSES the design basis feedwater controller failure, maximum demand event, described in the SSAR, Section 15.1.2 (Ref. 2). Opening the bypass valves during the pressurization event mitigates the increase in reactor vessel pressure, which affects the MCPR during the event.
An inoperable Main Turbine Bypass System may result in an MCPR penalty.
(continued) j ABWR TS B 3.7-1 P&R, 08/06/93 4:54pm
Main Turbine Bypass System B 3.7.6 4
BASES APPLICABLE The Main Turbine Bypass System satisfies Criterion 3 of the SAFETY ANALYSES NRC Policy Statement.
(continued)
LC0 The Main Turbine Bypass System is required to be OPERABLE to limit peak pressure in the main steam lines and maintain reactor pressure within acceptable limits during events that r
cause rapid pressurization, such that the Safety Limit MCPR is not exceeded. With the Main Turbine Bypass System inoperable, modifications to the MCPR limits (LCO 3.2.2,
" MINIMUM CRITICAL POWER RATIO (MCPR)") may be applied to allow continued operation.
An OPERABLE Main Turbine Bypass System requires the bypass valves to open in response to increasing main steam line pressure or in the Fast Opening Mode, as applicable. This response is within the assumptions of the applicable analysis (Ref. 2). The MCPR limit for the inoperable Main Turbine Bypass System is specified in the COLR.
APPLICABILITY The Main Turbine Bypass System is required to be OPERABLE at 2 40% RTP to ensure that the fuel cladding integrity Safety Limit and the cladding I% plastic strain limit are not violated during the feedwater controller failure, maximur demand event. As discussed in the Bases for LCO 3.2.I,
" AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)," and LC0 3.2.2, sufficient margin to these limits exists
< 40% RTP. Therefore, these requirements are only necessary when operating at or above this power level.
ACTIONS A,1 If the Main Turbine Bypass System is inoperable (one or more bypass valves inoperable), or the MCPR limits for an inoperable Main Turbine Bypass System, as specified in the COLR, are not applied, the assumptions of the design basis transient analysis may not be met. Under such circumstances, prompt action should be taken to restore the Main Turbine Bypass System to OPERABLE status or adjust the MCPR limits accordingly. The 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Completion Time is (continued) l ABWR TS B 3.7-2 P&R, 08/06/93 4:54pm
Main Turbine Bypass Systcm 1
B 3.7.6 BASES I
ACTIONS A.1 (continued)
I reasonable, based on the time to complete the Required Action and the low probability of an event occurring during i
this period requiring the Main Turbine Bypass System.
Em1 If the Main Turbine Bypass System cannot be restored to OPERABLE status or the MCPR limits for an inoperable Main Turbine Bypass System are not applied, THERMAL POWER must be reduced to < 40% RTP. As discussed in the Applicability section, operation at < 40% RTP results in sufficient margin to the required limits, anc the Main Turbine Bypass System is not required to protect fuel integrity during the i
feedwater controller failure, maximum demand event. 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, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.
SURVEILLANCE SR 3.7.5.1 REQUIREMENTS Cycling each main turbine bypass valve through one complete cycle of full travel demonstrates that the valves are mechanically OPERABLE and will function when required. The 31 day Frequency is based on engineering judgment, is consistent with the procedural controls governing valve operation, and ensures correct valve positions. Therefore, the Frequency is acceptable from a reliability standpoint.
1 SR 3.7.5.2 The Main Turbine Bypass System is required to actuate automatically to perform its design function. This SR demonstrates that. with the required system initiation i
signals, the valves will actuate to their required position.
l The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a unit outage and because of the potential for an unplanned transient if the Surveillance were performed with the (continued)
ABWR TS B 3.7-3 P&R, 08/06/93 4:56pm
Main Turbine Bypass System B 3.7.6 BASES SURVEILLANCE SR 3.7.5.2 (continued)
REQUIREMENTS reactor at power. Operating experience has shown the 18 month Frequency, which is based on the refueling cycle,is acceptable from a reliability standpoint.
SR 3.7.5.3 This SR ensures that the TURBINE BYPASS SYSTEM RESPONSE TIME is in compliance with the assumptions of the appropriate safety analysis. The response time limits are specified in
[ unit specific documentation]. The [18] month Frequency is based on the need to perform this Surveillance under the conditions that apply during a unit outage and because of the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has shown the [18] month Frequency, which is based on the refueling cycle, is acceptable from a reliability standpoint.
REFERENCES 1.
2.
1 1
l ABWR TS B 3.7-4 P&R, 08/06/93 4:58pm i
Fuel Pool Water Level B 3.7.7 B 3.7 PLANT SYSTEMS B 3.7.7 Fuel Pool Water Level BASES i
BACKGROUND The minimum water level in the spent fuel storage pool meets the assumptions of iodine decontamination factors following a fuel handling accident.
A general description of the spent fuel storage pool design is found in the SSAR, Section 9.1.2 (Ref. 1). The assumptions of the fuel handling accident are found in the SSAR, Section 15.7.4 (Ref. 2).
APPLICABLE The water level above the irradiated fuel assemblies is an SAFETY ANALYSES explicit assumption of the fuel handling accident. A fuel handling accident is evaluated to ensure that the radiological consequences (calculated whole body and thyroid a
doses at the exclusion area and low population zone boundaries) are s 25% (NUREG-0800, Section 15.7.4, Ref. 3) of the 10 CFR 100 (Ref. 4) exposure guidelines.
A fuel handling accident could release a fraction of the fission product inventory by breaching the fuel rod cladding as discussed in the Regulatory Guide 1.25 (Ref. 5).
e The fuel handling accident is evaluated for the dropping of an irradiated fuel assembly onto the reactor core which bounds the consequences of dropping an irradiated fuel assembly onto stored fuel bundles. The consequences of a fuel handling accident inside the reactor building are documented in Reference 2.
The water level in the spent fuel storage pool provides for absorption of water soluble fission product gases and transport delays of soluble and insoluble gases that must pass through the water before being released to the reactor building atmosphere. This absorption and transport delay reduces the potential radioactivity of the release during a fuel handling accident.
The fuel pool water level satisfies Criterion 2 of the NRC Policy Statement.
(continued)
ABWR TS B 3.7-1 P&R, 08/06/93 5:03pm
Fuel Pool Water Level B 3.7.7 4
BASES (continued)
LCO The specified water level preserves the assumption of the fuel handling accident analysis (Ref. 2). As such, it is the minimum required for fuel movement within the spent fuel storage pool.
APPLICABILITY This LC0 applies whenever movement of irradiated fuel assemblies occurs in the associated fuel storage racks since the potential for a release of fission products exists.
ACTIONS A.1 Required Action A.1 is modified by a Note indicating that LCO 3.0.3 does not apply.
If moving irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations. Therefore, inability to suspend movement of irradiated fuel assemblies is not sufficient reason to require a reactor stutdown.
~
When the initial conditions for an accident cannot be met.
steps should be taken to preclude the accident from occurring. With either fuel pool level less than required, the movement of irradiated fuel assemblies in the associated storage pool is suspended immediately. Suspension of this activity shall not preclude completion of movement of an irradiated fuel assembly to a safe position. This l
effectively precludes a spent fuel handling accident from occurring.
SURVEILLANCE SR 3.7.6.1 REQUIREMENTS This SR verifies that sufficient water is available in the event of a fuel handling accident. The water level in the spent fuel storage pool must be checked periodically. The 7 day Frequency is acceptable, based on operating experience, considering that the water volume in the pool is normally stable and water level changes are controlled by unit procedures.
(continued)
)
ABWR TS B 3.7-2 P&R, 08/06/93 5:03pm 1
i Fuel Pool Water Level B 3.7.7 o.
BASES (continued) l REFERENCES 1.
2.
3.
NUREG-0800, Section 15.7.4, Revision 1, July 1981.
4.
5.
Regulatory Guide 1.25, March 1972.
)
i ABWil TS B 3.7-3 P&R, 08/06/93 5:05pm
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