ML17289A332
| ML17289A332 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 02/14/1992 |
| From: | WASHINGTON PUBLIC POWER SUPPLY SYSTEM |
| To: | |
| Shared Package | |
| ML17289A331 | List: |
| References | |
| NUDOCS 9202240273 | |
| Download: ML17289A332 (10) | |
Text
BASES INDEX SECTION 3/4.5 EMERGENCY CORE COOLING SYSTEMS PAGE 3/4.5.1 and 3/4.5.2 ECCS - OPERATING and SHUTDOWN......-...
B 3/4 5"1 3/4.5.3 SUPPRESSION CHAMBER..............................
B 3/4 5-2 3/4.6 CONTAINMENT SYSTEMS 3/4. 6. 1 PRIMARY CONTAINMENT Primary Containment Integrity..................--
Primary Containment Leakage.............'......;
Primary Containment Air Locks.................
MSIV Leakage Control System.........,.............
Primary Containment Structural Integrity.........
Drywell and Suppression Chamber Internal ressure.......................................
P Drywell Average Air Temperature.
Drywell and Suppression Chamber Purge System...
B 3/4 6-1 B 3/4 6-1 B 3/4 6-1 B 3/4 6-1 B 3/4 6-2 B 3/4 6-2 8 3/4 6-2 B 3/4 6"2 3/4.6.2 DEPRESSURIZATION SYSTEMS.........................
B 3/4 6-3 3/4. 6. 4 VACUUM RELIEFo
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3/4. 6. 3 PRIMARY CONTAINMENT ISOLATION VALVES.............
B 3/4 6-4 B 3/4 6-X5 3/4.6.5 SECONDARY CONTAINMENT............................
8 3/4 6-5 3/4.6.6 PRIMARY CONTAINMENT ATMOSPHERE CONTROL.......-...
B 3/4 6-5 3/4. 7 PLANT SYSTEMS 3/4.7.1 SERVICE WATER SYSTEMS.......'....................
B 3/4 7,"1 3/4. 7. 2 CONTROL ROOM EMERGENCY FILTRATION SYSTEM.........
3/4.7.3 REACTOR CORE ISOLATION COOLING SYSTEM............
B 3/4 7-1 B 3/4 7-1 3/4.7. 4 NUB8 ERS t ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ e
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WASHINGTON NUCLEAR - UNIT 2 9202240273 920214 PDR
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C 3/4.6 CONTAINMENT SYSTEMS BASES 3/4. 6. 1 PRIMARY CONTAINMENT 3/4. 6. 1. 1 PRIMARY CONTAINMENT INTEGRITY PRIMARY CONTAINMENT INTEGRITY ensures that the release of radioactive mate-rials from the containment atmosphere will be restricted to those leakage paths and associated leak rates assumed in the safety analyses.
This restriction, in conjunction with the leakage rate limitation, will limit the SITE BOUNDARY radiation doses to within the limits of 10 CFR Part 100 during accident conditions.
3/4. 6. 1. 2 PRIMARY CONTAINMENT LEAKAGE The limitations on primary containment leakage rates ensure that the total containment leakage volume will not exceed the value assumed in the safety analyses at the peak. accident pressure of 34.7 psig, P
As an added conserva-tism, the measured overall integrated leakage rate is further limited to less than or equal to 0.75 L
during performance of the periodic tests to account for possible degradation of the containment leakage barriers between leakage tests.
a Operating experience with the main steam line isolation valves has indicated that degradation has occasionally occurred in the leak tightness of the valves; therefore the special requirement for testing these, valves.
The surveillance testing for measuring leakage rates is consistent with the requirements of Appendix J of 10 CFR Part 50 with the exception of exemp-tions granted for main steam isolation valve leak testing and testing the airlocks after each opening.
3/4.6.1. 3 PRIMARY CONTAINMENT AIR LOCKS The limitations on closure and leak rate for the primary containment air locks are required to meet the restrictions on PRIMARY CONTAINMENT INTEGRITY and the primary containment leakage rate given in Specifications 3.6. 1. 1 and 3.6. 1.2.
The specification makes allowances for the fact that there may be long periods of time when the air locks will be in a closed and secured posi-tion during reactor operation.
Only one closed door in each air lock is required to maintain the integrity of the containment.
3/4.6. 1.4 MSIV LEAKAGE CONTROL SYSTEM Calculated doses resulting from the maximum leakage allowance for the main steamline isolation valves in the postulated LOCA situations would be a small fraction of the 10 CFR Part 100 guidelines, provided the main steam line system from the isolation valves up to and including the turbine condenser remains intact.
Operating experience has indicated that degradation has occasionally occurred in the leak tightness of the MSIVs such that the specified leakage requirements have not always been maintained continuously.
The requirement for the leakage control system will reduce the untreated leakage from the MSIVs when isolation of the primary system and containment is required.
WASHINGTON NUCLEAR - UNIT 2 8 3/4 6-1
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BASES 3/4. 6. l. 5 PRIMARY CONTAINMENT STRUCTURAL INTEGRITY This limitation ensures that the structural integrity of the containment steel vessel will be maintained comparable to the original design standards for the life of the unit.
Structural integrity is required to ensure that the containment will withstand the maximum pressure of 34.7 psig in the event of a LOCA.
A visual inspection in conjunction with Type A leakage tests is sufficient to demonstrate this capability.
3/4.6. 1.6 DRYMELL AND SUPPRESSION CHAMBER INTERNAL PRESSURE The limitations on drywell and suppression chamber internal pressure ensure that the containment peak pressure of 34.7 psig does not exceed the design pressure of 45 psig during LOCA conditions or that the external pressure differential does not exceed the design maximum external pressure differential of 2 psid.
The limit of l.'5 psig for initial positive contain-ment pressure will limit the total pressure to 34.7 psig which is less than the design pressure and is consistent with the safety analysis.
3/4.6. 1.7 DRYWELL AVERAGE AIR TEMPERATURE The limitation on drywell average air temperature ensures that the containment peak air temperature does not exceed the design temperature of 340'F during LOCA conditions and is consistent with the safety analysis.
3/4.6. 1.8 DRYMELL AND SUPPRESSION CHAMBER PURGE SYSTEM The 24-inch and 30-inch drywell and suppression chamber purge supply and exhaust isolation valves are required to be sealed closed during plant operation since these valves have not been demonstrated capable of closing during a
LOCA or steam line break accident. 'aintaining these valves sealed closed during plant operations ensures that excessive quantities of radio-active materials will not be released via the purge system.
To provide assurance that the 24-inch and 30-inch valves cannot be inadvertently opened, they are sealed closed in accordance with Standard Review Plan 6.2.4, which includes mechanical devices to seal or lock the valve closed or prevent power from being supplied to the valve operator.
The use of the drywell and suppression chamber purge lines is restricted to the 2-inch purge supply and exhaust isolation valves since, unlike the 24-inch and 30-inch valves, the 2-inch valves will close during a LOCA or steam line break accident and therefore the SITE BOUNDARY dose guidelines of 10 CFR Part 100 would not be exceeded in the event of an accident during PURGING operations.
The design of the 2-inch purge supply and exhaust isolation valves meets the requirements of Branch Technical Position CSB 6-4, "Containment Purging During Normal Plant Operations."
Leakage integrity tests with a maximum allowable leakage rate for purge supply and exhaust isolation valves will provide early indication of resilient material seal degradation and will allow the opportunity for repair before gross leakage fai lure develops.
The 0.60 L
leakage limit shall not be exceeded when the leakage rates determined 3y the leakage integrity tests of those valves are added to the previously determined total for all valves and penetrations subject to Type B and C tests.
MASHINGTON NUCLEAR - UNIT 2 8 3/4 6-2
INSERT Design specifications require the system to accommodate a leak rate of five times the Technical Specification leakage allowed for the MSIVs while maintaining a
negative pressure downstream of the MSIVs.
The allowed leakage value per each valve is 11.5 scfm, or a total of 230 scfh (3.8 scfm)."'hen corrected for worst case
- pressure, temperature and humidity expected to be seen during surveillance testing conditions, the flow would never exceed an indicated value (uncorrected reading from local flow instrumentation) of 5 cfm.
The 30 cfm acceptance criterion provides significant margin to this design basis require-ment and provides a benchmark for evaluating long term blower performance.
The Technical Specification limit for pressure of -17" H~O W.C. was also established based on a benchmark of the installed system performance capability.
This -17" HzO W.C. provides assurance that the negative pressure criterion can be met.
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- Letter, G02-75-238, dated August 18,
"Response to Request for Information Main Steam Isolation Valve Leakage Control System"
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CONTAINMENT SYSTEM 1S BASES 3/4. 6. 2.
OEPRESSURIZATION SYSTEMS The specifications of this section ensure that the primary containment pressure will not exceed the design pressure of 45 psig during primary system b'lowdown from full operating pressure.
The suppression chamber water provides the heat sink for the reactor coolant system energy release following a postulated rupture of the system.
The suppression chamber water volume must absorb the associated decay and structural sensible heat released during 'reactor coolant system blowdown from 1020 psig.
Since all of the gases in the drywell are purged into the suppres-sion chamber air space during a loss-of-coolant
- accident, the pressure of the liquid must not exceed 45 psig, the suppression chamber maximum pressure.,
The design volume of the suppression
- chamber, water and air, was obtained by con-sidering that the total volume of reactor coolant and to be considered is discharged to the suppression chamber and that the drywell volume is purged to the suppression chamber.
Using the minimum or maximum water volumes given in this specification, containment pressure during the design basis accident is approximately 34.7 psig which is below the design pressure of 45 psig.
Haximum water volume of 128,827 fthm results in a downcomer submergence of 12 ft and the minimum volume of 127,197 ft results in a submergence approximately 4 inches less.
The majority of the Bodega tests were run with a submerged length of 4 feet and with complete condensation.
Thus, with respect to the downcomer submergence, this specification is adequate.
The maximum temperature at the end of the blowdown tested during the Humboldt Bay and Bodega Bay tests was 170 F and this is conservatively taken to be the limit for complete condensation of the reactor coolant, although condensation would occur for temperatures above DOoF Should it;be necessary to make the suppression chamber inoperable, this shall only be done as specified in Specification 3.5'.
Under full power operating conditions, blowdown from an initial suppres-
'sion chamber water temperature of 90'F results in a water temperature of approximately 135'F immediately following blowdown which is below the 200'F used for complete condensation via quencher.
devices.
At this temperature and atmospheric
- pressure, the available NPSH exceeds that required by both the RHR and core spray
- pumps, thus, there is no dependency on containment overpressure during the accident injection phase.
If both RHR loops are used for contain-ment cooling, there is no dependency on containment overpressure for post-LOCA operations.
Experimental data indicate that excessive steam condensing loads can be avoided if the peak bulk temperature of the suppression pool is maintained below 200'F during any period of relief valve operation with sonic conditions at the discharge exit for quencher devices.
Specifications have been placed on the envelope of reactor operating conditions so that the reactor can be depressurized in a timely manner to avoid the regime of potentially high suppression chamber loadings.
WASHINGTON NUCLEAR - UNIT 2 B 3/4 6-3
CONTAINMENT SYSTEMS BASES DEPRESSURIZATION SYSTEMS (Continued)
Because of the large volume and thermal capacity of the suppression
- pool, the volume and temperature normally changes very slowly and monitoring these parameters daily is sufficient to establish any temperature trends.
By requiring the suppression pool temperature to be frequently recorded during periods of significant heat addition, the temperature trends will be closely followed so that appropriate action can be taken.
The requirement for an external visual examination following any event where potentially high loadings could occur provides assurance that no significant damage was encountered.
In addition to the limits on temperature of the suppression chamber pool water, operating procedures define the action to be taken in the event a
safe'."/relief valve inadvertently opens or sticks open.
As a minimum this acti~
shall include: 'l) use of all available means to close the valve, (2) initiate suppression pool water cooling, (3) initiate reactor
- hutdown, and (4) if other safety/relief valves are used to depressurize the reactor, their discharge shall be separated from that of the stuck-open safety/relief valve to assure mixing and uniformity of energy insertion to the pool.
3/4. 6. 3 PRIMARY CONTAINMENT ISOLATION VALVES The OPERABILITY of the primary containment isolation valves ensures that the containment atmosphere will be isolated from the outside environment in the event of a release of radioactive material to the containment atmosphere or pressurization of the -containment.
Containment isolation within the time limits specified ensures for those isolation valves designed to close auto-matically that the release of radioactive material to the environment will be consistent with the assumptions used in the analyses for a LOCA.
3/4.6.4 VACUUM RELIEF Vacuum relief breakers are provided to equalize the pressure between the suppression chamber and drywell and between the reactor building and suppres-sion chamber.
This system will maintain the structural integrity of the primary containment under conditions of large differential pressures.
The vacuum breakers between the suppression chamber and the drywell must not be inoperable in tne open position since this would allow bypassing of the suppression pool in case of an accident.
There are nine pairs of valves to provide redundancy and capacity so that operation may continue indefinitely with no more than two pairs of vacuum breakers inoperable in the closed position.
MASHINGTON NUCLEAR UNIT 2 B 3/4 6-4 Amendment No.
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COHTAINMEHT SYSTEMS BASES 3/4.6.5 SECONDARY CONTAIHMEHT Secondary containment is designed to minimize any ground level release of radioactive material which may result from an accident.
The reactor building and associated structures provide secondary containment during normal opera-tion when the drywell is sealed and in service.
At other times the drywell may be open and, when required, secondary containment integrity is specified.
Establishing and maintaining a vacuum in the reactor building with the standby gas treatment system once per 18 months, along with the surveillance of the doors,
- hatches, dampers, and valves, is adequate to ensure that there are no violations of the integrity of the secondary containment.
~ The OPERABILITY of,/he standby gas treatment systems ensures that suf-ficient iodine removal dagggility will be available in the event of a LOCA.
The reduction in containmhAt@odine inventory reduces the resulting SITE SOUHOARY radiation doses asso~c" ted with containment leakage.
The operation of this system and resuitantRodjne removal capacity are consistent with the assumptions used in the LOCA aria/'s.
Continuous operation of the system with the heaters OPERABLE for ld Mrs during each 31 day period is sufficient to reduce the buildup of moisture go
+
he adsorbers and HEPA filters.*
3/4. 6. 6 PRIMARY CONTAIHMEHT ATMOSP CONTROL The OPERABILITY of the systems requi d for the detection and control o'ydrogen gas ensures that these systems be available to maintain the hydrogen concentration within the primary inment below its flammable limit during post-LOCA conditions.
Either r+~ll and suppression chamber hydrogen recombiner system is capable of controlling the expected hydrogen generation associated with (1) zirconium-water reactions, (2) radiolytic-decomposition of water, and (3) corrosion of metals within containment.
The hydrogen control system is consistent with the recommendations of Regulatory Guide 1.7, "Control of Combustible Gas Concentrations in Containment Following a LOCA," September 1976.
WASHINGTON NUCLEAR - UNIT 2 8 3/4 6-5