ML20217C294
| ML20217C294 | |
| Person / Time | |
|---|---|
| Site: | McGuire, Mcguire  |
| Issue date: | 07/03/1991 |
| From: | DUKE POWER CO. |
| To: | |
| Shared Package | |
| ML20217C240 | List: |
| References | |
| NUDOCS 9107150206 | |
| Download: ML20217C294 (3) | |
Text
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4 PLANT SYSTEMS
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BASES STAN08Y NUCLEAR SERVICE WATER POND (Continued)
The limitations on minimum water level and maximum temperature are based on providing a 30-day cooling water supply to safety-related equipment without exceeding their design basis temperature and is consistent with the reconmend-ations of Regulatory Guide 1.27, " Ultimate Heat Sink for Nuclear Plants,"
March 1974.
The Surveillance Requirements specified for the das inspection will confom to the recommendations of Regulatory Guide 1.127, Revision 1, March 1978.
3/4.7.6 CONTROL AREA VENTILATION SYSTEM i
The OPERA 81LITY of the Control Area Ventilation Systes ensures that:
(1) the ambient air temperature does not exceed the allowable temperature for continuous duty rating for the equipment and instrumentation cooled by this systes, and (2) the control room will remain habitable for operations personnel during and following all credible accident conditions.
Cumulative operation
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of the system with the heaters on for 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> over a 31-day period is fficjent to reduce the buildup of moisture on the adsorbers and HEPA filters.
The
- NcA 1 OPERA 8ILITY of this system in conjunction with control rocca design provisions % rem is based on limiting the radiation exposure to personnel occupying the control g4 room to 5 ren or less whole body, or its equivalent.
This limitation is consistent with the requirements of General Design Criterion 19 of Appendix A, 10 CFR 50.
ANSI N510-1975 will be used as a procedural guide for surveillance testing.
3/4.7.7 AUXILIARY SUILDING FILTERED VENTILATION EXHAUST SYSTEM The OPERA 8ILITY of the Auxiliary Building Filtered Ventilation Exhaust Systes ensures that radioactive materials leaking from the ECCS equipment within the auxiliary building following a LOCA are filtered prior to reachirg the environment.
The operation of this system and the resultant effect on offsite dosage calculations w re assumed in the accident analyses.
ANSI N510-1980 will be used as a procedural guide for surveillance testing.
The methyl iodido penetration test criterion for the carbon semples has been established at 105 (i.e., 9C% rteoval) which is greater than the iodine removal in the accident analysis.
9107150206 910703 PDR ADOCK 05000369 P
PDR McGuirw - UNITS 1 and 2, B 3/4 7-4 Amendment No. 113(Unit 1)
Amendment No.
95(Unit 2)
CONTAI!NEtiT SYSTEMS BASES 1
3/4.6.1.7 REACTOR BUILDING STRUCTURAL INTEGRITY This limitation ensures that the structural integrity of the containment F
reactor building will be maintained comparacle to the original design standards f(O for the life of the facility.
Structural integrity is required to provide:
(1) protection for the steel vessel from external missiles, (2) radiation shielding in the event of a LOCA, and (3) an annulus surrounding the steel
-f-vessel that can be maintained at a negative pressure during accident condi-p tions.
A visual inspection is sufficient to demonstrate this capability.
Q 3/4.6.1.8 ANNULUS VENTILATION SYSTEM i
The OPERABILITY of the Annulus Ventilation System ensures that during LOCA Q4 conditions, containment vessel leakage into the annulus will be filtered througn 4 the HEPA filters and charcoal adsorber trains prior to discharge to the atmos-9 phere.
Cuaulative operation of the systes with the heaters on for 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> over M
_ a 31-day oeriod i sufficient to reduce the buildup of moisture on the adsorters and HEPA filter.
Rh r:;;ir:::nt 1: necessary to meet the assumptions used in t/) the accident analyses and limit the SITE BOUNDARY radiation doses to within the
' dose guidelind values of 10 CFR Part 100 during LOCA conditions.
A"5! '510-1975 A
will 5: :::d :: : pr:::dar:1 guid; f r ;ure:ill:n : t::tkg.
A0',0% O OS, T;3:
t h:d A, will 5: c:d fer Urit 1 :gracilhn;; t::th; (12:ntc~j te t) f:r g
=ey! iodide pretratier " 'i:e :f the 1e:r:t: y t::t :;::if t:d " ':ge':t: y
'Q \\
Side 1. 50, ":v. 2, "r:h 1973, *:g h:: y F::itha 0. 5.:. n AO', # r3 M
- t ::th:d h und fer : r:hth: h :idity f 05* :t 20*C.
B: :: cf ah
- t ed the sce: *ec: critechn of : ::thyl iodid: pen:tr tha Of h:: than 0.71.' : : : =: h & t d th an = d d=:nt r k:t hn ef' S h n:i:: ;f 95%.
This
-chr; rer!tM % Nr YE cy+tr hut:r =;=ity = Unit 1.
3/4.6.1.9 CONTAINMENT VENTILATION SYSTEM The containment purge supply and exhaust isolation valves for the lower compartment (24-inch) and instrument room (12-inch and 24-inch) are required to be sealed closed during plant operations sinco these valves have not been demonstrated capable of closing during a LOCA.
Maintaining these valves sealed closed during plant operation ensures that excessive quantities of radioactive material will not be released via the Containment Purge System.
To provide assurance that these containment valves cannot be inadvertently opened, the valves are sealed closed in accordance with Standard Review Plan 6.2.4 which includes mechanical devices to seal or lock the valve closed, or prevents power free being supplied to the valve operator.
The use of the containment purge lines is restricted to the purge supply and exhaust isolation valves in the upper compartment (24-inch) since, wiike the valves in the lower compartment and instrument room, the upper compartment valves will close during a LOCA.
Therefore, the SITE BOUNDARY dose guideline values of 10 CFR Part 100 would not b3 excaeded in the event of an accident during containment purging operation.
Operation with these valves open will be limited to 250 hours0.00289 days <br />0.0694 hours <br />4.133598e-4 weeks <br />9.5125e-5 months <br /> during a calendar year.
Leakage integrity tests with a maximum allowable leakage rate for contain-ment purge supply and exhaust supply valves will provide early indication of resilient material seal degradation and will allow opportunity for repair before gross leakage failures could develop.
The 0.60 L leakage limit of Specifica-tion 3.6.1.2b. shall not be exceeded when the leakage rates determined by the leakage integrity tests of these valves are added to the previously determined total fer all valves and penetrations subject to Type B and C tests.
McGUIRE - UNITS 1 and 2 B 3/4 6-3 Amendment No.109(Unit 1)
Amenament Ho. 01(Unit 2)
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1.
Tre specified laboratory test methoo, namely, ASTN 03803 89 implies that heaters may be unavailable for controlling the relative nu11dity of the influent air entering the charcos) aosorber section to 1 70 percent. This is acceptable since accident analysis with appropriate adscreer efficiencies for radiciodine in elemental and organic forms based on the asova test shows the site boundary radiation doses to be within the 10 CFR Part 100 lirits during design basis LOCA conditions. However, specifications are included to ensure heater coerability and corrective ACTION $ are identified to adcress tre contingency cf inoperable heaters; these are in place to increase the safety margin of the filters.
2.
The specified laboratory test method, namely, ASIN 03803-89, implies that heaters tay be 1.,navailable for controlling the reistive humidity of the influent air entering the charcoal adsoroer section to c 70 percent. This is acceptable, since accident analysis with appropriate adsoi9er officiercies for radiciedine in elemental and organic foris based on the above tast shows that the control room radiation doses to be within the 10 CFR Part 50, Aependix A CDC 19 limits during design basis LOCA I
conditions.
Howevar, spec.fications are included to ensure heater operability and corrective ACTIONS are identified to address the contingency of inoperable heaters; thews are in place to increase ins safety margir of the filters.
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