ML20248H838

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Proposed Tech Specs Re Internal Pressure of Containment Sys & Containment Purge Supply & Exhaust Isolation Valves
ML20248H838
Person / Time
Site: Calvert Cliffs  
Issue date: 09/29/1989
From:
BALTIMORE GAS & ELECTRIC CO.
To:
Shared Package
ML20248F046 List:
References
NUDOCS 8910120112
Download: ML20248H838 (5)
v • d • e


Text

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CONTAINMENT SYSTEMS d

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3/4.6.1.4 INTERNAL PRESSURE The limitations on containment internal pressure ensure that 1) the j

Ia* \\**;,M containment structure is prevented from exceeding its design negative pressure b

differential with respect to the outside atmosi,here of 3.0 psig and 2) the he* Np containment peak pressure does not exceed the design pressure of 50 psig I

during LOCA ccedMienr. or a f s-l!" e 6ce. A c..udif toas.

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The maximum peak pressure expected to be obtained from a LOCA event is 47.6 psigl The limit of 1.8 psig for initial positive containment-

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pressure will limit tha total pressure to 49.4 psig which is less than the design pressure and ir

,nsistent with the accident analyses. 74e m *. x;mu m r*~k pressure. ups :

d to be ebkraed frw ~ sfean (;ne bec~k esc ~f' tuiltd m./a; _,J p,,, m j jg,,5 is 4'i2.psip a.rsm.

n au 3/4.6.1.5 AIR TEMPERATUR

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The. limitation on containment everage air temperature ensures that the containment peak air temperature does not exceed the design temperature of 276'F during LOCA conditions. The containment temperature limit is consis-tent'with the accident analyses.

3/4.6.1s6 CONTAINMENT STRUCTURAL INTEGRITY This limitation ensures that the structural integrity of the contain-ment will be maintained ccmparable to the original design standards for the

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life of the facility. Structural integrity is required to ensure that the containment will withstand the maximum pressure of 47.6 psig in the event of a LOCA. The measurement of containment tendon lift off force, the visual and metallurgical examination of tendons, anchcrages and liner and the Type A leakage tests are sufficient to demonstrate this capability.

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i The surveillance requirements for demonstrating the containw nt's structural integrity are consistent with the intent of the recommendations e nca rj

@88-of Regulatory Guide 1.35 "Ir. service Surveillance of Ungrouted Tendons in oo Prestressed Concrete Containment StructuresJ January 1976.

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The end anchorage concrete exterior surfaces are checked visually for u

indications of abnormal material behavior during tendon surveillance.

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Inspections of pre-selected concrete crack patterns are performed during the Type A containment leakage rate tests, consistent with the Structural l

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5 Integrit.) Test.

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1 CONTAINMENT SYSTEMS BASES 3/4.6.2 DEPRESSURIZATION AND. COOLING SYSTEMS 3/4.6.2.1 CONTAINMENT SPRAY SYSTEM The OPERABILITY of the containment spray system ensures that con-tainment depressurization and cooling capability will be available in the event of a LOCA.

The pressure reduction and resultant lower con-tainment leakage rate are consistent with the assumptions used in the accident analyses.

3/4.6.2.2 CONTAINMENT COOLING SYSTEM The OPERABILITY of the containment cooling system ensur'es that 1) the containment air temperature will be maintained within limits during normal operation, and 2) adequate heat removal capacity is available when Operated in :cnjuncticn "ith the cent;inment :pr:y :ystem> during post-LOCA conditions.

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3/4.6.3 IODINE REMOVAL SYSTEM The OPERABILITY of the containment iodine filter trains ensures that sufficient iodine removal capability will be ava'ilable in the event of a LOCA.

The reduction in containment iodine inventory' reduces the resulting site boundary radiation doses associated with containment leakage.

The operation of this sys. tem and resultant iodine removal capscity are consistent with the assumptions used in the LOCA analyses.

3/4.6.4 CONTAINMENT ISOLATION VALVES The OPERABILITY of the 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 atmos-;

phere or pressurization of the containment.

Containment isolation within the time limits specified ensures that the release of radioactive mate-rial to the environment will be consistent with the assumptions used in the analyses for a LOCA.

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POWER DISTRIBUTION LIMITS BASES and Local Power Density - Hign LCOs and LSSS setpoints remain valid. An-AZIMLTTHAL POWER TILT > 0.10 is not expected and if it should occur subsequent operation would be restricted to only those operations required to identify the cause of this unexpected tilt.

that must be used in the equation Fh = F (1 + T )

The value of Tq q

and F +Fr (1 + T ) is the measured tilt.

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T The surveillance requirements for verifying that F F and T are within 7

q T

their limits provide assurance that the actual values of F,y, F and T do'not 7

q Verifying Ffy and Ff after each fuel loading prior exceed the assumed values.

to exceeding 75% of RATED THERMAL POWER provides additional assurance that the core was properly leaded.

3/4.2.5 DNB PARAMETERS The limits on the DNB related parameters assure that each of the parameters are maintained within the nonnal steady state envelope of operation assumed in the transient and accident analyses. The limits are consistent with the safety analyses assumptions and have been analytically demonstrated adequate to main-tainaminimumQNBRofI hroughout each analyzed transient.

l In addition to the DNB criterion, there are two other criteria which set the specification in Figure 3.2-4.

The second criterion is to ensure that the existing core power distribution at full power is less severe than the power distribution factored into the small-break LOCA analysis. This results in a limitation on'the allowed negative AXIAL SHAPE INDEX value at full power. The third criterion is to maintain limitations on peak linear heat rate at low power levels resulting from Anticipated Operational Occurrences (A00s).

Figure 3.2-4 is used to assure the LHR criteria for this condition because the linear heat rate LCO, for both ex-core and in-core monitoring, is set to Min-tain on1,y the LOCA kw/ft requirements which are limiting at high power levels.

At reduced power levels, the kw/ft requirements.of certain A00s (e.g., CEA withdrawal), tend to become more limiting than that for LOCA.

The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> periodic surveillance of these parameters through instrument readout is sufficient-to ensure that the parameters are restored within their limits following load changes and other expected transient operation. The 18 month periodic measurement of the RCS total flow rate is adequate to detect flow degradation and er.sure correlation of the flow indication channels with measured flow such that the indicated percent flow will provide sufficient verification of flow rate on a 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> basis.

i CALVERT CLIFFSi-UNIT 2 '

B 3/4 2-2 Amendment No. J8 JJ,38,57. 90 9-43

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CONTAINMENT SYSTEMS BASES 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS 3/4.6.2.1 CONTAINMENT SPRAY SYSTEM The OPERABILITY of the containment spray system ensures that con-tainment depressurization and cooling capability will be available in the event of a LOCA.

The pressure reduction and resultant lower con-tainment leakage rate are consistent with the assumptions used in the accident analyses.

3/4.6.2.2 CONTAINMENT COOLING SYSTEM The OPERABILITY of the containment cooling system ensures that 1) the containment air temperature will be maintained within limits during normal operation, and 2) adequate heat removal capacity is available "her operated in conjunctier "ith the containmer.t spr:y system:-during l

post-LOCA conditions.

3/4.6.3 IODINE REMOVAL SYSTEM The OPERABILITY of the containment iodine filter trains ensures that sufficient iodine removal capability will be available in the event of a LOCA.

The reduction in containment iodine inventory reduces the resulting site boundary radiation doses associated with containment 1

leakage. The operation of this system and resultant iodine removal capacity are consistent with the assumptions used in the LOCA analyses.

3/4.6.4 CONTAINMENT ISOLATION VALVES The OPERABILITY of the 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 atmos-phere or pressurization of the containment.

Containment isolation within the time limits specified ensures that the release of radioactive mate-l rial to t"ne environment will be consistent with the assumptions used in the analyses for a LOCA.

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8ASES 3/4'.6.1.4 INTERMAL PRESSURE

,The limitations on containment internal: pressure ensure that 1) the containment structure is prevented from exceeding its design negative pressure differential with respect to the outside atmosphere of 3.0 psig and 2) the containment. peak pressure does not exceed the design ressure

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of 50 psig during LOCA :rdth~. er s te a.- /tae brak c. dl /.ss.

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The' maxi peak pressure expected to be obta ned from a LOCA event is 47.6 psig, The limit of 1.8 psig for initial positive containment l

pressure will limit the' total pressure to 49.4 psig which is less than m.x te m

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the design press and s con istent with the a cident anal ses. The.res k s'es>t is

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( g, g,,;g, 3/4.6.1.5 AIR TEMPERATURE The limitation on containment average air temperature ensures that the containment peak air temperature does not exceed the design temperature of 276'T during LOCA conditions. The containment temperature limit is consistent with the accident analyses.-

3/4.6.1.6 CONTAINMENT STRUCTURAL INTEGRITY Thisl limitation ensures that the structural integrity of the' contairiment

- will be maintained comparable to the original design standards for the life of the facility.. Structural integrity is required to ensure that the contain-ment will withstand the maximum pressure of 47.6 psig in the event of a LOCA.

The measurement of containment tendon lift off force, the visual and metal-lurgical examination of tendons, anchorages and liner and the Type A leakage tests'are sufficient to demonstrate this capability.

The surveillance requirements for demonstrating the containment's structural integrity are consistent with the intent of the recommendations of Regulatory Guide 1.35 " Inservice Surveillance of Ungrouted Tendons in Prestressed Concrete Containment Structures", January 1976.

.The end anchorage concrete exterior surfaces are checked visually for indications of abnormal material behavior during tendon surveillance.

Inspections of pre-selected concrete crack patterns are performed during the Type A containment leakage rate tests, consistent with the Structural Integrity Test.

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