ML20077A899

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Proposed Tech Specs Re Suppression Chamber Water Level
ML20077A899
Person / Time
Site: Pilgrim
Issue date: 11/22/1994
From:
BOSTON EDISON CO.
To:
Shared Package
ML20077A897 List:
References
NUDOCS 9411280139
Download: ML20077A899 (8)
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Text

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ATTACHMENT B to BECo Letter 94-128 a Amended Technical Specification Paaes i

9411280139 941122 PDR ADOCK 05000293 P PDR

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LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.7 CONTAINME"T SYSTEMS (Cont) 4.7 CONTAINMENT SYSTEMS (Cont)

A. ' Primary Containment (Cont)

k. The differential pressure may be reduced to less than 1.17 psid for a maximum of four (4) hours for maintenance activities on the differential pressure control system and during required operability testing of the HPCI system, the relief valves, the RCIC system and the drywell-suppression chamber vacuum breakers.
1. If the specifications of Item 1, above, cannot be met, and the differential pressure cannot be restored within the subsequent (6) hour period, an orderly shutdown shall be initiated and the react,r shall be in a cold shutdown condition in twenty-four (24) hours.
m. Suppression chamber water level shall be maintained l between -6 to -1 inches on torus level instrument which corresponds to a downcomer submergence of 3 feet to 3 feet 5 inches.
n. The suppression chamber can be drained if the conditions as specified in Sections 3.5.F.3 and 3.5.F.S of this Technical Specification are adhered to.

Amendment No. 17;-113 3/4.7-3

. ~ _ , - .. - + , . -. . .- --- -- - ~ . - - - - - - .- .

2 .

BASES:

3/4.7 CONTAINMENT SYSTEMS A. . Primary Containment

. The integrity of the primary containment and operation of the core standby cooling system in combination limit the off-site doses to values less than those suggested in 10CFR100 in the event of a break in the primary-system piping. Thus, containment integrity is specified whenever the potential for violation of the primary reactor system integrity exists. Concern'about such a violation exists whenever'the. reactor is critical and above atmospheric pressure. An exception was made to this requirement during initial core loading and while the low power test program was being conducted and ready access to the reactor vessel was required. There was no pressure on the 7 system at this time, thus greatly reducing the chances of a pipe break.

Should this type of testing be necessary in the future, the reactor may be taken critical; however, restrictive operating procedures would be in effect again to minimize the probability of an accident. Procedures and the Rod Worth Minimizer would limit control worth such that a rod drop would not result in any fuel damage. In addition, in the unlikely event that an

, excursion did occur, the secondary containment and standby gas treatment j system, which shall be operational during this time, offer a sufficient barrier to keep off-site doses well below 10CFR100 limits.

The pressure suppression pool water provides the heat sink for the reactor primary system energy release following a postulated rupture of the system.

The pressure suppression chamber water volume must absorb the associated decay and structural sensible heat released during primary system blowdown from 1035 psig. Since all of the gases in the drywell are purged into ti:e pressure -

j suppression chamber air space during a loss-of-coolant accident, the pressure resulting from isothermal compression plus the vapor pressuie of the liquid must not exceed 62 psig, the suppression chamber maximum pressure. The design volume of the suppression chamber (water and air) was obtained by considering that the total volume of reactor coolant to be condensud 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 the specification, containment pressure during the design basis accident is approximately 45 psig which is below the maximum of 62 psig. Maximum water volume of 94,000 ft3 results in a downcomer submergency of 4'-0" and the minimum volume of 84,000 ft3 results in a submergence approximately 12-inches less.- Mark I Containment Long Term Program Quarter Scale Test Facility (QSTF) testing was performed at various downcomer submergences and a number of wetwell to drywell pressure di f ferentials. The results of these tests were used to demonstrate that a downcomer submergence of 3'0" to 3'5" at a wetwell to drywell differential pressure of 1.17 psi is acceptable.

Should,it be necessary to drain the suppression chamber, provision will be made to maintain those requirements as described in Section 3.5.F BASES of this Technical Specification.

Amendment No. 427-113 B3/4.7-1 ,

g b e e ~ w , * -

ATTACHMENT C to BECo Letter 94-128 Marked-up Pages from Current Technical Specifications b

PNPS

, TABLE 3.2.F SURVEILLANCE INSTRUMENTATION Minimum # of Operable Instrument Type Indication Channels Instrument # Parameter and Range Notes 2 640-29A & B Reactor Water Level Indicator 0-60" (1) (2) (3) 2 640-25A & B Reactor Pressure Indicator 0-1200 psig (1) (2) (3) 2 TRU-9044 Dyrwell Pressure TRU-9045 Recorder 0-80 psia (1) (2) (3) 2 TRU-9044 Drywell Temperature Recorder, Indicator (1) (2) (3)

TI-9019 0-400aF 2 TRU-9045 Suppression Chamber Air TI-9018 Recorder, Indicator (1) (2) (3)

Temperature 0-400aF 2 LR-5038 Suppression Chamber Water (1) (2) (3)

LR-5049 Level Recorder ;f.+7 @

m Ac 1 NA Control Rod Position 28 Volt Indicating )

Lights )

1 NA

) (1) (2) (3) (4)

Neutron Monitoring SRM, IRM, LPRM )

O to 100% power )

l Amendment No. 31;-48;-83 i

3/4.2-25

, LIMITI,NG CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.7 CONTAINMENT SYSTEMS (Cont) 4.7 CONTAINMENT SYSTEMS (Cont)

A. ' Primary Containment (Cont)

k. The differential pressure may be reduced to less than 1.17 psid for a maximum of four (4) hours for maintenance activities on the differential pressure control system and during required operability testing of the HFCI system, the relief valves, the RCIC system and the drywell-suppression chamber vacuum breakers,
1. If the specifications of Item 1, above, cannot be met, and the differential pressure cannot be restored within the subsequent (6) hour period, an orderly shutdown shall be initiated and the reactor shall be in a cold shutdown condition in twenty-four (24) hours.
m. Suppression chamber water j level shall be maintained '

between -6 to @ lnches on torus level instrument which corresponds to a downcomer submergence of 3-Ohnd--3-25 3 fe<J 4 3 EM J-[<> </ca .

fanc-respectivnly.

n. The suppression chamber can be drained if the conditions as specified in Sections 3.5.F.3 and 3.5.F.S of this Technical Specification are adhered to.

b s_iR!L

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. Amendment No. 17:-113 3/4.7-3

. BASES:

, 3/4.7 CONTAINMENT SYSTEMS ,

A. Primary Containment

. The inta,rity of the primary containment and operation of the core standby cooling system in combination limit the off-site doses to values less than these suggested in 10CFR100 in the event of a break in the primary system piping. Thus, containment integrity is specified whenever the potential for violation of the primary reactor system integrity exists. Concern about such a violation exists whenever the reactor is critical and above atmospheric pressure. An exception was nade to this requirement during initial core loading and while the low power test program was being conducted and ready access to the reactor vessel was required. There was no pressure on the system at this time, thus greatly reducing the chances of a pipe break.

Should this type of testing be necessary in the future, the reactor may be taken critical; however, restrictive operating procedures would be in effect again to minimize the probability af an accident. Procedures and the Rod Worth Minimizer would limit control worth such that a rod drop would not t

result in any fuel damage. 'In addition, in the unlikely event that an excursion did occur, the secondary containment and standby gas treatment system, which shall be operational during this time, offer a sufficient barrier to keep off-site doses well below 10CFR100 limits.

1 The prec.sure suppression pool water provides the heat sink for the reactor primary system energy release following a postulated rupture of the system.

The pressure suppression chamber water volume must absorb the associated decay-and structural sensible heat released during primary system blowdown from 1035 psig. Since all of the gases in the drywell are purged into the pressure suppression chamber air space during a loss-of-coolant accident, the pressure resulting from isothermal compression plus the vapor pressure of the liquid

~

must not exceed 62 psig, the suppression chamber maximum pressure. The design volume of the suppression chamber (water and air) was obtained by considering

" that the total volume of reactor coolant to be condensed is discharged to the suppression chamber and that the drywell volume is purged to the suppression chamber. wca perken mf a.-f- JineJ.s Usingtheminimumormaximumwatervolumesgiveninthedpecification,  ;

containment pressure during the design basis accident is pproximately 45 psig.

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which is below the maximum of 62 psig. Maximumwatervolu(meof94,000ft 3

%M E

results in a downcomer submergency of 4'-0" and the minimum \ volume of 84,000

'Nf 3 results in a rubmergence approximately 12-inches less. NarkIContainment Long etnt-h ogram Quarter Scale Test Facility (QSTF) testing CJdowncomer submergency D T-75 bt-ant %psD wetwell to drywell pressure differentials.

showws1 nificant-suppressiar 6 hNr-loadduction-and4eng-TumJcogram antlys 6-andsnodifica t-lons-a rebosed-orsha--above-submety,enesndwliff a r a n e i n 1 ,:. 1 ,

g+rassure Should it be necessary to drain the suppression chamber, provision will be l made to maintain those requirements as described in Section 3.5.F BASES of '

3- this l'echnical Specification.  ;

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& W 4 g l A- Ob S hctr qpw^*d- M f e r<n M 3W ,f ngt, l hW gj~;v~~ision 1 h Amendment No. 42:-113 B3/4.7-1 j

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