ML20077Q686

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Proposed Tech Spec 4.4.D.1,which Changes Interval for Performance RHR Sys Leakage Test from Once Every 12 Months to Perform Test During Each Refueling Shutdown
ML20077Q686
Person / Time
Site: Prairie Island  Xcel Energy icon.png
Issue date: 01/13/1995
From:
NORTHERN STATES POWER CO.
To:
Shared Package
ML20077Q630 List:
References
NUDOCS 9501190317
Download: ML20077Q686 (5)
v • d • e


Text

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. TS 4.4-4 REV 62 2/23/83

b. Cold DOP testing shall be performed after each complete or partial replacement of a HEPA filter bank or after any structural maintenance on the system housing that could affect the HEPA bank bypass leakage,
c. Halogenated hydrocarbon testing shall be performed after each complete or partial replacement of a char-coal adsorber bank or after any structural maintenance on the system housing that could affect the charcoal adsorber bank bypass leakage.
d. Each circuit shall be operated with the heaters on at least 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> every month.
5. Perform an air distribution test on the HEPA filter bank  ;

after any maintenance or testing that could affect the air distribution within the systems. The test shall be performed 1 at rated flow rate (i10%). The results of the test shall show the air distribution is uniform within i20%.

C. Containment Vacuum Breakers The air-operated valve in each vent line shall be tested at quarterly intervals to demonstrate that a simulated contain-ment vacuum of 0.5 psi will open the valve and a simulated accident signal will close the valve. The check valves as well as the butterfly valves will be leak-tested during each refueling shutdown in accordance with the requirements of Speci-fication 4.4.A.2.

D. Residual Heaf Removal System

1. Those portions of the residual heat removal system external l to the isolation valves at the containment, shall be hydro- i statically tested for leakage et 12 nenth intere:1: ggp[pg i kNE55MINN3E5M- l
2. Visual inspection shall be made for excessive leakage from components of the system. Any visual leakage that cannot be stopped at test conditions shall be measured by collec-tion and weighing or by another equivalent method.
3. The acceptance criterion is that maximum allowable leakage from either train of the recirculation heat removal system components (which includes valve stems; flanges and pump seals) shall not exceed two gallons per hour when the system is at 350 psig. l l
4. Repairs shall be made as required to maintain leakage within the acceptance criterion in Specification 4.4.D.3
5. If repairs are not completed within 7 days, the reactor shall be shut down and depressurized until repairs are effected and

, the acceptance criterion in 3. above is satisfied.

1 9501190317 950113 i PDR ADDCK 05000282 p PDR .

. B.4.4-2 REV 107 7/29/93

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l 4.4 CONTAINMENT SYSTEM TESTS Bases continued i i

Several penetrations of the containment vessel and the shield building  !

could, in the event of leakage past their isolation valves, result in i leakage bein5 conve ed acrose the annulus by the penetrations themselves, l thus bypassing the function >f the Shield Building Ventilation System 1 (Reference 5). Such leakage is estimated not to exceed .025% per day.

A special zone of the auxiliary building has minimum-leakage construc-tion and controlled access, and is designated as a special ventilation zone where such leakage would be collected by aither of two redundant trains of the Auxiliary Building Special Ventilation System. This system, l when activated, will supplant the normal ventilation and draw a vacuum l throughout the zone such that all outleakage will be through particulate )

and charcoal filters which exhaust to the shield building exhaust stack.  !

l The design basis loss-of-coolant accident was initially evaluated by the AEC staff (Reference 3) assuming primary containment leak rate of 0.5%

per day at the peak accident pressure. Another conservative assumption in the calculation is that primary containment leakage directly to the ABSVZ is 0.1% per day and leakage directly to the environs is 0.01% per day. I The resulting two-hour doses at the nearest SITE BOUNDARY and 30-day doses at the low population zone radius of 14 miles are less than guidelines presented in 10CFR100. )

Initial leakage testing of the shield building and the ABSV resulted ]

in a greater inleakage than the design basis. The staff has reevaluated i doses for these higher inleakage rates and found that for a primary containment leak rate of 0.25% per day at peak accident pres-sure, the offsite doses are about the same as those initially calculated for higher primary containment leakage and lower secondary containment in-leakage (Reference 6).

The Residual Heat Removal Systems functionally become a part of the  !

containment volume during the post-accident period when their operation is changed over from the injection phase to the recirculation phase.

Redundancy and independence of the systems permit a leaking system to be isolated from the containment during this period, and the possible consequences of leakage are minor relative to those of the Design Basis ,

Accident (Reference 4); however, their partial role in containment '

warrants surveillance of their leak-tightness. l The limiting leakage rates from the recirculation heat removal system are judgment values based primarily on assuring that the components could operate without mechanical failure for a period on the order of 200 days after a design basis accident. The test pressure, 350 psig,

__te_.._; _e.t_. t.. _____, _..______...,__-.t.2.__.<__,,_ .__.<__

gives an adequate margin over the highest pressure within the system after a design basis accident. A recirculation heat removal system leakage of 2 gal /hr will limit off-site exposure due to leakage to insignificant levels relative to those calculated for leakage directly from the containment in the design basis accident.

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Exhibit C i

Prairie Island Nuclear Generating Plant Licer.se Amendment Request Dated January 13, 1995 l evised Technical Specification Pages l l

i Exhibit C consists of revised pages for the Prairie Island Nuclear Generating Plant Technical Specifications with the proposed changes incorporated. The revised pages are listed below  ;

TS.4.4-4 I B.4.4-2 ,

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TS 4.4-4  ;

i REV i

b. Cold DOP testing shall be performed after each complete '

- or partial replacement of a HEPA filter bank or after any structural maintenance on the system housing that could affect the HEPA bank bypass leakage.

c. Halogenated hydrocarbon testing shall be performed i after each complete or partial replacement of a char-coal adsorber bank or after any structural maintenance on the system housing that could affect the charcoal adsorber bank bypass leaksge.
d. Each circuit shall be operated with the heaters on at least 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> every month.
5. Perform an air distribution test on the HEPA filter bank after any maintenance or testing that could affect the air distribution within the systems. The test shall be performed at rated flow rate ( 10%). The results of the test shall show the air distribution is uniform within i20%.

C. Containment Vacuum Breakers The air-operated valve in each vent line shall be tested at quarterly intervals to demonstrate that a simulated contain-ment vacuum of 0.5 psi will open the valve and a simulated accident signal will close the valve. The check valves as well as the butterfly valves will be leak-tested during each refueling shutdown in accordance with the requirements of Speci-fication 4.4.A.2.

D. Residual Heat Removal System

1. Those portions of the residual heat removal system external to the isolation valves at the containment, shall be hydro-statically tested for leakage during each refueling shutdown.
2. Visual inspection shall be made for excessive leakage from components of the system. Any visual leakage that cannot be stopped at test conditions shall be measured by collec-tion and weighing or by another equivalent method.
3. The acceptance criterion is that maximum allowable leakage from either train of the recirculation heat removal system ,

components (which includes valve stems; flanger and pump '

seals) shall not exceed two gallons per hour when the system is at 350 psig.

4. Repairs shall be made as required to maintain leakage within the acceptance criterion in Specification 4.4.D.3
5. If repairs are not completed within 7 days, the reactor shall be shut down and depressurized until repairs are effected and the acceptance criterion in 3. above is satisfied.

l l

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, B.4.4-2 REV 4.4 CONTAINMENT SYSTEM TESTS 1

Bases continued Several penetrations of the containment vessel and the shield building could, in the event of leakage past their isolation valves, result in leakage being conveyed across the annulus by the penetrations themselves, thus bypassing the function of the Shield Building Ventilation System  ;

(Reference 5). Such leakage is estimated not to exceed .025% per day.

A special zons of the auxiliary building has minimum-leakage construc- ,

tion and controlled access, and is designated as a special ventilation zone where such leakage would be collected by either of two redundant ,

trains of the Auxiliary Building Special Ventilation System. This system, when activated, will supplant the normal ventilation and draw a vacuum throughout the zone such that all outleakage will be through particulate and charcoal filters which exhaust to the shield building exhaust stack.

The design basis loss-of-coolant accident was initially evaluated by the AEC staff (Reference 3) assuming primary containment leak rate of 0.5%

per day at the peak accident pressure. Another conservative assumption in ,

the calculation is that primary containment leakage directly to the ABSVZ is 0.1% per day and leakage directly to the environs is 0.01% per day.

The resulting two-hour doses at the nearest SITE BOUNDARY and 30-dsy doses at the low population zone radius of lb miles are less than guidelines presented in 10CFR100,

'i Initial leakage testing of the shield building and the ABSV resulted in a greater inleakage than the design basis. The staff has reevaluated doses for these higher inleakage rates and found that for a primary containment leak rate of 0.25% per day at peak accident pres-sure, the offsite doses are about the same as those initially calculated for higher primary containment leakage and lower secondary containment j in-leakage (Reference 6).

The Residual Heat Removal Systems functionally become a part of the '

containment volume during the post-accident period when their operation is changed over from the injection phase to the recirculation phase.

Redundancy and independence of the systems permit a leaking system to be isolated from the containment during this period, and the possible consequences.of leakage are minor relative to those of the Design Basis Accident (Reference 4); however, their partial role in containment warrants surveillance of their leak-tightness. l The limiting leakage rates from the recirculation hett removal system are judgment values based primarily on assuring that the components could operate without mechanical failure for a period on the order of i 200 days after a design basis accident. The test pressure, 350 psig, l gives an adequate margin over the highest pressure within the system after I a design basis accident. A recirculation heat removal system leakage of 2 gal /hr will limit off-site exposure due to leakage to insignificant levels ,

relative to those calculated for leakage directly from the containment in the design basis accident. j l

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