ML20045D809
| ML20045D809 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 06/24/1993 |
| From: | NORTHERN STATES POWER CO. |
| To: | |
| Shared Package | |
| ML20045D805 | List: |
| References | |
| NUDOCS 9306300085 | |
| Download: ML20045D809 (6) | |
Text
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1 TS.4.4-2 I
2.
Initial and periodic type B (except airlocks) and type C tests of penetrations shall be performed at a pressure of 46 psig (P ) in l
accordance with the provisions of Appendix J,Section III.B and Section III.C, and Specification 4.4.A.S.
The airlocks shall be
'j tested initially and at six-month intervals at 46 psig by pressurizing 4
the inner volume.
In addition, when CONTAINMENT INTEGRITY is l
required, each airlock shall be tested every 3 days if it is in use by pressurizing the intergasket space to 10 psig.
3.
Type A tests will be considered to be satisfactory if the acceptance criteria delineated in Appendix J,Section III.A are met.
4.
Type B and C tests will be considered to be satisfactory if the combined leakage rate of all components subjected to Type B and C tests does not exceed 60% of the L, and if the following conditions are met.
a.
For pipes connected to systems that are in the auxiliary building special ventilation zone, the total leakage past isolation valves shall be less than 0.1 weight percent per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at pressure P..
b.
For pipes connected to systems that are exterior to both the shield building and the auxiliary building special ventilation zone, the total leakage past isolation valves shall be less than 0.01 weight percent per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at pressure P,.
c.
For airlocks, the leakage shall be less than 1% of the L, at 10 psig for door intergasket tests and 5% of the L, at 46 psig for overall airlock tests.
5.
The retest schedules for Type A, B, and C tests will be in accordance with Section III.D of Appendix J.
Each shield building shall be retested in accordance with the Type A test schedule for its containment. The auxiliary building special ventilation zone shall be retested in accordance with the Type A test schedule for Unit 1 centainment.
6.
Type A, B and C tests will be in accordance with Section V of Appendix J.
Inspection and reporting requirements of each shield building test shall be the same for Type A tests.
The auxiliary building special ventilation zone shall have the same inspection and reporting requirements as for the Type A tests of Unit 1.
J 9306300085 930624 PDR ADDCK 05000282 p
l B.3.6-1 3.6 CONTAINMENT SYSTEM Bases The OPERABILITY of the containment isolation valves ensures that the containment atmosphere will be isolated from the outside environment in i
the event of a release of radioactive material to the containment atmosphere or pressurization of the containment.
The opening of normally closed containment isolation valves on an intermittent basis under administrative control includes the following considerations: (1) designation of an operator who is in constant communication with the control room and capable of closing the affected j
valve (s) within one minute, (2) instructing this operator to close these valves in ar. accident situation, and (3) assuring that environmental i
conditions will not preclude access to close the valves and that this action will prevent the release of radioactivity outside the containment.
i Proper functioning of the Shield Building vent system is essential to the performance of the containment system.
Therefore, except for reasonable i
periods of maintenance outage for one redundant chain of equipment, the system should be wholly in readiness whenever above 200'F.
Proper functioning of the auxiliary building special vent system and isolation of the auxiliary building normal vent system are similarly necessary to preclude possible unfiltered leakage through penetrations that enter the l
special ventilation zone.
The auxiliary building special ventilation zone and its associated ventilation system have been designed to serve as secondary containment following a loss of coolant accident (Reference 2).
Special care was taken to design the access doors in the boundary and isolation valves in normal ventilation systems so that AUXILIARY BUILDING SPECIAL VENTILATION ZONE INTEGRITY can be intact during reactor operation.
The zone can l
perform its accident function with openings if they can be closed within 6 minutes, since the accident analysis assumed direct leakage of primary containment atmosphere to the environs when the shield building is at positive pressure (6 minutes). As noted in Reference 2, part of the Shield Building is part of the Auxiliary Building Special Ventilation Zone Integrity. The part of the Shield Building which is part of the Auxiliary Building Special Ventilation Zone is subject to the Technical l
Specifications of the Shield Building Integrity and not thosc~ associated with Auxiliary Building Special Ventilation Zone Integrity.
l The action statement which allows Shield Building Integrity to be lost for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> will allow for minor modifications to be made to the Shield Building during power operations.
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The COLD SHUTDOWN condition precludes any energy release or buildup of l
containment pressure from flashing of reactor coolant in the event of a l
system break.
l The shutdown margin for the COLD SHUTDOWN condition assures sub-criticality with the vessel closed, even if the most reactive rod control cluster assembly were inadvertently withdrawn.
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l B.3.6-2 3.6 CONTAINMENT SYSTEM l
Bases continued The 2 psig limit on internal pressure provides adequate margin between the maximum internal pressure of 46 psig and the peak accident pressure resulting from the postulated Design Basis Accident (Reference 1)..
The containment vessel is designed for 0.8 psi internal vacuum, the occurrence of which will be prevented by redundant vacuum breaker systems.
The containment has a nil ductility transition temperature of 0*F.
Specifying a minimum temperature of 30*F will provide adequate margin j
above NDTT during power operation when containment is required.
l The conservative calculation of off-site doses for the loss of coolant accident (References 2, 4) is based on an initial shield building annulus-air temperature of 60*F and an initial containment vessel air temperature of 104*F.
The calculated period following LOCA for which the shield building annulus pressure is positive, and the calculated off-site doses are sensitive to this initial air temperature difference.
The specified 44*F temperature difference is consistent with the LOCA accident analysis (Reference 4).
The initial testing of inleakage into the shield building and the l
auxiliary building special ventilation zone (ABSVZ) has resulted in greater specified inleakage (Figure TS.4.4-1, change No. 1) and the necessity to deenergize the turbine building exhaust fans in order to achieve a negative pressure in the ABSVZ (TS.3.6.E.2).
The staff's conservative calculation of doses for these conditions indicated that changing allowable containment leak rate from 0.5% to 0.25%/ day would offset the increased leakage (Reference 3).
j High efficiency particulate absolute (HEPA) filters are installed before l
the charcoal adsorbers to prevent clogging of the iodine adsorbers for all emergency air treatment systems.
The Charcoal adsorbers are installed to reduce the potential release of radioiodine to the environment.
The operability of the equipment and systems required for the control of hydrogen gas ensures that this equipment will be available to maintain the
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hydrogen concentration within containment below its flammable limit during post-LOCA conditions.
Either recombiner unit is capable of controlling i
l the expected hydrogen generation associated with (1) zirconium-water reactions, (2) radiolytic decomposition of water, and (3) corrosion of j
metals within containment. These hydrogen control systems are consistent I
with the recommendations of Regulatory Guide 1.7, " Control'of Combustible Gas Concentrations in Containment Following a LOCA", March 1971.
l Air locks are provided with two doors, each of which is designed to seal against the maximum containment pressure resulting from the limiting DBA.
Should an air lock become inoperable as a result of an inoperable air lock door or an inoperable door inter.ock, power operation may continue provided that at least one OPERAB'I air lock door is cla sed.
With an air lock door inoperable, access throujh the closed or locked OPERABLE door is only permitted for repair of inopersble air lock equipment.
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B.3.6-3 3.6 CONTAINMENT SYSTEM Bases continued OPERABILITY of air locks is required to ensure that CONTAINMENT INTEGRITY maintained. Should an air lock become inoperable for reasons other than an inoperable air lock door, the air lock leak tight integrity must be restored within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or actions must be taken to place the unit in a condition for which CONTAINMENT INTEGRITY is not required.
1 References 1.
USAR, Section 5 2.
USAR, Section 10.3.4 and FSAR Appendix G 3.
Letter to NSP dated November 29, 1973 4.
Letter to NSP dated September 16, 1974 l
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B.6.4-1 i
'4.4 CONTAINMENT SYSTEM TESTS Bases The Containment System consists of a steel containment vessel, a con-l crete shield building, the Auxiliary Building Special Ventilation Zone (ABSVZ), a Shield Building Ventilation System, and an Auxiliary Building Special Ventilation System.
In the event of a loss-of-coolant accident, a vacuum in the shield building annulus will cause most leakage l
from the containment vessel to be mixed in the annulus volume and I
recirculated through a filter system before its deferred release to the environment through the exhaust fan that maintains vacuum.
Some of the l
1eakage goes to the ABSVZ from which it is exhausted through a filter. A small fraction bypasses bath filter systems.
The freestanding containment vessel is designed to accommodate the maximum internal pressure that would result from the Design Basis Acci-dent (Reference 1).
For initial conditions typical of normal operation, 120*F and 15 psia, an instantaneous double-ended break with minimum safeguards results in a peak pressure of less than 46 psig at 268'F.
The containment will be strength-tested at 51.8 psig and leak-tested at 46.0 psig to meet acceptance specifications.
l License Amendment Nos. 62 and 56 dated February 23, 1983 revised the Prairie Island Technical Specifications to conform to the requirements of Appendix J to 10 CFR Part 50.
That License Amendment approved several clarifications and exemptions to the Type B and C testing requirements of Appendix J to 10 CFR Part 50.
Those clarifications and exemptions were incorporated into the Prairie Island Technical Specifications in the form of Notes 1, 2 and 5 of Table TS.4.4-1.
Table TS.4.4-1 was subsequently relocated from the Prairie Island Technical Specifications in response to Generic Letter 91-08, " Removal of Component Lists From Technical Specifications". While the reference of these notes to specific containment penetrations was relocated out of the Technical Specifications with Table TS.4.4-1, the specific clarifications and exemptions approved by License Amendment Nos. 62 and 56 are still binding.
The applicability of the Type B and C testing clarifications and exemptions contained in Notes 1, 2 and 5 of relocated Table TS.4.4-1, to specific containment i
penetrations, is maintained in the Prairie Island Updated Safety Analysis Report.
The safety analysis (
Reference:
2, 3) is based on a conservatively l
chosen reference set of assumptions regarding the sequence of events l
relating to activity release and attainment and naintenance of vacuum in the shield building annulus and the Auxiliary Building Special Ventilation Zone, the effectiveness of filtering, and the leak rate of the containment vessel as a function of time. The effects of variation in these assumptions, including that for leak rate, has been investigated thoroughly. A summary of the items of conservatism involved in the reference calculation and the magnitude of their effect upon off-site dose l
demonstrates the collective effectiveness of conservatism in these assumptions, i
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1 B.6.4-2 i
4.4 CONTAINMENT SYSTEM TESTS 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 leahage 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 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 t'l outleakage will be through particulate and charcoal filters which exh ist 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-day doses
- <. the low population zone radius of lb miles are less than guidelines pre-sented 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 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 Hear. 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 wcrrants surveillance of their leak-tightness.
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 i
200 days after a design basis accident.
The test pressure, 350 psig, achieved either by normal system operation or hydrostatically testing 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.