ML20046B632
| ML20046B632 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 07/29/1993 |
| From: | Bill Dean Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20046B633 | List: |
| References | |
| GL-91-08, GL-91-8, NUDOCS 9308050287 | |
| Download: ML20046B632 (27) | |
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UNITED STATES l
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NUCLEAR REGULATORY COMMISSION
- 'g' Cs WASHINGTON, D.C. 20555-0001
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NORTHERN STATES POWER COMPANY DOCKET NO. 50-282 PRAIRIE ISLAND NUCLEAR GENERATING PLANT. UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE 9
Amendment No.107 License No. DPR-42 l
1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Northern States Power Company (the licensee) dated May 7, 1992, as revised June 24, 1993, complies with the standards and requirements of the Atomic Energy Act of 1954, as t
amended (the Act), and the Commission's rules and regulations set 1
forth in 10 CFR Chapter I; 1
B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; j
C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifica-tions as indicated in the attachment to this license amendment, and para-graph 2.C.(2) of Facility Operating License No. DPR-42 is hereby amended to read as follows:
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'9309050287 930729 i
PDR ADOCK 05000282 p
PDR.
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Technical Specifications l
The Technical Specifications contained in Appendix A, as revised through Amendment No.103 are hereby incorporated in the license The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION ~
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Wi am M. Dean, Acting D rector Project Directorate III-I I
Division of Reactor'Projectsz III/IV/V Office of Nuclear Reactor Regulation i
Attachment-Changes to the Technical Specifications Date of Issuance: July 29,1993 l
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ATTACHMENT TO LICENSE AMENDMENT NO. 107 i
I FACILITY OPERATING LICENSE NO. DPR-42 i
DOCKET NO. 50-282 i
Revise Appendix A Technical Specifications by removing the_pages identified below and inserting the attached pages. The revised pages are identified by l
amendment number.and contain marginal lines indicating the area of change.
i REMOVE INSERT l
i TS-xii TS-xii TS.1-2 TS.1-2 i
TS.3.4-2 TS.3.4-2 l
TS.3.6-1 TS.3.6-1 TS.4.4-2 TS.4.4-2 TABLE TS.4.4-1 (Pgs 1-5)
B.3.6-1 B.3.6-1 B.3.6-2 B.3.6-2 l
B.3.6-3 B.4.4-1 B.4.4-1 B.4.4-2 B.4.4-2 l
TS-xii t
TECHNICAL SPECIFICATIONS
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LIST OF TABLES TS TABLE TITLE 3.5-1 Engineered Safety Features Initiation Instrument Limiting Set Points 3.5-2 Instrument Operating Conditions for Reactor Trip t
3.5-3 Instrument Operating Conditions for Emergency Cooling System 3.5-4 Instrument Operating Conditions for Isolation Functions 3.5-5 Instrument Operating Conditions for Ventilation Systems 3.5-6 Instrument Operating Conditions for Auxiliary Electrical System 3.9-1 Radioactive Liquid Effluent Monitoring Instrumentation 3.9-2 Radioactive Gaseous Effluent Monitoring instrumentation 3.14-1 Safety Related Fire Detection Instruments 3.15-1 Event Monitoring instrumentation - Process & Containment 3.15-2 Event Monitoring instrumentation - Radiation 4.1-1 Minimum Frequencies for Checks, Calibrations and Test of Instrument Channels 4.1-2A Minimum Frequencies for Equipment Tests 4.1-2B Minimum Frequencies for Sampling Tests 4.2-1 Special Inservice Inspection Requirements 4.10-1 Radiat! ion Environmental Monitoring Program (REMP) l Sample Collection and Analysis 4.10-2 RFMP - Maximum Values for the Lower Limits of Detection 4.10-3 RFMP - Reporting Levels for Radioactivity Concentrations in r
l Environmental Samples i
I 4.12-1 Steam Generator Tube Inspection 4.13-1 Snubber Visual Inspection Interval 4.17-1 Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirements 4.17-2 Radioactive Gaseous Effluent Monitoring instrumentation Surveillance Requirements 4.17-3 Radioactive Liquid Waste Sampling and Analysis Program 4.17-4 Radioactive Gaseous Waste Sampling and Analysis Program I
5.5-1 Anticipated Annual Release of Radioactive Material in Liquid Effluents From Prairie Island Nuclear Generating Plant (Per Unit) 5.5-2 Anticipated Annual Release of Radioactive Nuclides in Gaseous Effluent From Prairie Island Nuclear Generating Plant (Per Unit) 6.1-1 Minimum Shift Crew Composition Prairie Island Unit 1 Amendment No. 98, g Prairie Island Unit 2 Amendment No. $1, g l
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1 1
TS.1-2 r
l CONTAINMENT INTEGRITY CONTAINMENT INTEGRITY shall exist when:
1.
Penetrations required to be isolated during accident conditions are i
either-Capable of being closed by an OPERABLE containment automatic isolation a.
valve system, or b.
Closed by manual valves, blind flanges, or deactivated automatic valves secured in their closed positions, except as provided in Specifications 3.6.C and 3.6.D.
2.
The equipment hatch is closed and sealed.
3.
Each air lock is la compliance with the requirements of Specification 3.6.M.
4.
The containment leakage rates are within their required limits.
COLD SHUTDOWN A reactor is in the COLD SHUTDOWN condition when the reactor is suberitical by l
at least 1% Ak/k and the reactor coolant average temperature is less than l
200* F.
CORE ALTERATION l
l CORE ALTERATION is the movement or manipulation of any component within the reactor pressure vessel with the vessel head removed and fuel in the vessel, which may affect core reactivity.
Suspension of CORE ALTERATION shall not preclude completion of movement of a component to a safe conservative position.
CORE OPERATING LIMITS REPORT The CORE OPERATING LIMITS REPORT is the unit-specific document that provides core operating limits for the current operating reload cycle. These cycle-specific core operating limits shall be determined for each reload cycle in accordance with Specification 6.7.A.6.
Plant operation within these l
operating limits is addressed in individual specifications.
1 Prairie Island Unit 1 Amendment No. 92,107 Prairie Island Unit 2 Amendment No. $5,100
i TS.3.4-2 3.4.B.1.d.
A minimum of 100,000 gallons of water is available in the condensate storage tanks and a backup supply of river water is available through the cooling water system.
e.
Motor operated valves MV-32242 and MV-32243 (Unit 2 valves MV-32248 and MV-32249) shall have valve position monitor lights OPERABLE and shall be locked in the open position by having the motor control center supply breakers physically locked in the off position.
f.
Manual valves in the above systems that could (if one is improperly positioned) reduce flow below that assumed for accident analysis shall be locked in the proper position for emergency use.
During POWER OPERATION, changes in valve position will be under direct administrative control.
g.
The condensate supply cross connect valve C-41-2, to the auxiliary l
feedwater pumps shall be blocked and tagged open. Any changes in position of this valve shall be under direct administrative l
I control.
2.
During STARTUP OPERATION or POWER OPERATION, any one of the following conditions of inoperability may exist for each unit provided STARTUP OPERATION is discontinued until OPERABILITY is restored.
If OPERABILITY is not restored within the time specified, place the affected unit (or either unit in the case of a motor driven AFW pump inoperability) in at least HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and reduce reactor coolant system average temperature below 350* F within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, a.
A turbine driven AFW pump, system valves and piping may be inoperable for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, b.
A motor driven AFW pump, system valves and piping may be inoperable for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, c.
The condensate storage tanks may be inoperable for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> l
provided the cooling water system is available as a backup supply of water to the auxiliary feedwater pumps.
d.
The backup supply of river water provided by the cooling water system may be inoperable for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> provided a minimum of l
100,000 gallons of water is available in the condensate storage tanks.
The valve position monitor lights for motor operated valves e.
MV-32242 and MV-32243 (Unit 2 valves MV-32248 and MV-32249) may be l
inoperable for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> provided the associated valves' positions are verified to be open once each shift.
Prairie Island Unit 1 Amendment No. $1,10~7 Prairie Island Unit 2 Amendment No. $4,100 l
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TS.3.6-1
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3.6 CONTAINMENT SYSTEM Applicability Applies to the integrity of the containment system.
Obiective To define the operating status of the containment system for plant operation.
Specification A.
Containment Intenrity 1.
A reactor shall not be made or maintained critical nor shall reactor coolant system average temperature exceed 200* F unless CONTAINMENT j
INTEGRITY is maintained.
2.
If these conditions cannot be satisfied, within one hour initiate the action necessary to place the unit in HOT SHUTDOWN, and be in at least HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
l B.
Vacuum Breaker System l
1.
Both valves in each of two vacuum breaker systems, including actuating and power circuits, shall be OPERABLE when CONTAINMENT INTEGRITY is required (except as specified in 3.6.B.2 and 3.6.B.3 below).
2.
With one vacuum breaker inoperable with respect to its containment isolation function, apply the requirements of Specification 3.6.C.3, to the isolation valves associated with the inoperable vacuum breaker.
3.
One vacuum breaker may be inoperable with respect to its vacuum relief function for 7 days.
C.
Containment Isolation Valves 1.
Non-automatic containment isolation valves shall be locked closed or shall be under direct administrative control and capable of being closed within one minute following an accident when CONTAINMENT INTEGRITY is required (except as specified in 3.6.C.3 below).
2.
Automatic containment isolation valves, including actuation circuits, shall be OPERABLE when CONTAINMENT INTEGRITY is required (except as specified in 3.6.C.3 below).
3.
With one or more of the containment isolation valve (s) inoperable, within four hours:
(a) restore the inoperable valve (s) to operable status or, (b) deactivate the operable valve in the closed position or, (c) lock closed at least one valve in ecch penetration having one inoperable valve.
Prairie Island Unit 1 Amendment No. 91,107 Prairie Island Unit 2 Amendment No. $5,100
-1 TS.4.4-2 j
1 1
l 2.
Initial and periodic type B (except airlocks) and type C tests of g
penetrations shall be performed at a pressure of 46 psig (P ) in i
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accordance with the provisions of Appendix'J,Section III.B and j
Section III.C. and Specification 4.4.A.5. 'The~airlocks shall be tested initially and at six-month intervals at 46 psig by pressurizing the inner volume.
In addition, when CONTAllGENT INTEGRITY is l
required, each airlock shall be tested every 3 days if it is in use by j
pressurizing the intergasket space.to 10 psig.
3.
Type A tests will be considered to be satisfactory if tho' acceptance-i criteria delineated in Appendix J,Section III.A are met.
l 4.
Type B and C tests will be considered to be satisfactory if the
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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 i
are met.
For pipes connected to systems that are in the auxiliary building a.
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.,
i 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..
For airlocks, the leakage shall' be less thEn 1% of the L at 10 -
I c.
psig for door intergasket tests and 5% of the L at 46 psig for i
overall airlock tests, j
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 l
containment. The auxiliary building special ventilation zone shall be retested in accordance with the Type A test schedule for Unit 1
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containment.
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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.
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Prairic Island Unit 1 amendment No. 62,107 Prairie Island Unit 2 Amenda.ent No. 56,100
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 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 valve (s) within one minute, (2) instructing this operator to close these valves in an accident situation, and (3) assuring that environmental conditions will not preclude access to close the valves and that this action will prevent the release of radioactivity outside the containment.
Proper functioning of the Shield Building vent system is essential to the performance of the containment system.
Therefore, except for reasonable 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 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 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 Specifications of the Shield Building Integrity and not those associated with Auxiliary Building Special Ventilation Zone Integrity.
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.
The COLD SHUTDOWN condition precludes any energy release or buildup of containment pressure from flashing of reactor coolant in the event of a system break.
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.
Prairie Island Unit 1 Amendment No. $1,107 Prairie Island Unit 2 Amendment ho. $#,100
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B.3.6 3.6 CONTAINMENT SYSTEM J
2 133,gg continued The 2 psig limit on internal pressure provides adequate margin between th>
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.
i The containment has a nil ductility transition temperature of 0*F.
Specifying a minimum temperature of 30*F will provide adequate margin i
above NDTT during power operation when containment is required.
+
l The conservative calculation of off-site doses for the loss of coolant
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accident (References ~2,
- 4) is based on an initial shield building annulus air temperature of 60*F and an initial containment vessel air temperature-l of 104*F.
The calculated period following LOCA for which.the shield building annulus pressure is positive, and-the calculated off-site doses 1
are sensitive to this initial air temperature difference. The specified 44'F-temperature difference is consistent with the LOCA accident analysis 1
(Reference 4).
The initial testing ~of.inleakage-into the shield building and.the 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 I
achieve a negative pressure in the ABSVZ (TS.3.6.E.2):.
Tha staff's conservative calculation of doses for.these conditions indicated that changing allowable containment leak rate ~from 0.54 to 0.254/ day would:
-l offset the increased leakage (Reference 3).
j High efficiency particulate absolute (HEPA) filters are installed before j
the charcoal adsorbers to prevent clogging'of the iodine adsorbers for_all emergency air treatment systems. The Charcoal adsorber's are installed to reduce the potential release of radiciodine 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 hydrogen concentration within containment below its flammable limit during post-LOCA conditions. Either recombiner unit is capable of controlling the expected hydrogen generation associated'with (1) zirconium-water reactions, (2) radiolytic decomposition of water, and (3) corrosion of metals within containment..These hydrogen control' systems are consistent' with the' recommendations of Regulatory Guide 1.7, " Control of Combustible Gas Concentrations in Containment Following a IDCA", March' 1971.
Air locks are provided with two doors, each of which-is designed to seal against the maximum containment pressure resulting from the limiting DBA.
i Should an air lock become inoperable as. a result' of an inoperable air lock
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door or an inoperable door interlock, power operation may continue.
1 provided that at least one OPERABLELair lock door is closed. With an air
' lock door inoperable, access through'the closed or locked OPERABLE door is only permitted for repair of inoperable air lock equipment.
Prairie Island Unit 1
1 Amendment No. $/,107 Prairie Island Unit Amendment No. $4,100 ;
j
4 B.3.6-3 3.6 CONTAINMENT SYSTEM I
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JAEeg continued l
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OPERABILITY of air locks is required to ensure that CONTAINMENT INTEGRITY is l'
j maintained.
Should an air lock become inoperable for reasons other than an inoperable air lock door, the air lock leak tight integrity must be
- l 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 l
1.
USAR, Section 5 2.
USAR, Section 10.3.4 and FSAR Appendix G 3.
Letter to NSF dated November 29, 1973 4.
Letter to NSP dated September 16, 1974
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l Amendment No. $/, g Prairie Island Unit 1 Amendment No. $f,' w '
Prairie Island Unit 2
B.4.4-1 4.4 CONTAINMENT SYSTEM TEST,S Bases The Containment System consists of a steel containment vessel, a con-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 from the containment vessel to be mixed in the annulus volume and recirculated through a filter system before its deferred release to the environment through the exhaust fan that maintains vacuum.
Some of the leakage goes to the ABSVZ from which it is exhausted through a filter.
A small fraction bypasses both 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.
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 penetrations, is maintained in the Prairie Island Updated Safety Analysis Report.
The safety analysis (References 2, 3) is based on a conservatively chosen reference set of assumptions regarding the sequence of events relating to activity release and attainment and maintenance 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 demonstrates the collective effectiveness of conservatism in these assumptions.
Prairie Island Unit 1 Ame ndment No. $1,107 Prairie Island Unit 2 Amendment No. $4,100
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B.4.4-2 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 j
(Reference 5).
Such leakage is estimated not to exceed.0254 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 j
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 la that primary containment leakage directly to the ABSVZ is 0.1% per day and leakage directly to the environs is 0.01% per day.
l The resulting two-hour doses at the nearest SITE BOUNDARY and 30-day doses at the low population zone radius of lb 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 reevaluat<ed 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 j
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.
The limiting leakage rates from the recirculation heat removal system are judgment values based primarily on assuring that the components j
could operate without mechanical failure for a period on the order of 200 days after a design basis accident. The test pressure, 350 psig, j
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.
Prairie Island Unit 1 Amendment No. $f, 107 Prairie Island Unit 2 Amendment No. $9, 100 J
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E UNITED STATES NUCLEAR REGULATORY COMMISSION
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WASHINGTON, D.C. 20555-0001 NORTHERN STATES POWER COMPANY DOCKET NO. 50-306 PRAIRIE ISLAND NUCLEAR GENERATING PLANT.' UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE' Amendment No.100 License No. DPR-60 1.
The Nuclear Regulatory Commission (the Commission) h,as found that:
A.
The application for amendment by Northern States Power Company (the licensee) dated May 7, 1992, as revised' June 24, 1993, complies with the standards and requirements.of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I, B.
The facility will operate in conformity with the application,-the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and
.I safety of the public, and (ii) that such activities.will be conducted in compliancr with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all-applicable requirements' have been I
satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifica-tions as indicated in the attachment to this license amendment, and para-graph 2.C.(2) of Facility Operating License No. DPR-60 is hereby amended to read as follows:
1 l-
i Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.10Q are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the l
Technical Specifications.
I 3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Wi iam M. Dean, Acting irector Project Directorate III-I Division of Reactor Projects III/IV/V Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications l
Date of Issuance: July 29,1993 i
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1 ATTACHMENT TO LICENSE AMENDMENT NO.100 FACILITY OPERATING LICENSE NO. DPR-60 i
DOCKET NO. 50-306 Revise Appendix A Technical Specifications by removing the pages identifi'ed j
below and inserting the attached pages. The revised pages are identified by amendment number and contain marginal lines indicating the area of change.
REMOVE INSERT TS-xii TS-xii l
TS.1-2 TS.1-2 l
TS.3.4-2 TS.3.4-2 l
TS.3.6-1 TS.3.6-1 TS.4.4-2
'TS.4.4-2 l
TABLE TS.4.4-1 (Pgs 1-5) l B.3.6-1 B.3.6-1 B.3.6-2 B.3.6-2 B.3.6 -----
l B.4.4-1
-B.4.4-1 B.4.4-2 B.4.4-2 I
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6 TS-xii TECHNICAL SPECIFICATIONS LIST OF TABLES TS TABLE TITLE 3.5-1 Engineered Safety Features Initiation Instrument Limiting Set Points 3.5-2 Instrument Operating Conditions for Reactor Trip 3.5-3 Instrument Operating Conditions for Emergency Cooling System 3.5-4 Instrument Operating Conditions for Isolation Functions 3.5-5 Instrument Operating Conditions for Ventilation Systems 3.5-6 Instrument Operating Conditions for Auxiliary Electrical System 3.9-1 Radioactive Liquid Effluent Monitoring Instrumentation 3.9-2 Radioactive Gaseous Effluent Monitoring instrumentation 3.14-1 Safety Related Fire Detection Instruments 3.15-1 Event Monitoring instrumentation - Process 6 Containment 3.15-2 Event Monitoring instrumentation - Radiation 4.1-1 Minimum Frequencies for Checks, Calibrations and Test of Instrument Channels 4.1-2A Minimum Frequencies for Equipment Tests 4.1-2B Minimum.'requencies for Sampling Tests 4.2-1 Special Inservice Inspection Requirements 4.10-1 Radiat' ion Environmental Monitoring Program (REMP)
Sample Collection and Analysis 4.10-2 RFMP - Maximum Values for the Lower Limits of Detection 4.10-3 RFMP - Reporting Levels for Radioactivity Concentrations in Environmental Samples 4.12-1 Steam Generator Tube Inspection 4.13-1 Snubber Visual Inspection Interval 4.17-1 Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirements 4.17-2 Radioactive Ga.=eous Effluent Monitoring instrumentation l
Surveillance Lequirements 4.17-3 Radioactive Li nid Waste Sampling and Analysis Program 4.17-4 Radioactive Gaseous Waste Sampling and Analysis Program 5.5-1 Anticipated Annual Release of Radioactive Material in Liquid Effluents From Prairie Island Nuclear Generating Plant (Per Unit) 5.5-2 Anticipated Annual Release of Radioactive Nuclides in Gaseous Effluent From Prairie Island Nuclear Generating Plant (Per Unit) 6.1-1 Minimum Shift Crew Composition Prairie Island Unit 1 Amendment No. $$.1Cr/
Prairic Island Unit 2 Amendment No. $1, g l
TS.1-2 CONTAINMENT INTEGRITY CONTAINMENT INTEGRITY shall exist when:
1.
Penetrations required to be isolated during accident conditions are either:
Capable of being closed by an OPERABLE containment automatic isolation a.
valve system, or b.
Closed by manual valves, blind flanges, or deactivated automatic valves secured in their closed positions, except ss provided in Specifications 3.6.C and 3.6.D.
2.
The equipment hatch is closed and sealed.
3.
Each air lock is in compliance with the requirements of Specification 3.6.M.
4.
The containment leakage rates are within their required limits.
COLD SHUTDOVN A reactor is in the COLD SHUTD0k'N condition when the reactor is suberitical by at least 1% Ak/k and the reactor coolant average temperature is less than 200*F.
CORE ALTERATION CORE ALTERATION is the movement or manipulation of any component within the reactor pressure vessel with the vessel head removed and fuel in the vessel, which may affect core reactivity.
Suspension of CORE ALTERATION shall not preclude completion of movement of a component to a safe conservative position.
COPE OPERATING LIMITS REPORT The CORE OPERATING LIMITS REPORT is the unit-specific document that provides core operating limits for the current operating reload cycle.
These cycle-specific coce operating limits shall be determined for each reload cycle 3
in accordance with Specification 6.7.A.6.
Plant operation within these operating limits is addressed in individual specifications.
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Prairie Island Unit 1 Amendment No. 92,107 Prairie Island Unit 2-Amendment No. 85,100 l
a TS.3.4-2 3.4.B.1.d.
A minimum of 100,000 gallons of water is available in the condensate storage tanks and a backup supply of river water is availabic through the cooling water system, e.
Motor operated valves MV-32242 and MV-32243 (Unit 2 valves MV-32248 and MV-32249) shall have valve position monitor lights OPERABLE and shall be locked in the open position by having the motor control center supply breakers physically locked in the off position.
f.
Manual valves in the above systems that could (if one is improperly positioned) reduce flow below that assumed for accident analysis shall be locked in the proper position for emergency use.
During POWER OPERATION, changes in valve position will be under direct administrative control.
g.
The condensate supply cross connect valve C-41-2, to the auxiliary l
feedwater pumps shall be blocked and tagged open. Any changes in position of this valve shall be under direct administrative l
control.
2.
During STARTUP OPERATION or POWER OPERATION, any one of the following conditions of inoperability may exist for each unit provided STARTUP OPERATION is discontinued until OPERABILITY is restored.
If OPERABILITY is not restored within the time specified, place the affected unit (or either unit in the case of a motor driven AFW pump inoperability) in at least HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and reduce reactor coolant system average temperature below 350* F within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
a.
A turbine driven AFW pump, system valves and piping may be inoperable for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, b.
A motor driven AFW pump, system valves and piping may be inoperable for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, c.
The condensate storage tanks may be inoperable for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> provided the cooling water system is available as a backup supply of water to the auxiliary feedwater pumps.
d.
The backup supply of river water provided by the cooling water system may be inoperable for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> provided a minimum of 100,000 gallons of water is available in the condensate storage tanks.
e.
The valve position monitor lights for motor operated valves MV-32242 and MV-32243 (Unit 2 valves MV-32248 and MV-32249) may be inoperable for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> provided the associated valves' positions are verified to be open once each shift.
Prairie Island Unit 1 Amendment No. $1,107 Prairie Island Unit 2 Amendment No. $f,100
TS.3.6-1 3.6 CONTAINMENT SYSTEM Applicability Applies to the integrity of the containment system.
Obiective To define the operating status of the containment system for plant operation.
Specification A.
Containment Intecrity 1.
A reactor shall not be made or maintained critical nor shall reactor coolant system average temperature exceed 200* F unless CONTAINMENT INTEGRITY is maintained.
2.
If these conditions cannot be satisfied, within one hour initiate the action necessary to place the unit in HOT SHUTDOWN, and be in at least HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
B.
Vacuum Breaker System 1.
Both valves in each of two vacuum breaker systems, including actuating and power circuits, shall be OPERABLE when CONTAINMENT INTEGRITY is required (except as specified in 3.6.B.2 and 3.6.B.3 below).
2.
With one vacuum breaker inoperable with respect to its containment isolation function, apply the requirements of Specification 3.6.C.3, to the isolation valves associated with the inoperable vacuum breaker.
3.
One vacuum breaker may be inoperable with respect to its vacuum relief function for 7 days.
C.
Containment Isolation Valves 1.
Non-automatic containment isolation valves shall be locked closed or shall be under direct administrative control and capable of being closed within one minute following an accident when CONTAINMENT INTEGRITY is required (except as specified in 3.6.C.3 below).
2.
Automatic containment isolation valves, including actuation circuits, shall be OPERABLE when CONTAINMENT INTEGRITY is required (except as specified in 3.6.C.3 below).
3.
With one or more of the containment isolation valve (s) inoperable, within four hours:
l (a) restore the inoperable valv-is) to operable status or, i
(b) deactivate the operable valve in the closed position or, 1
(c) lock closed at least one valve in each penetration having one i
inoperable valve.
l Prairie Island Unit 1 Amendment No. 91,107 Prairie Island Unit 2 Amendment No. $5,100
i TS.4.4-2 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 i
accordance with the provisions of Appendix J,Section III.B and l
Section III.C, and Specification 4.4.A.5.
The airlocks shall be tested initially and at six-month intervals at 46 psig by pressurizing the inner volume.
In addition, when CollTA110GNT IltTEGRITY is l.
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required, each airlock shall be tested every 3 days if it is in use by i
pressurizing the intergasket space to 10 peig.
3.
Type A tests will be considered to be satisfactory'if the acceptance criteria delineated in Appendix J,.Section III.A are met.
i 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 60s of the 1, and if the following conditions--
l 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 e
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 it of the L, at 10 psig for door intergasket tests and 5% of the 1, at 46 psig for overall airlock tests.
t 5.
The retest schedules for Type A, B, and C tests will be in accordance l
with Section III.D of Appendix J.
Each shield building shall be-l retested in accordance with the Type A test schedule.for its containment. The raxiliary building special ventilation zone shall be i
retested in accordance with the Type A test schedule for Unit 1 containment.
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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 r
l ventilation zone shall have the same inspection and reporting '
l requirements as for the Type A tests of Unit.1.
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Prairie' Island Unit 1 amendment No. 62,107 Prairie Island Unit 2 Amendeent No. 56,100 i
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I B.3.6-1 t
3.6 CONTAINMENT SYSTEM
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i The OPERABILITY of the containment isolation valves ensures that the containment atmosphere will be isolated from the outside environment in j
the event of a release of radioactive material to the_ containment i
atmosphere or pressurization of the containment.
'l 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
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communication with the control room and capable of closing the affected valve (s) within one minute, (2) instructing this operator to_close these 4
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valves in an accident situation, and (3) assuring that environmental j
conditions will not preclude access'to close the valves and that this action will' prevent the release of radioactivity outside the containment.
l Proper functioning of the Shield Building vent system is essential to the performance of the containment system. '1herefore, except for reasonable j
periods of maintenance outage for one redundant chain of equipment, the
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i system should be wholly in readiness whenever above 200*F.
Proper l
functioning of the auxiliary building special vent system and isolation of i
the auxiliary building normal vent system are similarly necessary to i
preclude possible unfiltered leakage through penetrations that enter the special ventilation zone.
l The auxiliary building special ventilation zone and its associated l
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 j
normal ventilation systems so that AUXILIARY BUILDING SPECIAL VENTI 1ATION ZONE INTEGRITY can be intact during reactor operation. -The zone can I
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 4
Building Special Ventilation Zone is subject to the Technical Specifications of the Shield Building Integrity and not those associated with Auxiliary Building Special Ventilation Zone Integrity, j
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.
The COLD SHUTDOWN. condition precludes any energy release or buildup of containment pressure from flashing of reactor coolant in the event of_a system break.
The shutdown margin for the Co1D SHUTDOWN condition assures sub-criticality with the vessel closed, even if the most reactive rod control cluster. assembly were inadvertently withdrawn.
Prairie Island Unit 1 Amendment No. H,107 Prairie Island Unit 2 Amendment ho. H,100 -
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B.3.6-2 3.6 CONTAINMENT SYSTEM IAlig 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 above NDTT during power operation when containment is required.
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 vesse1~ 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 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 t
changing allowable containment leak rate from 0.54 to 0.25t/ day would offset the increased leakage (Reference 3).
High efficiency particulate absolute (HEPA) filters are installed before i
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 radiciodine 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 j
hydrogen concentration within containment below its flammable limit during post-LOCA conditions. -Either recombiner unit is capable of controlling the expected hydrogen generation associated with (1) zirconium water reactions, (2) radiolytic decomposition of water, and (3) corrosion of metals within containment. These hydrogen control systems are consistent with the recommendations of Regulatory Guide 1.7, " Control of Combustible Gas concentrations in Containment Following a IDCA", March 1971.
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 interlock, power operation may continue
-l provided that at_least one OPERABLE air lock door is closed. With an air lock door inoperable, access through the closed or locked OPERABLE door is only permitted.for repair of inoperable air lock equipment.
Prairie Island Unit 1 Amendment No. 91,107 Prairie Island Unit 2 Amendment No. $#,100 1
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B.3.6-3 3.6 CONTAINMENT SYSTEM lateE continued OPERABILITY of air locks is required to ensure that CONTAINMENT INTEGRITY is l
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.
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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Amendment No. H, y Prairie Island Unit 1 Amendment No. M,,w Prairie Island Unit 2
i B.4.4-1 I
2 4.4 CONTAINMENT SYSTEM TESTS f
i The. Containment System consists of a steel containment. vessel, a con-crete shield building, the Auxiliary Building Special Ventilation Zone
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(ABSVZ), a Shield Building Ventilation System. and an Auxiliary j
Building Special Ventilation System.
In the event of a. loss-of-coolant accident, a vacuum in the shield building annulus will cause most leakage from the containment vessel to be mixed in the annulus volume and j
recirculated through a filter system before its deferred release to the d
environment through the exhaust fan that maintains vacuum. Some of the-1 leakage goes to the ABSVZ from which it is exhausted through a filter. A small fraction bypasses both filter systems.
i The freestanding containment vessel is designed to accommodate the j
maximum internal pressure that would result from the Design Basis Acci-j dent (Reference 1).
For initial conditions typical of normal operation, 120*F and 15 psia, an instantaneous double-ended break with minimum j
safeguards results in a peak pressure of less than 46 psig at 268'F.
a The containment will be strength-tested at'51.8 psig and leak-tested at 46.0 psig to meet acceptance specifications.
j License Amendment Nos. 62 and 56 dated February 23, 1983 revised the j
Prairie Island Technical Specifications to conform.to the requirements of i
Appendix J to 10 CFR Part 50.
That License Amendment approved several clarifications and exemptions to the Type B and C testing requirements of j
Appendix J to 10 CFR Part 50.
Those clarifications and exemptions were incorporated into the Prairie Island Technical Specifications in the form l
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 3
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Generic Letter 91-08, " Removal of Component Lists From Technical i
Specifications". While the reference-of these notes to specific containment penetrations was relocated out of the Technical Specifications l-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 j
penetrations, is maintained in the Prairie Island Updated Safety Analysis j.
Report.
1 The safety analysis (References 2, 3) is based on a' conservatively chosen reference. set.of assumptions regarding the sequence of events j
relating to activity release and attainment and maintenance of vacuum in the shield building annulus and the Auxiliary Building Special
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Ventilation' Zone, the effectiveness of filtering, and the leak rate of the containment vessel as a function of time. The effects of variation in 4
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 demonstrates the collective effectiveness of conservatism in these f
assumptions.
4 Prairie Island Unit 1-Amendment. No. H,107 j
Prairie Island Unit 2 Amendment No. M,100 4
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..,o B. 4. 4. 't 4.4 CONTAINMENT SYSTEM TESTS An n 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.0254 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 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 0.54 per day at the peak accident pressure. Another conservative assumption in the calculation is.that primary containment leakage directly to the ABSVZ is 0.14 per day and leakage directly to the environs is 0.014 per day.
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 doses for these higher inleakage rates ~and found that for a primary containment leak rate of 0.254 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.
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, 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-i relative to those_ calculated for leaksgs directly from the containment in the design basis accident.
Prairie Island Unit'1
- Amendment No. H,107 Prairie Island Unit 2 Amendmen t No. Q, 100