ML20099A363
| ML20099A363 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 07/09/1992 |
| From: | Marsh L Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20099A367 | List: |
| References | |
| NUDOCS 9207290126 | |
| Download: ML20099A363 (32) | |
Text
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UNITED STATES 5?
.e NUCLEAR REGULATORY COMMISSION 8
WASHINGTON. O C. 20555
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NORTHERN STATES POWER COMPANY DOCKET NO. 50-282 PRAIRIE ISLAND NUCLEAR GENERATING PLANT. UNIT NO. 1 AMENDMENT TO FACILITY OPERATING llCENSE Amendment No. 99 License No. DPR-42 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Northern States Power Company (the licensee) dated October 4, 1991, 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 facili'y will operate in conformity with the application, the provisions of the Act, and the rules and reguletions of the Commission; 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 Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. OPR-42 is hereby amended to read as follows:
9207290126 920709 ADOCK0500g2 DR
. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.99, are hereby incorporated in the li;ense.
The licensee shall operate the facility in accordance with the Technical Specifications.
3.
'his license amendment is effective as of its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION L. B. Marsh, Director Project Directorate III-I Division of Reactor Projects III/IV/'!
Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: July 9,1992 i
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ATTACHffLNT TO LICENSE AMENDMENT NO. oo FACILITY OPERATING LICENSE NO. DPR-42 DOCKET NO. 50-282 Revise-Appendix A Technical-Specifications by removing the pages identified 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-iii-TS-iii TS-vi TS-vi TS-vii 15-vii TS-xi TS-xi TS.3.8-3 TS.3.8-3 Table TS.4.1-28 (page 1 of 2)
Table TS.4.1-28 (page 1 of 2)
Table TS.4.1-2B (page 2 of 2)
Table TS.4.1-2B (page 2 of 2)
TS.4.19-1 T:,5. 6-1 1S.5.6-1
-TS.5.6-2 TS.S.6-2 TS.5.6-3
.B.3-8-1 B.3.8-1 B.3.8-2 B.3.8-2 B.3.8-3 B.4.19-1
l TS-iii 1
TABLE OF CONTENTS (Continued)
TS'SECTION TITLE PACE 3 '. 6 Containment System TS.3.6 1 A. Containment Integrity TS.3.6-1 B. Vacuum Breaker System TS.3.6-1 C. Containment Isolation Valves TS.3.6-1 D. Containment Purge System TS.3.6 2 E. Auxiliary Building Special Ventilation Zone Integrity.
TS.3.6 2 F.-Auxiliary Building Special Ventilation System TS.3.6-3 G. Shield Building Integrity TS.3.6-3 H. Shield. Building Ventilation System
.TS.3.6 3 I. Containment Internal Pressure TS.3.6-3 J. Containment and Shield Building Air Temperature TS.3.6-4 K. Containment Shell Temperature TS.3.6-4 L. Electric Hydrogen Recombiners TS.3.6-4 M. Containment Air Locks TS.3.6-4 3.7 Auxiliary Electrical System TS.3.7 1 3.8-Refueling and-Fuel Handling TS.3.8-1
- A. Core Alterations TS.3.8-1 B, Fuel Handling Operations TS.3.8-3 l
C. Small Spent Fuel Pool Restrictions TS.3.8-4 D.' Spent Fuel Pool Special Ventilation System TS.3,8-4 E. Storage of Low Burnup Fuel TS.3.8-4 3.9 Radioactive Effluents TS.3.9-1 A. Liquid Effluents TS.3.9-1
- 1. Concentration TS.3.9-1
- 2. Dose TS.3.9-1
-3. Liquid Radwaste System TS.3.9-2
- 4. Liquid Storage Tanks TS.3.9-2 B. Gaseous Effluents TS.3.9-3
-1. Dose Rate-TS.3.9-3
~ Particulate TS.3.9-4
- 4. Gaseous Radwaste: Treatrent System and Ventilation Exhaust Treatment Systems TS.3.9-4
- 5. Containment Purging TS.3.9-5 C. Solid Radioactive Waste TS.3.9-6 D. Dose from All Uranium Fuel Cycle Sources TS.3.9 E. Radioactive Liquid Effluent Moni: cing-Instrumentation TS.3.9-7
. F. - Radioactive Gaseous Effluent Monitoring Instrumentation TS.3.9-7
- Prairie Island Unit 1 - Amendment No. //,7f,7p,$J,99 Prairie Island. Unit 2 - Amendment No. ##,f7,7),Ef,92 4
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TS vi h
TABLE OF CONTENTS (Continued)
TS SECTION TITLE PAGE 4.12 Steam Generator Tube Surveillance TS.4.12 1 A. C7eam Generstor Sample Selection and TS.4.12-1 Inspection B. Steam Generator Tube Sample Selection TS.4.12-1 and Inspection C. Inspection Frequencies TS.4.1.-3 D. Acceptance Criteria TS.4.12 4 E. Reports TS.4.12-5 4.13. Snubbers TS.4.13-1 4.14 Control Room Air Treatment System Tests TS.4.14-1 4.15 Spent Fuel Pool Special Ventilation System TS.4.15-1
- 4.16 Fire Detection and Protection Systems TS.4.16-1 A. Fire Detection Instrumentation TS.4.16-1 B. Fire Suppression Water System TS.4.16-1 C. Spray and Sprinkler Systems TS.4,16-3 D. Carbon Dioxide System TS.4.16-3 E. Fire Hose Stations TS.4.16-3 F.. Fire Hydrant Hose Heuses TS.4.16-4 G. Penetration Fire Barriers TS.4.16-4 4.17-Radioactive Effluents Surveillance TS.4.17-1 A. Liquid Effluents TS.4.17-1 B. Gaseous Effluents
.TS.4.17-2 C. Solid Rad.ioactive Waste TS.4.17-4 D. Dose from All Uranium Fuel-Cycle Sources TS.4.17-4 4.18 Reactor Coolant Vent System Paths TS.4.18-1 A, Vent Path Operability TS.4.18-1 B._ System Flow Testing TS.4.18-1 l
4.19 Auxiliary Building Crane Lifting Devices TS.4.19.
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Prairle Island Unit 1 - Amendment No. 32,73,91,99 P airie Island Unit 2 -- Amendment No. 2E,ES,84,92 L
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TS vii IABLE OF CONTENTS (Continued.1 TS SECTION TITLE PAGE 5.0 DESIGN FEATURES TS.S.1-1 5.1 Site TS.S.1-1 5.2 A. Containment Structures TS.5.2 1
- 1. Containment Vessel TS.S.2-1
- 2. Shield Building TS.S.2-2
- 3. Auxiliary Building Special Ventilation Zone B. Special Ventilation Systems TS.S.2-2 C. Containment System Functional Design TS.5.2-3 5.3 Reactor TS.S.3 1 A. Reactor Core TS.5.3 1 B. Reacter Coolant System TS.5.3-1 C. Protection Systems TS.S.3.1 5.4 Engineered Safety Features TS.S.4-1 5.5 Radioactire Waste Systems TS.5.5-1 A. Accidental Releases TS.S.5-1 B. Routine Releases TS.S.5-1
- 1. Liquid Wastes TS.S.5-1
- 2. Gaseous Wastes TS.5.5-2
- 3. Solid Wastes TS.5.5-3 C. Process and Effluent Radiological Monitoring TS.5.5-3 System 5.6 Fuel Handling TS.S.6-1 A. Criticality Consideration TS.S.6-1 B. Spent Fuel Storage Structure TS.5.6-1 C. Fuc1 Handling TS.5.6-2 l
D. Spent Fuel Storage Capacity TS.S.6-2 Prairie Island Unit 1 - Amendment No. 73,80,31,99 Prairie Island Unit 2 - Amendment No M,73,M,92 l
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a b TS-xi TABLE OF CONTENTS (continued)
TS BASES SECTl.0H TITIE PAGE 0
4.0 BASES FOR SURVEILLANCE REQUIREMENTS 4.1 Operational Safety Review B.4.1-1 4.2 Inservice Inspection and Testing of Pumps B.4.2-1 a.nd Valves Requirements 4.3 PrLmary Coolant Sy:. tem Pressure Isolation B.4.3-1 Ve.lvo 4.4 Contairment System Tests B.4.4-1 4.5 Engineeted Safety Features B.4.5-1 4.6 Periodic Testing of Emergene'l Power Systems B.A.C-1 4.7 Main Steam Isolation Valves B.4.7-1 4.8 Steam and Power Conversion Systems B.4.8-1 4.9 Reactivity Anomalies B.4.9 1 4.10 Radiation Environmental Monitoring Program B.4.10-1 A. Sample Collectirn and Analysis B.4.10-1 B. Land Use Census B.4.10 1 o
C. Interlaboratory Comparison Program B.4.10-1 4.11 Radioactive Source Leakage Test B.4.ll-1 4.12 Steam Generator Tube Surveillance B.4.12-1 4.13 Snubters B.4.13-1 4.14 Control Room Air Treatment System Tests B.4.14-1 4.15 Spent Fuel Pool Special Ventilation System B.4.15-1 4.16 Fire Detection and Protection Systems B.4.16-1 4.17 Radioactive Effluents Surveillance B.4.17 1 4.18 Reactor Coolant Vent System Paths B.4.18-1 4.19 Auxiliary Building Crane Lifting Devices B.4.19-1 l
o Prairie Is'and Unit 1 - Amendment tio, H,99 Prairie Island Unit 2 - Amendment tio. M,92
j TS.3.8-3 i
I 3.8.B.
' Fuel Handlinc Operations 1.
During fuel handling operations the following conditions shall be satisfied:
Radiation levels in the spent fuel storage pool area shall be a.
monitored continuously during fuel handling operations.
b.
Prior to fuel handling operations, fuel-handling cranes shall be load testad for OPERABILITY of limit switches, interlocks and alarms.
c.
A minimum boron concentration of 1800 ppm shall be maintained in the spent fuel pool whenever a spent fuel cask containing fuel is located in the spent fuel pool.
2.
If any of the conditions in 3.8.B.1, above, cannot be met, suspend all fuel handling operations and initiate the actions necessary to re-establish comp?iance with the requirements of 3.8.B.1.
Prairie' Island Unit 1 - Amendment No, J7,2E,73,7A,90,$1,99 Prairie Island Unit 2 - Amendment No. 71,79,EE.97,$3,E7,92
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4 Table TS.4.1 2B (Page 1 of 2)
TABLE TS.4.1-2B MINIMUM FREOUENCIES FOR SAMPLING TESTS FSAR Section TEST FREQUENCY Reference 1.
RCS Gross 5/ week Activity Determination
.2. :RCS Isotopic Analysis for DOSE 1/14 days (when at power)
. EQUIVALENT I-131-Concentration 3.
RCS Radiochemistry E determination 1/6 months (1) (when at power) 4.
RCS Isotopic Analysi-for Iodine a) Once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, whenever Including I-131,-I-133, and I 135 the specific activity ex-ceeds 1.0 uCi/ gram DOSE _
EQUIVALENT I-131 or 100/E uCi/ gram (at or above cold shutdown), and b) One-sample between 2 and 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> following THERMAL POWER change exceeding 15 percent of the RATED THERMAL POWER within a one hour period ( above hot shutdown) 5.
RCS' Radiochemistry (2)
Monthly-
-6. -RCS Tritium Activity' Weekly 7.
RCS Chemistry (Cl*,F*, 0 )
5/ Week 2
B, RCS Boron Concentration *(3) 2/ Week (4) 9.2 9.
RWST Boron Concentratien Weekly
-10. Boric Acid Tanks Boron Concentration 2/ Week 11-Caustic Standpipe NaOH Concentration Monthly 6.4 12,1 Accumulator Boron concentration Monthly 6
'13.-Spent Fuel. Pit Boron Concentration Monthly (7) 9.5.5 Prairie Island. Unit 1 - Amendment No. 25,E2,99
- Prairie Island Unit 2 - Amendment No. JS,AE,92
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Table TS,4.1-2B (Page 2 of 2)
TABLE TS.4,1-2B MINIMUM FREOUENCIES FOR SAMPLING TESTS FSAR Section TEST FREOUENCY Reference 14 Secondary Coolant Cross
-Weekly
- Beta-Gamma activity 15.
Secondary Coolant-Isotopic 1/6 months (5)
Analysis for DOSE EQUIVALENT 1-131 concentration 16, Seconcary Coolant Chemistry pH 5/ week (6) pHl Control Additive 5/ week (6)
Sodium 5/ week (6)
' Notes:
- 1. -Sample to be taken after a minimum of 2 EFPD and 20 days of POWER
. 0PERATION have elapsed since. reactor was last suberitical for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> 1or longer.
2.
To determine activity of corrosion products having a half-life greater than 30 mf.nutes.
3.
D- 'ing REFUELING, the boron concentration shall be verified by chemical
. alysis daily.
4.
The maximum interval between analyses shall not exceed 5 days.
5.
If activity of the samples is greater than 10% of the limit in Specification 3.4.D, the frequency shall be once per month.
6.
The maximum interval between analyses shall not exceed 3 = days.
7.
The minimum spent fuel pool boron concentration from Specification 3.8.B.1.b shall be. verified by chemical analysis weekly while a spent fuel cask'containing fuelLis located in the spent fuel pool.
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See Specification 4.1.D Prairie Island Unit 1 - Amendment No. 25,EJ,EZ,91,99 Frairie Island Unit 2 - Amendment No. J),fE,fE.EA,92
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TS.4 19-1
_ 4.19 Auxiliary Building Crane Lifting Devices-s Aeolicability Applies to surveillance requirements for the auxiliary building crane special lifting devices and slings before handling heavy loads carried over safe shutdown equipment or spent fuel in the spent fuel pool.
Obiective To verify that special lifting devices and slings used in conjunction with the auxiliary building crane are operable prior to their use in supporting heavy loads over safe shutdown equipment or spent fuel in the spent fuel pool.
Specification
-Slings and special lifting devices which will be used in supporting heavy loads 1from the auxiliary building crane shall be visually inspected and verified OPERABLE within 7 days prior to their use in handling heavy loads over_ safe shutdown equipment or spent fuel in the spent fuel pool.
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- 1 TS.5.6 1 5.6 FUEL HANDLING A.
.C_riticality consideration The;new and spene fuel pit structures are desigt.ed to withstand the anticipated earthquake loadings as Class I (seismic) structures.
The spent fuel pit t.as a stairless steel liner to ensure against loss of water (Reference 1).
The new and spent fuel storage racks are designed so that it is impossible to insert assemblies in other than the prescribed locations.
The design of~the new fuel storage pit and racks (Reference 1) ensures a new fuel pit K,g of less than or equal to 0.95, including uncertainties, even if unborated water were used to fill the pit.
The new fuel rack configuration also ensures K,g less than or equal to 0.98, including uncertainties, even if the new fuel racks were acciden* ally filled with a low density moderator which resulted in optimum low density moderation conditions.
Fuel stored in the new fuel storage racks will have-a maximum enrichment of 4.25 weight percent U-235.
The spent fuel storage rack design (Reference 1) and the limitations on the storage of low burnup fuel contained in Technical-Specification Section 3.8.E ensure a spent fuel pool K,g of less than or equal to 0.95, including uncertainties.
The maximum enrichment of fuel to be stored in the spent fuel pool will be 4.25 weight percent U-235.
Fuel will not be inserted into a spent fuel cask in the pool, unless a minimum boron' concentration of 1800 ppm is present.
The 1800 ppm will ensure that k,g for the spent fuel cask, including statistical uncertainties, will be less than or equal to 0.95 for all postulated arrangements of fuel within the cask.
The criticality analysis for the TN-40 spent fuel storage cask was based on fresh fuel enriched to 3.85
- weight percent U-235'.
B.
Spent Fuel Storage Structure The spent fuel storage pool is enclosed with a reinforced concrete building having 12-to 18-inch thick walls and roof (Reference 1).
The pool and pool enclosure are Class I (seismic) structures that afford protection against loss of integrity from postulated tornado missiles. The storage compartments and the fuel transfer canal are connected by fuel transfer slots that can be closed off with pneumatically sealed gates. The bottoms of-the slots are above the tops of the active fuel in the fuel assemblies which will be stored vertically in specially constructed racks.
The spent fuel pool has a reinforced concrete bottom slab nearly 6 feet thick and has been designed to minimize loss of water due to a dropped cask accident. Piping to the pool is arranged so that failure of any pipe cannot drain the pool below the tops of the stored fuel-assemblies.
Prairie Island' Unit 1 - Amendment No. 77,22,AE,7f,ED,$$,99 Prairie Island Unit 2 - Amendment No. JJ,16,42,E7.73,$3,92
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TS.S.6-2
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C.
' Fuel Handling The. fuel handling system provides the means of transporting and handling fuel from the time it reaches the plant in an unirradiated condition until it leaves after post-irradiation cooling.
The system consists of the refueling cavity, the fuel f$ansfer system, the spent fuel storage pit, and the spent fuel cask transfer system.
Major components of the fuel handling system are the manipulation crane, the spent fuel pool bridge, the auxiliary building crane, the fuel transfer system, the spent fuel storage racks, the spent fuel cask, and the rod cluster control changing fixture. The reactor vessel stud tensioner, the reactor vessel head lifting device, and the reactor internals lifting device are used for preparing the reactor for refueling and for assembling the reactor after refueling.
Upon arrival in the storage pit, spent fuel will be removed from the transfer system and placed, one assembly at a time, ', storage racks using a long handled manual tool suspended from the spent fuel pit bridge crane. After sufficient decay, the fuel will be loaded into storage casks for storage in the Independent Spent Fuel Storage Installation or into shipping casks for removal from the site.
The casks will be handled by the auxiliary building crane.
Spent fuel casks will be handled by a single failure proof handling system l
meeting the requirements of Section 5.1.6 of NUREG-0612. " Control of Heavy Loads at Nuclear Power Plants", July 1980, The auxiliary building crane has been upgraded to conform with the single failure proof requirements of Section 5.1.6 of NUREG-0612. The auxiliary building crane is designed to l_
not allow a load drop as a result of any single failure.
The improved reliability of the auxiliary building crane is achieved through increased
-factors of safety and through redundancy or duality in certain active
-components.
l D.
Spent Fuel Storage Caparily The spent fuel storage facility is a two-compartment pool that, if-completely filled with fuel storage racks, provides up to 1582 storage locations.
The southeast corner. of the small pool (pool no. 1) also servcs as-the cask lay-down area. During times when the cask is being used,'four racks are removed from the small pool. With the four storage racks in the southeast corner of pool 1 removed, a total of i
1386 storage locations are provided. To allow insertion of a spent fuel l
cask total storage is limited to 1386 assemblies, not including those assemblies which can be returned to the reactor.
l l-Reference 1.
USAR, Section 10.2 Prairie Island Unit 1 - Amendment No. /E,EI,77,DR,$$,99 Prairie Island Unit 2 - Amendment No. 42,EE,E7,73,E3,92
.sm u B.3.8-1 9
3.8 REFUELING AND FUEL HANDLING Bases The eqtipment and general procedures to be utilized durtng refueling are discussed in the FSAR. Detailed instructions, the precautions specified above, and the design of the fuel handling equipment incorporating built-in interlocks and safety features, provide assurance that no incident could occur during CORE ALTERATIONS that would result in a hazard to public health and safety (Reference 1).
Whenever changes aresnot being made in core geometry, one flux monitor is sufficient. This permits maintenance of the instrumenta-tion. Continuous monitoring of radiation levels and neutron flux provides immediate indication of an unsafe condition. The residual heat removal pump is used to maintain a uniform boron concentration.-
Under rodded and unrodded conditions, the L,,, of the reactor must be less than or equal to 0.95 and the boron concentration must be greater L
than or equal to 2000 ppm.
Periodic checks of refueling water boron concentration insure that proper shutdown margin is maintained.
3.8.A.l.h. allows the control room operator to inform the manipulator operator of any impending unsafe condition detected from the main control board indicators during fuel movement.
No movement of fuel in the reactor is permitted until the reactor has been suberitical for at least 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> to permit decay of the fission products in the fuel. The delay time is consistent with the l
fuel handling accident analysis (Reference 2).
Fuel will not be inserted into a spent fuel cask unless a minimum boron concentration of 1800 ppm is present. The 1800 ppm will ensure that k,,,
for the spent fuel cask, including statistical uncertainties, will be less than or equal to 0.95 for all postulated arrangements of fuel within the cask.
.The number of recently discharged assemblies in Pool No. 1 has been limited to 45 to provide assurance that in the event of loss of pool cooling capability, at least eight hours are available under worst case conditions to make repairs until the onset of boiling.
The Spent _ Fuel Fool Special Ventilation System (Reference 3) is a safeguards system which maintains a negative-pressure _in the spent fuel enclosure _upon detection of high area radiation. The Spent Fuel Pool Normal Ventilation System is automatically isolated and exhaust air is drawn through filter modules containing a roughing filter, particulate-filter, and a charcoal filter before discharge to the environment via one of the Shield Building exhaust stacks. Two completely redundant trains are provided. The exhaust fan and filter of each train are
-shared with the corresponding train of the Containment In-service Purge System. High efficiency particulate absolute (HEPA) filters are installed before che charcoal adsorbers to prevent clogging of the iodine adsorbers in each SFPSVS filter train.
The charcoal adsorbers are installed to L
reduce the potential release of radiciodine to the environment.
Prairie Island Unit 1 - Ar.endment No. 9J,99 Prairie Island Unit 2 - Amendment No. E/,92
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3.8 REFUELING AND ICEL llANDLING Pases continued During movement of irradiated fuel assemblies or control rods, a water level of 23 feet is maintained to provide sufficient shielding.
The water level may be lowered to the top of the RCCA drive shafts for latching and unlatching. The water level may also be luwered below 20 feet for upper internals removal / replacement. The basis for these allowance (s) are (1) the refueling cavity pool b a sufficient level to allow time to initiate repairs or emergency procedures to cool the core, (2) during latching / unlatching and upper internals removal / replace-ment the level is closely monitored because the activity uses this level as a reference point, (3) the time spent at this level is minimal.
L The requirements for the storage of low burnup fuel in the spent fuel pool er/ure that the spent fuel pool will remain suberitical durir.g fuel storage.
Fuel stored in the spent fuel pool will be limited to a maximum enrichment of 4.25 weight percent U 235.
It has been shown by criticality analysis that the use of the three out of four storage configuration will assure that the K,,, will remain less than 0.95, including uncertainties, when fuel with a maximum enrichment of 4.25 weight percent U 235 and average assembly burnup of less than 5,000 E D/MTU is stored in the spent fuel pool.
The requirement for maintaining the spent fuel pool boron concentratior greater than 500 ppm whenever fuel with average assembly burr.up of lest (nan 5,000 WD/MTU is stored in the spent fuel pool ensures that K,,, for C.e spent fuel pool will remain less than 0.95, including uncertainties, even if a fuel assembly is inadvertently inserted in the empty cell of the thrae out of four storage configuration Re fereneca 1.
USAR, Section 10.2.1.2 2.
USAR, Section 14.5.1 3.
USAR, Section 10.3.7 Prairie Island Unit 1 - Amendment No. N,99 Prairie Island Unit 2 Amendment No. M,92
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e B.4.19 1 4.19 Auxiliary Buildinr Crane 1.1f tine Devlees Bases The auxiliary builcing crane has been modified to conform with the single failure proof requirements of Section 5.1.6 of NUREC 0612, acontrol of Heavy-Loads at Nuclea< Power Plants", July 1980. The auxiliary building crane is designed to not allow a load drop as a result of any single failure. As the slings and special lifting devices are, by thefr nature, an integral part of the load bearing path, their surveillance is necessary to ensure agt wt a load drop as a result of deficient rigging. Any load that weighs tro.. than the combined weight of a singic fuel assembly and its associated handling tool is considered a heavy load.
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i Prairie Island Unit 1 - Amendment tio. 99 Prairie Icland Unit 2 - Amendment lio. 92
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UNITED STATES n
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NUCLEAR REGULATORY COMMISSION 8
WASHINGTON. D.C. 20Z6 o[s'
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NORTHERN STATESJ0WER COMPANY DOCKET NO. 50-306 g
PRAIRIL11 LAND NUCLEAR GEt!ERATING ELANT. UNii NO. 2 AMENDMENT TO FACILITY OPERATING LICEl{1[
Amendment No. 92 License No. DPR-60 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Northern States Power Company (the licensee) dated
, 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 (i) that the activities authorized by this amendment can be conducted without endangering the health
- ind 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 safet-7f 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 sat'-fied.
2.
According y, the license is amended by changes to the Technical Specifice ions as indicated in the attachment to this license amendment, and paragtaph 2.C.(2) of facility Operating License No. CPR-60 is hereby amended tc read as follows:
. Technical Specifications The Technical Specifications contained to Appendix A, as revised through An.endment No.92, 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 its date of issuance.
FOR THE NUCLEAR REGULA10RY COMMISSION
[
L. B. Marsh, Director Project Directorate 111-1 Division of Reactor Projects Ill/IV/V Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
July 9,1992
r.
ATTACHMENT TO LLCENSE AMENDMENT NO. o?
FACILITY OPERATING LICENSE NO. OpR-60 DOCKET NO. 50-306 Revise Appendix A Technical Specifications by removing the pages identified 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-tii TS-iii TS-vi TS-vi TS-vii TS-vii TS-xi TS-xi TS.3.8-3 T5.3.8-3 Table TS.4.1-2B (page 1 cf 2)
Table TS.4.1-28 (page 1 of 2)
Table T3.4.1-2B (paga 2 of 2)
Table TS.4.1-28 (page 2 of 2)
TS.4.19-1 TS.5.6-1 TS.5.6-1 TS.5.6-2 TS.S.6-2 TS.S.6-3 B.3.8-1 B.3.8-1 B.3.8-2 B.3.8-2 B.3.8-3 B.4.19-1
.,, - +
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7
T5 111 TABLE OF CONTI C fContinuedi U ECTION IlTJJ PACE 3.6 Containment System TS.3.6 1 A. Containment Integrity TS.3.6 1 B. Vacuum Breaker System TS.3.6 1 C Centaincent Isolation Valves TS.3.6 1 D. Containment Purge System 15.3.6 2 E. Auxiliary Building Special Ventilation Zone Integrity TS.3.6 2 F. Auxiliary Building Special Ventilation System TS.3.6 3 C. Shield Building Integrity TS.3.6 3 H. Shield Building Ventilation System TS.3.6-3 r
- 1. Containment Internal Pressure TS.3.6 3 J. Contairaent and Shield Building Air Temperature TS.3.6 4 K. Containment Shell Temperature TS 3.6 4 L. Electric Hydrogen Recombiners TS.3.6 4 M.
- ontainment Air Locks 75.3.6 4 3.7 Auxiliary Electrical $ stem TS.3.7 1 3
3.8 Refueling and Tuel Handling TS.3.8 1 A. Core Alterations TS.3.8 1 B. Tuel Handling Operations TS.3.8 3 l
C. Small Spent Fuel Pool Restrictions TS.3.8-4 D. Spent Tuel Fool Special Ventilation System TS.3.8 4 E. Storage of Low Burnup Fuel TS.3.8-4 3.9 Radioactive Effluents TS.3 9 1 A. Liquid Effluents TS.3.9 1
- 1. Concentration TS.3.9 1
- 2. Dese TS.3.9-1
- 3. Liquid Radwaste System TS.3.9 2
- 4. Liquid Storage Tanks TS.3.9 2 B. Caseous Effluents TS.3.9 3
- 1. Dose Rate 75.3.9 3
- 2. Dose from Noble Cases
- !S.3.9-3
- 3. Dose from I 131, Tritium and Radioactive Particulate TS.3.9 4 4 Caseous Radvasta Treatment System and Ventilation Exhaust Treatment Systems TS.3.9 4
- 5. Containment Purging TS.3.9 5 C. Solid Radioactive Waste TS.3.9 6 D. Dose from All Uranium Fuel Cycle Sources TS.3.9 6 E. Radioactive Liquid Effluent Monitoring Ins tr umenta tion 75.3.9-7
)
F. Radioactive Caseous Effluent Monitoring Instrumentation TS.3.9 7 Prairie Island Unit 1 - Amendment No. //,7A,7#,5),99 Prairie Island Unit 2 - Amendment No. Ef,f7.7).Ef 92
TS vi TABLE OF CONTENTS (Continued)
TS SECTION 11I11 PACE 4.12 Steam Generator Tube Surveillance T5.4.12-1 A. Steam Generator Sanple Selection and TS.4.12-1 Inspection B. Steam Generator Tube Sample Selection TS.4.12-1 and Inspection C. Inspection Frequencies TS.4.12 3 D. Acceptance Criteria TS.4.12 4 E. Reports TS.4.12 5 4.13 Snubbers TS.4.13 1 4.14 Control Room Air Treatment System Tests TS.4.14 1 4.15 Spent Tuel Fool Special Ventilation System TS.4.15 1 4.16 Fire Detection and Protection Systems TS.4.16 1 A. Fire Detection Instrumentation TS.4.16 1 B. Fire Suppression Water System TS.4 16 1 C. Spray and Sprinkler Systems TS.4.16-3 D. Carbon Dioxide System TS.4.16 3 E. Fire Hose Stations TS.4.16 3 T. Fire Hydrant Hose Houses TS.4.16-4 G. Penetration Fire Barriers TS.4.16 4 4.17 Radioactive Effluents Surveillance TS.4.17 1 A. Liquid Effluents TS.4.17 1 B. Gaseous Effluents TS.4.17-2 C. Solid Radioactive Vaste T5.4.17-4 D. Dose from All Uranium Tuel Cycle Sources 75.4.17 4 4.16 Reactor Coolant Vent System Paths TS.4.18 1 A. Vent Path Operability TS.4.18 1 B. System riow Testing TS.4.18-1 l
4.19 Auxiliary Building Crane Lifting Devices TS.4.19-1 Prairie Island Unit 1 - Amendment No. 22,73,92,99 Prairie Island Unit 2 - Amendment No. 26,ff,EA,92
TS vii IABLE OF CONTENTS (Continued)
TS SECTION TITLE PAGE__
5.0 DESIGN FEATURES TS.S.1 1 5.1 Site 7S.5.1-1 5.2 A. Containment Structures 75.5.2 1
- 1. Containment Vessel TS.S.2 1
- 2. Sh':1d Building TS.S.2 2
- 3..$'iciliary Building Special Ventilation Zone B. Speci: 1 Ventilation Systees TS.5.2 2 C. Containment Systes Functional Design TS.5.2 3 5.3 Reactor TS.5.3 1 A. Reactor Core TS.S.3 1 B. Reactor Coolant System TS.5.3 1 C. Protection Systems TS.S.3 1 5.4 Engineered Safety Features TS.5.4 1 5.5 Radioactive Vaste Systems T5.5.5 1 A. Accidental Releases TS.S.5 1 B. Routine Releases TS.5.5 1
- 1. Liquid Vastes TS.S.5-1
- 2. Gaseous Vastes TS.5.5-2
- 3. Solid Vastes TL.5.5 3 C. Process and Effluent Radiological Monitoring TS.S.5 3 System 5.6 fuel Handling TS.S.6-1 A. Criticality Consideration TS.5.6 1 B. Spent Fuel Storage Structure TS.S.6-1 C. Fuel Handling TS.5.6-2 D. Spent Fuel Storage Capacity TS.5.6 2 l
Prairie Island Unit 1 - Amendment No. 73,80,9J,99 Prairie Island Unit 2 - Amendment No. EE,73.BA,92
TS ni TABLE OF CONTENTS (continued)
TS. BASES SECTION TITLE PAGE.,_
4.0 BASES FOR SURVEILLANCE REQUIREMENTS 4.1 Operational Safety Review B.4.1-1 4.2 Inservice Inspection and Testing of Pumps B.4.2 1 and Valves Requirements 4.3 Primary Coolant System Pressure Isolation B.4.3-1 Valves 4.4 Containment System Tests B.4.4 1 4.5 Engineered Safety Teatures B.4.5 1 4.6 Periodic Testing of Emergency Power System.
B.4.6 1 4.7 Main Steam Isolation Valves B.4.7 1 4.8 Steam and Power Conversion Systems B.4.8-1 4.9 Reactivity Anomalies B.4.9 1 4.10 Radiation Environmental Monitoring Program B.4.10-1 A. Sample Collection and Analysis B.4.10 1 B. Land Use Census B.4.10 1 C. Interlaboratory Comparison Program B.4.10-1 4.11 Radioactive Source Leakage Test B.4.11-1 4.12 Steam Generator Tube Surveillance B.4.12 1 4.13 Snubbert, B.4.13 1 4.14 Control Room Air Treatment System Tests B.4.14-1 4.15 Spent Fuel Pool Special Ventilation System B.4.15 1 4.16 Fire Detection and Protection Systems B.4.16 1 4.17 Radioactive Effluents Surveillance B.4.17 1 4.18 Reactor Coolant Vent System Paths B.4.18-1 4.19 Auxiliary Euilding Crane Lifting Devices B.4.19 1 l
Prairie Island Unit 1 - Amendment No. 91,99 Prairie Island Unit 2 - Amendment No. EA,92
TS.3,8-3 3.8.B.
Fuel Handlinr. Operations 1.
During fuel handling operations the following conditions shall be satisfied Radiation levels in the spent fuel storage pool area shall be a.
monitored continuously during fuel handling operations.
b.
Prior to fuel handling operations, fuel. handling cranes shall be load tested for OPERABILITY of limit switches, interlocks and alarms.
A minimum horon concentration of 1800 ppm shall be maintained c.
in the spent fuel pool whenever a spent fuel cask containing fuel is located in the spent fuel pool.
2.
If any of the conditions in 3.8.B.1, above, cannot be met, suspend all fuel handling operations and initiate the actions necessary to re establish compliance with the requirements of 3.8.B.1.
I i
l t
L Prairie Island Unit 1 - Amendment No. 27,2#,73,7A,90,$1,99 l
Prairie Island Unit 2 - Amendment No. II.JP,El,E/,$3,E/,92
Table TS.4.1 2B (Page 1 of 2)
TABLE TS.4.1-28 i
MINIMUM FREOUENCIES FOR SAMPLING TESTS FSAR Section TEST FREOUENCY Reference 1.
RCS Gross 5/ week Activity Determination 2.
RCS Isotopic Analysis for DOSE 1/14 days (when at power)
EQUIVALENT I 131 Concentration 3.
RCS Radiochemistry E determination 1/6 months (1) (when at power) 4.
RCS Isotopic Analysis for Iodine a) Dnce per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, whenever Including I 131, I-133, and I-135 the specific activity ex-ceeds 1.0 uCi/ gram DOSE _
EQUIVALENT I 131 or 100/E uCi/ gram (at or above cold shutdown), and b) One sample between 1 and 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> following THEFyAL POWER change exceeding 15 percent of the RATED THERMAL POWER within a one hour period ( above hot shutdown) 5.
RCS Radiochemistry (2)
Monthly 6.
RCS Tritium Activity Weekly 7.
P.CS Chemistry (Cl*, P*, O )
5/ Week g
8.
RCS Boron Concentration *(3) 2/ Week (4) 9,2 9.
RWST Boron Concentratinn Weekly
- 10. Boric Acid Tanks Boron Concentration 2/ Week
- 11. Caustic Standpipe NaOH Concentrution Monthly 6.4
- 12. Accumulator Boron Concentration Monthly _
6 L
- 13. Spent Fuel Pit Boron Concentration Monthly (7) 9.5.5 L
Prairie Island Unit 1 - Amendment No. 2),52,99 Prairie Island Unit 2 - Amendment No. 29,AE,92 l
L 1
. _. ~. _... _ _,
Tchle 75.4.1 2B (Pass 2 of 2)
TABLE TS.4.1-2B MINIMtM FREOUENCIES FQR SAMPLING TESTS i
TSAR Section TEST FP.EOUENCY Referenec 14 Secondary Coolant Cross Weekly Beta Gamma activity 15.
Secondary Coolant Isotopic 1/6 uonths (5)
Analysis for DOSE EQUIVALENT I.131 concentration 16.
Secondary Coolant Chemistr/
pH S/ week (6) pH Control Additive 5/ week (6)
+
Sodium 5/ week (6)
Notes-1.
Sample to be taken af ter a minimum of 2 EFPD and 20 days of POWER OPERATION have elapsed since reactor was last suberitical for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or longer.
2.
To determine activity of corrosion products having a half life greater than 30 minutes.
3.
During REFUELING, the boron concentratien shall be verified by chemical analysis daily.
4.
The. maximum interval between analyses shall not exceed 5 days.
5.
If activity of the samples is greater than 10% of the limit in specification 3.4.D, the frequency shall be once per month.
l' 6.
The maximum interval between analyses shall not exceed 3 days.
7.
The minimum spent fuel pool boron concentration from Specification 3.8.B.1.b shall be verified by chemical analysis weekly while a spent fuel cask containing fuel is located in the spent fuel pool.
See Specification 4.1.D.
l' Prairie Island Unit 1 - Amendment No. 2E EJ,E2,SJ.99 l
Prairie Island Unit 2 - Amendment No.19,/E,/f,EA,92 l
l
TS.4.19 1 4.19 Au3111ary Buildine Crane liftinc Devices Annlicabilin Applies to surveillance requirements for the auxiliary building crane special lifting devices and slings before handling heavy loads :arried over safe shutdown equipment or spent fuel in the spent fuel pool.
Obieetive To verify that special lifting devices and slings used in conjunction with the auxiliary 'ouilding crane are operable prior to their use in support.ing heavy loads over safe shutdown equipment or spent fuel in the spent fuel pool.
Snecifiention Slings and special lifting devices which will be used in supporting heavy loads from the auxiliary building crant shall be visually inspected and verified OPERABLE within 7 days prior to their use in handlin5 heavy loads over safe shutdovn equipment or spent fuel in the spent fuel pool.
Prairie Island Unit 1 - Amendment No. 99 Prairie Island Unit 2 - Amendment No. 92
o 9
TS.5.6 1 5.6 RIEL HANDLING A.
Criticality Consideration The new and spent fuel pit structures are designed to withstand the anticipated earthquake loadings as Class I (seismic) structures. The spent fuel pit has a stainless steel liner to ensure against loss of water (Reference 1).
The new and spent fuel storage racks are designed so that it is impossible to insert assemblies in other than the prescribed locations.
The design of the new fuel storage pit and rtaks (Reference 1) ensures a nev fuel pit K,g of less than or equal to 0.95, including uncertainties, even if unboreted water were used to fill the pit.
The new fuel rack configuration also ensuren K less than or equal to 0.98, including g
uncertainties, even if the n,eu fuel racks were accidentally filled with a low dernity moderator which resulted in optimum low density moderation conditions.
Fuel stored in the new fuel storage racks will have a maximus enrichment of 4.25 weight percent U-235.
The spent fuel storage rack design (Reference 1) and the limitetions on the storage of low burnup fuel contained in Technical Specification Section 3.8.E ensure a spent fuel pool K,n of less than or equal to 0.95, including uncertainties.
The maximum enrichment of fuel to be stored in the spent fuel pool will be 4.25 weight percent U 235.
Fuel will not be inserted into a spent fuel cask in the pool, unless a minimum boron concentration of 1800 ppm is present. The 1800 ppm will ensure that k,g for the spent fuel cask, including statistical uncertainties, will be less than or equal to 0.95 for all postulated arrangements of fuel within the cask. The criticality analysis for the TN 40 spent fuel storage cask "as based on fresh fuel enriched to 3.85 weight percent U 235.
3.
Spent Fuel Storare Structure The spent fuel storage pool is enclosed with a reinforced concrete building having 12-to 18-inch thick walls and roof (Reference 1).
The pool and pool enclosure are Class I (seismic) structures thae afford protection against loss of integrity from postulated tornado missiles.
The storage compartments and the fuel transfer canal are connected by fuel transfer slots that can be closed off with pneumatically sealed gates. The bottoms of the slots are above the tops of the active fuci in the fuel assemblics which will be stored vertically dn specially constructed racks.
The spent fuel pool has a reinforced concrete bottom slab nearly 6 feet thick cr.d has been designed to minimize loss of water due to a dropped cask accident.
Piping to the pool is arranged so that failure of any pipe cannot drain the pool below the tops of che stored fuel assemblies.
Prairie Island Unit 1 - Amendmer,t No. 77,27,/E.7/,RD,70,99 Prairie Island Unit 2 - Amendnent Ho_ JJ.JE,#7,E7,7),E3,92
A TS.5.6 2 C.
Fuel Handline The fuel handling system provides the means of transporting and handling fuel from the time it reaches the plant in an unirradiated condition until it leaves after post irradiation cooling. The system consists of the refueling cavity, the fuel transfer system, the spent fuel storage pit, and the spent fuel cask transfer system.
Major components of the fuel handling system are the manipulation crane, the spent fuel pool bridge, the auxiliary buf1 ding crane, the fuel transfer system, the spent fuel storage racks, the spent fuel cask, and the rod cluster control changing fixture.
The reactor vessel stud tensionor, the reactor vessel head lifting device, and the reactor internals lifting device are used for preparing the reactor for refueling and for assembling the reactor after refueling.
Upon arrival in the storage pit, spent fuel vill be removed from the transfer system and placed, one ascembly at a time, in storage racks using a long-handled manual tool suspended from the spent fuel pit bridge crane. After sufficient decay, the fuel will be loaded into storage casks for storage in the Independent Spent Tuel Storage Installation or into shipping casks for removal from the site.
The casks will be handled by the auxiliary building crane.
Spent fuel casks will be handled by a single failure proof handlin6 system meeting the requirements of Section 5.1.6 of NUREG 0612 " Control of Heavy Loads at Nuclear Power Plants", July 1980.
The auxiliary building crane has been upgraded to conform with the single failure proof requirements of Section 5.1.6 of NUREG 0612. The auxiliary building crane is designed to not allow a loa 6 drop as a result of any single failure.
The improved reliability of the auxiliary building crane is achieved through increased factors of safety and through redundancy or duality in certain active components.
D.
Spent Fuel Storace Cacacity The spent fuel storage facility is a two-compsrtment pool that, if completely filled with fuel storage racks, provides up to 1582 storage locations. The southeast corner of the small pool (pool no. 1) also servec as the cask lay down area.
During times when the cask is being used, four racks are removed from the small pool. With the four storage racks in the southeast corner of pool I removed, a total of 1386 storage locatier.s are provided. To allow insertion of a spent fuel cask, total storage is limited to 1385 assemblies, not including those assemblies which can be returned to the reactor.
Reference 1.
USAR, Section 10.2 Prairie Island Unit 1 - Amendment No. /E,$1,7/,00,$E,99 Prairie Island Unit 2 - Amendment No. 62,EE E7,73,$3,92 I-
. - ~.
8.3.8 1
~
3.8 REFUELING AND TUE1. HANDL7NG Bases The eg'aipment and general procedures to be utilized during refueling are discussed in the FSAR. Detailed instructions, the precautions specified above, and the design of the fuel handling equipment incorporating built-in interlocks and safety features, provide assurance that no incident could occur during CORE ALTERATIONS that would result in a hazard to public health and safety (Reference 1).
i Vhenever changes are not being made in core geometry, one flux monitor is sufficient. This permits maintenance of the instrumento-tion.
Continuous monitoring of radiation levels and neutron flux provides immediate indication of an unsafe condition. The residual heat removal pump is used to maintain a uniform boron concentration.
Under rodded and unrodded conditions, the K,g of the reactor must be less than or equal to 0.95 and the boron concentration must be greater than or equal to 2000 ppm. feriodic checks of refueling water boron concentration insure that proper shutdown margin is maintained.
3.8.A.1.h allows the contrel room operator to inform the manipulator operator of any impending unsafe condition detected from the main control board indicators during fuel movement.
No movement of fuel in the reactor is permitted until the reactor has been suberitical for at least 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> to permit decay of the fission products in the fuel. The delay time is consistent with the fuel' handling accident analysis (Reference 2).
Fuel-will not be inserted into a spent fuel cask unless a minimum boron concentration of 1800 ppm is present.
The 1800 ppm vill ensure that k forthespentfuelcask,includingstatisticaluncertainties,willbelgess than or equal to 0.95 for all postulated arrangements of fuel within the cask.
The number of recently discharged assemblier in Pool No. I has been limited to 45 to provide assurance that in the event of loss of pool cooling capability, at least eight hours are available under worst case conditions to make repairs until the onset of boiling.
The Spent Fuel pool Special Ventilation System (Refierence 3) is a safeguards system which maintains a negative pre sv e in the spent fuel enclosure upon detection of high area radiation. the Spent Fuel Pool Normal Ventilation System is automatically isolated and exhaust air is drawn through filter modules containingla roughing filter, particulate filter, and a charcoal filter before discharge to the environment via one of.the Shield Building exhaust seseks. Two completely redundant trains'are provided. The exhaust fan and filter of each train are shared with the corresponding train of the Containment In service Purge System. High efficiency particulate absolute (HEPA) filters are installed before the charcoal adsorbers to. pn vent clogging of the iodine adsorbers in each SFPSVS filter train. The charcoal adsorbers are installed to
. reduce the potential release of ra.ioiodine to the environment.
Prairie. Island Unit 1 - Amendment No. SJ,99 Prairie Island Unit 2 - Amendment No. EA,92
B.3.8 2 3.8 REFUELING AND FUEL HANDLINC Bases continued During movement of irradiated fuel assemblies or control rods, a water level of 23 feet is naintained to provide sufficient shielding.
4he water level may be lowered to the top of the RCCA drive shafts for latching and unlatching.
The water level may also be lowered below 20 feet for upper internals removal / replacement.
The basis for these allowance (s) are (1) the refueling cavity poc1 has sufficient level to allow time to initiate repairs or emergency procedures to cool the core, (2) during latching / unlatching and upper internals removal /re lace-the level is closely monitored because the activity uses this ment level as a reference point. (3) the time spent at this level is minimal.
The requirements for the storage of low burnup fuel in the spent fuel pool ensure that the spent fuel pool will remain suberitical during fuel storage.
Fuel stored in the spent fuel peol will be limited to a maximum enrichment of 4.25 weight percent U 235.
It has been shovn by criticality analysis that the use of the three out of four storage configuration will assure that the K will remain less than 0.95, including uncertainties, when fuel with amaxigmum enrichment of 4.25 weight percent U 235 and average assembly burnup of less than 5,000 MVD/MTU is stored in the spent fuel pool.
The requirement for maintaining the spent fuel pool boron concentration greater than 500 ppm whenever fuel with average assembly burnup of less than 5,000 MVD/MTU is stored in the spent fuel pool ensures that K,g for the spent fuel pool will remain less than 0.95, including uncertainties, even if a fuel assembly is inadvertently inserted in the empty cell of the three out of four storage configuration.
l l.
l l
(
References L
1.
USAR, Section 10.2.1.2 i
2.
USAR, Section 14.5.1 3.
USAR, Section 10.3.7 Prairie Island Unit 1 - Amendaent No. 91,99 Prairie Island Unit 2 - Amendment No. EA,92
B.4.19 1 s
e.
4.19 6xiliarv Buildine Crane tiftinr Devices fases i
The auxiliary building crane has been nodified to conform with the single failure procf requiretents of Section 5.1.6 of NUREG 0612, " Control of Heavy loads at Nuclear Power Plants *, July 1980. The auxiliary building crane is de.tigned to not allow a load drop as a result of any single failure. As the slings and special lifting devicer are, by their nature, an integral part of the load bearing path, their s-.veillance is necessary to ensure against a load drop as a result of deficient rigging. Any load that weighs more than the combined weight of a single fuel assembly and its associated handling tool is considered a heavy load.
e Prairie Island Unit 1 - Amendment No. 99 Prairie Island Unit 2 - Amendment No. 92
-