ML17241A324
ML17241A324 | |
Person / Time | |
---|---|
Site: | Saint Lucie ![]() |
Issue date: | 05/06/1999 |
From: | Peterson S NRC (Affiliation Not Assigned) |
To: | |
Shared Package | |
ML17241A325 | List: |
References | |
NUDOCS 9905110202 | |
Download: ML17241A324 (19) | |
Text
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UNITED STATES NUCLEAR REGULATORY COMIVIISSION WASHINGTON, D.C. 205554001 FLORIDA POWER & LIGHTCOMPANY ORLANDO UTILITIESCOMMISSION OF THE CITYOF ORLANDO FLORIDA AND FLORIDA MUNICIPALPOWER AGENCY DOCKET NO. 50-389 ST. LUCIE PLANT UNIT NO. 2 AMENDMENTTO FACILITYOPERATING LICENSE Amendment No.
101 License No. NPF-16 The Nuclear Regulatory Commission (the Commission) has found that:
\\
A.
The application for amendment by Florida Power & Light Company (FPL), dated December 31, 1997, and supplemented May 15, 1998, September 15, 1998, November 25, 1998, and January 28, 1999, 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 facilitywilloperate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; 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 willbe 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.
9905ii0202 990S06 PDR ADOCK 05000389 P
PDR 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. NPF-16.
(2)
Technical S ecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 101 are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specification.
E This license amendment is effective as of its date of issuance and shall be implemented by the end of the next scheduled refueling outage, which is currently scheduled to begin in April of 2000.
FOR THE NUCLEAR REGULATORYCOMMISSION 9,. PN~=
Sheri R. Peterson, Chief, Section 2 Project Directorate II Division of Licensing Project Management Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
ATTACHMENT'OLICENSE AMENDMENTNO. 101 TO FACILITYOPERATING LICENSE NO. NPF-16 DOCKET NO. 50-389 Replace the following page of the Appendix "A"Technical Specifications with the enclosed pages.
The revised pages are identified by amendment number and contain vertical lines indicating the areas of change.
Remove Pa e
IX XXII 3/4 9-12 B 3/4 9-3 5-4 5-4A Inse~ Pacne IX XXII 3/4 9-12 B 3/4 9-3 5-4 5-4A thru 5-4F
LIMITINGCONDITIONS FOR OPERATION AND SURVEILLANCEREQUIREMENTS I
I SECTION PAGE 3/4.8.2 3/4.8.3 D.C. SOURCES OPERATING.
SHUTDOWN.......................
ONSITE POWER DISTRIBUTIONSYSTEMS OPERATING...
SHUTDOWN.
....3/4 8-10
.... ~~.....3/4 8-13 3/4 8-14
~.3/4 8-16 3/4.8.4 3/4.9 ELECTRICALEQUIPMENT PROTECTIVE DEVICES MOTOR-OPERATED VALVESTHERMALOVERLOAD PROTECTION BYPASS DEVICES REFUELING OPERATIONS 3/4 8-17 3/4.9.1 BORON CONCENTRATION 3/4.9.2 INSTRUMENTATION.
~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~
~ ~ ~ ~ ~
o3/4 9 1
...............................3/4 9-2 3/4 9.3 3/4.9.4 3/4.9.5 3/4.9.6 DECAYTIME...
CONTAINMENTBUILDINGPENETRATIONS COMMUNICATIONS MANIPULATORCRANE
~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~
~ ~
3/4
..3/4 9-4
~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~
~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 9
...........3/4 9-6 3/4.9.7 CRANE TRAVELSPENT FUEL STORAGE POOL BUILDING.............~. ~~.......3/4 9-7 3/4.9.8 3/4.9.9 3/4.9.10 3/4.9.11 3/4.9.12 3/4 9-8
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~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~
~ ~ ~ ~ ~ ~ ~ ~ ~ ~i3/4 9 9
~.3/4 9-10
..3/4 9-11
.....~...3/4 9-12
........3/4 9-13 SHUTDOWN COOLING AND COOLANTCIRCULATION HIGH WATER LEVEL LOW WATER LEVEL CONTAINMENTISOLATIONSYSTEM.
WATER LEVELREACTOR VESSEL SPENT FUEL STORAGE POOL SPENT FUEL CASK CRANE.....................~...........................
3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 3/4.10.2 3/4.10.3 3/4.10.4 3/4.10.5 SHUTDOWN MARGIN MODERATOR TEMPERATURE COEFFICIENT, GROUP HEIGHT, INSERTION AND POWER DISTRIBUTIONLIMITS REACTOR COOLANT LOOPS CENTER CEA MISALIGNMENT.....................................
CEA INSERTION DURING ITC, MTC, AND POWER COEFFICIENT MEASUREMENTS
.. 3/4 10-1 3/4 10-2 3/4 10-3
....3/4 10-4 3/4 10-5 ST. LUCIE-UNIT2 IX Amendment No. 101
i
LIST OF FIGURES Continued FIGURE INDEX PAGE 3.4-3 3.4-4 4.7-1 B 3/4.4-5.1-1 5.6-1a 5.6-1b 5.6-1c 5.6-1d 5.6-1e 6.2-1 B 3/4 4-10 REACTOR COOLANTSYSTEM PRESSURE-TEMPERATURE LIMITATIONS FOR 15 EFPY, COOLDOWN AND INSERVICE TEST
............... 3/4 4-31b REACTOR COOLANTSYSTEM PRESSURE-TEMPERATURE LIMITATIONS FOR 15 EFPY, MAXIMUMALLOWABLECOOLDOWN RATES...................... 3/4 4-32 SAMPLING PLAN FOR SNUBBER FUNCTIONALTEST 3/4 7-25 1
NIL-DUCTILITYTRANSITIONTEMPERATURE INCREASE'AS A FUNCTION OF FAST (E )1 MeV) NEUTRON FLUENCE (550'F IRRADIATION)FOR REACTOR VESSEL BELTLINEMATERIALS~
SITE AREA MAP.
5-2 REQUIRED FUEL ASSEMBLY BURNUP vs INITIALENRICHMENTand DECAYTIME, REGION II, 1.3 w/o...........................................~..............
5-4B REQUIRED FUEL ASSEMBLY BURNUP vs INITIALENRICHMENTand DECAYTIME, REGION II, 1.5 w/o.....................................................
5-4C REQUIRED FUEL ASSEMBLY BURNUP vs INITIALENRICHMENTand DECAYTIME, REGION I, 1.4 w/o
~.. ~. ~. ~ ~... ~..~.................................
5-4D REQUIRED FUEL ASSEMBLYBURNUP vs INITIALENRICHMENTand DECAYTIME, REGION I, 1.82 w/o..........
~.. ~.~...............................................
5-4E REQUIRED FUEL ASSEMBLY BURNUP vs INITIALENRICHMENT, REGION I, 2.82 w/o.......;.............
~ ~.~...~...........
~. ~.~.......................................
5-4F DELETED.
........................ 6-3 6.2-2 DELETED.
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... 6-4 ST. LUCIE - UNIT2 XXII Amendment No. e,ee,ee, 101
REFUELING OPERATION 3/4.9.11 SPENT FUEL STORAGE POOL ON 3.9.11 The Spent Fuel Storage Pool shall be maintained with:
a.
The fuel storage pool water level greater than or equal to 23 ft over the top of irradiated fuel assemblies seated in the storage racks, and b.
The fuel storage pool boron concentration greater than or equal to 1720 ppm.
APPLICABILITY:Whenever irradiated fuel assemblies are in the spent fuel storage pool ~
ACTION:
a.
With the water level requirement not satisfied, immediately suspend all movement of fuel assemblies and crane operations with loads in the fuel storage areas and restore the water level to within its limitwithin 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
b.
With the boron concentration requirement not satisfied, immediately suspend all-movement of fuel assemblies in the fuel storage pool and initiate action to restore fuel storage pool boron concentration to within the required limit.
c.
The provisions of Specification 3.0.3 are not applicable.
4.9.11 The water level in the spent fuel storage pool shall be determined to be at least its minimum required depth at least once per 7 days when irradiated fuel assemblies are in the fuel storage pool.
4.9.11.1 Verifythe fuel storage pool boron concentration is within limitat least once per 7 days.
ST. LUCIE - UNIT2 3/4 9-12 Amendment No. M1
REFUELING OPERATION 3/4.9.10 and 3/4.9.11 WATER LEVEL-REACTORVESSEL and SPENT FUEL STORAGE POOL i
The restrictions on minimum water level ensure that sufficient water depth is available to remove 99% of the assumed 10% iodine gap activity released from the rupture of an irradiated fuel assembly.
The minimum water depth is consistent with the assumptions of the safety analysis.
The limiton soluble boron concentration in LCO 3/4.9.11 is consistent with the minimum boron concentration specified for the RWT, and assures an additional subcritical margin to the value of keII which is calculated in the spent fuel storage pool criticalitysafety analysis to satisfy the acceptance criteria of Specification 5.6.1. Inadvertent dilution of the spent fuel storage pool by the quantity of unborated water necessary to reduce the pool boron concentration to a value that would invalidate the criticalitysafety analysis is not considered to be a credible event. The surveillance frequency specified forverifying the boron concentration is consistent with NUREG-1432 and satisfies, in part, acceptance criteria established by the NRC staff for approval of criticality safety analysis methods that take credit for soluble boron in the pool water. The ACTIONS required for this LCO are designed to preclude an accident from happening or to mitigate the consequences of an accident in progress, and shall not preclude moving a fuel assembly to a safe position.
3/4.9.12 SPENT FUEL CASK CRANE The maximum load which may be handled by the spent fuel cask crane is limited to a loaded multi-element cask which is equivalent to approximately 100 tons; This restriction is provided to ensure the structural integrity of the spent fuel pool in the event of a dropped cask accident.
Structural damage caused by dropping a load in excess of a loaded multi-element cask could cause leakage from the spent fuel pool in excess of the maximum makeup capability.
ST. L'UCIE - UNIT2 B 3/4 9-3 Amendment No.
DESIGN FEATURES VOLUME 5.4.2 5.5 The total water and steam volume of the reactor coolant system is 10,931 2 275 cubic feet at a nominal T, of 572'.
METEOROLOGICALTOWER LOCATION 5.5.1 The meteorological tower shall be located as shown on Figure 5.1-1.
5.6 FUEL STORAGE CRITICALITY 5.6.1 a.
The spent fuel pool and spent fuel storage racks shall be maintained with:
1.
A k,equivalent to less than 1.0 when flooded with unborated water, including a conservative allowance for biases and uncertainties as described in Section 9.1 of the Updated Final Safety Analysis Report.
2.
A k,equivalent to less than or equal to 0.95 when flooded with water containing 520 ppm boron, including a conservative allowance for biases and uncertainties as described in Section 9.1 of the Updated Final Safety Analysis Report.
3.
A nominal 8.96 inch center-to-center distance between fuel assemblies placed in the storage racks.
5.6.1 b.
Fuel placed in Region I of the spent fuel storage racks shall be stored in a configuration that willassure compliance with 5.6.1 a.1 and 5.6.1 a.2, above, with the following considerations:
1.
Fresh fuel shall have a nominal average U-235 enrichment of less than or equal to 4.5 weight percent.
2.
The reactivity effect of CEAs placed in fuel assemblies may be considered.
3.
The reactivity equivaiencing effects of burnable absorbers may be considered.
4.
The reactivity effects of fuel assembly burnup and decay time may be considered as specified in Figures 5.6-1 c through 5.6-1 e.
5.6.1 c.
Fuel placed in Region II of the spent fuel storage racks shall be placed in a configuration that willassure compliance with 5.6.1 a.1 and 5.6.1 a.2, above, with the following considerations:
Fuel placed in Region II shall meet the burnup and decay time requirements specified in Figure 5.6-1 a or 5.6-1b.
2.
The reactivity effect of CEAs placed in fuel assemblies may be considered.
3.
The reactivity equivalencing effects of burnabie absorbers may be considered.
ST. LUCIE - UNIT2 5-4 Amendment No. P,W M1
DESIGN FEATURES cont' CRITICALITY(continued) 5.6.1 d.
The new fuel storage racks are designed for dry storage of unirradiated fuel assemblies
'having a U-235 enrichment less than or'equal to 4.5 weight percent, while maintaining a k,of less than or equal to 0.98 under the most reactive condition.
DRAINAGE 5.6.2 The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 56 feet.
CAPACITY 5.6.3 The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 1360 fuel assemblies.
5.7 COMPONENT CYCLIC OR TRANSIENT LIMITS 5.7.1 The components identified in Table 5.7-1 are designed and shall be maintained within the cyclic or transient limits of Table 5.7-1.
ST. LUCIE-UNIT2 5-4A Amendment No. g 101
Figure 5.6-1 a Required Fuel Assembly Burnup vs Initial Enrichment and Decay Time Region II, 1.3 w/o 50000 C
Pn 40000 2
U'C g
30000 I
20000 Ha 10000 r.
C 0
1.5 2.0 Acceptable Burnup 2.5 3.0 3.5 initial U-235 Enrichment (w/o) 4.0 0 years 5 years 15 years 20 years 4.5 (DA.IC/CEND497-F5.6. ra-RO) 5.0 0 0
40000 I
30000
(
20000
- U<
1OOOO Figure 5.6-1 b Required Fuel Assembly Burnup vs Initial Enrichment and Decay Time Region II, 1.5 w/0 Acceptable Burnup years 5 years 10 years 15 years 20 years 1.5 2.0 2.5 3.0 3.5 4.0 4.5 5.0 Initial U-235 Enrichment (w/o)
(D/LIC/CENCh387+5.6. Jb-RQj
Ql U
~ 50000 c
40000 C
30000 I
C c
20000 r
C7 P 10000 C
Figure 5.6-1 c Required Fuel Assembly Burnup vs Initial Enrichment and Decay Time Region I, 1.4 w/o Acceptable Burnup 0 years 5-yeafs 10 years 15 years 20 years 0
3 H.
'Z Q
1.5 2.0 2.5 3.0 3.5 Initial U-235 Enrichment (w/o) 4.0 5.0 (D/LIC/CEND487-F5.6. J cd)
CO l
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z
=I lO Ql m
35000 e
30000 25000 ta 20000 Wc H
15000 10000 C
5000 Acceptable Burnup 0
ears 5 years 10 years 15 ears 20 years Figure 5.6-1 d Required Fuel Assembly Burnup vs Initial Enrichment and Decay Time Region I, 1.82 w/o 1.5 2.0 2.5 3.0 3.5 4.0 4.5 5.0
~ 2z O
O Initial U-235 Enrichment (w/o)
(0/UC/CENLh387+5.6. rd40)
M I
O m
Z fO Figure 5.6-1e Required Fuel Assembly Burnup vs Initial Enrichment Region I, 2.82 w/o Ql n
P-10000 CL C
Kl J2 E
Cltll GXO LL Acceptable Burnup > - 484.92'E"3 + 4504.5'E"2 - 6086 6'E - 7783.1 0 years O
0 1.5 2.0 2.5 3.0 3.5 Initial U-235 Enrichment (w/o) 4.0 4.5 5.0 (MJG'F5 6 1E~