ML20203F651
| ML20203F651 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood  (NPF-72-A-086, NPF-72-A-86, NPF-77-A-086, NPF-77-A-86) |
| Issue date: | 12/04/1997 |
| From: | Dick G NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20203D363 | List: |
| References | |
| NUDOCS 9712170431 | |
| Download: ML20203F651 (20) | |
Text
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- UNITED STATES l
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NUCLEAR REGULATORY COMMISSION
- o WA9MINGToN, D.C. EssaHe01 COMMONWEALTH EDISON COMPANY DOCKET NO. STN 50-456 BRAIDWOOD STATION. UNIT NO.1 1
AMENDMENT TO FAClljTY OPERATING LICENSE i
Amendment No. 86 License No. NPF 72 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The applicatjon for amendment by Commonwealth Edisoi) Company (the I
licensee) dated June 50,1997, as supplemented on September 25,1997, complies with the stai)dards and requirements of the Atomic Energ Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chaptor 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 assura,nce (i) that the activities authorized by this amendment can be conducted without endangering tile health and safety of the public, and (ii) that such activities will be condueled 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 publ!:; 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 thw Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF 72 is hereby amended to read as follows:
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Technical Soecifications The Technical Specifications contained in Appendix A as revised through Amendment No.o6 and the Environmental Protection Plan contained in Appendix B, both of which are attact.ed hereto, are hereby incorporated into this license.
The licensee shall operate the facility in accordance with the Technical Specifications and the Environmt 'al Protection Plan.
3.
This license amendment is effective as of the date of its issuance and shall be implemented within 30 days.
FOR THE NUCLEAR REGULATORY COMMISSION i
Geor F. Dick, Senior Project Manager Project Directorate 1112 Division of Reactor Projects - lil/lV Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of issuance: December 4, 1997 i
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'i UNITED STATES j<
i NUCLEAR REIULATORY COMMisslON waswiwoToN, D.C. M1
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COMMONWEALTH EDISON COMP #,1f(
DOCKET NO. STN 50-457 BRAIDWOOD STATION. UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. ~ 86 License No. NPF-77 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Commonwealth Edison Company (the licensee) dated June 30,1997, as supplemented on September 25,1997, 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 1; 9.
The facility will operate in confonnity 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 and safety of ti,e public, and (ii) that such activities will be conducted in compliance with the Commission's regulation ;
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 l
E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the l
Commission's regulations and all applicable requirements have been satisfied.
l 2.
Accordingly, the license is amended by changes b the Technical Specifications as indicated in the attachment to this license amend.nent, and paragrap 2.C.(2) of Facility Operating License No. NPF 77 is hereby amended to read as follows:
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Technical Soecifications The Technical Specifications contained in Appendix A as revised through Amendment No.86 and the Environmental Protection Plan contained in Appendix B, both of which were attached to License No. NPF 72, dated July 2,1987, are hereby incorporated into this license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
This license amendment is effective as ni the date if its issuance and shall be implemented within 30 days.
FOR THE NUCLEAR REGULATORY COMMISSION g
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T E
Georg F. Dick, Senior Project Manager Project Directorate 1112 Division of Reactor Projects lil/IV Office of Nue. lear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: December 4, 1997 l
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ATTACHMENT TO LICENSE AMENDMENT NOS.86 AND 86 FACILITY OPERATING LICENSE NOS. NPF-72 AND NPF-77 DOCKET NOS. STN 50-456 AND STN 50-457 Replace the following pages of the Apper' dix "A" Technical Specifications with the attached pages. The revised pages are identified by amendment number and contain verticallines ladicating the area of change. Pages indicated by an asterisk are provided fur convenience ony.
Remove Paoes insert Paoes 1
I Xll Xil XVil XVil XVill XVill XX XX 1 1*
11*
1-2 12 3/4 9-13 3/4 9-13 8 3/4 9-3 B 3/4 9-3 5 3*
5 3*
5-4 5-4 5-5 5-5 5 5b 5-Sb 5-Sc 5 5d 6-23 6-23
.lHQlX DEFINITIONS SECTION PAGE 1.0 DEFINITIONS 1.1 ACTI0N........................................................
1-1 1.2 ACTUATION LOGIC TEST..........................................
1-1 1.3 ANALOG CHANNEL OPERATIONAL TEST..............................
1-1 1.4 AXIAL FLUX DIFFERENCE.........................................
1-1 l.5 CHANNEL CALIBRATION...........................................
1-1 1.6 CHANNEL CHECK.................................................
1-1 1.7 CONTAINMENT INTEGRITY.........................................
1-2 1.8 CONTROLLED LEAKAGE............................................
1-2 1.9 CORE ALTERATION................................................
1-2 l
1.10 DIGITAL CHANNEL OPERATIONAL TEST.............................
1-2 1.11 DOSE EQUIVALENT I-131...........................
1-2 1.12 E-AVERAGE DISINTEGRATION ENERGY.............'.................
1-3 1.13 ENGINEERED SAFETY FEATURES RESPONSE TIME......................
1-3 1.14 FREQUENCY N0TATION...........................................
1-3 1.15 IDENTIFIED LEAXAGE...........................................
1-3 1.15.a L,..........................................................
1-3 1.16 MASTER R; LAY TEST.............................................
1-3 1.17 MEMBER (S) 0F THE PUBLIC......................................
1-3 1.18 0FFSITE DOSE CALCULATION MANUAL..............................
1-4 1.19 OPERABLE - OPERABILITY.......................................
1-4 1.19.a OPERATING LIMITS REP 0RT.....................................
1-4 1.20 OPERATIONAL MODE - N0DE......................................
1-4 1.20.a P,..........................................................
1-4 1.21 PHYSICS TESTS................................................
1-4 1.22 PRESSURE BOUNDARY LEAKAGE....................................
1-4 1.23 PROCESS CONTROL PR0 GRAM......................................
1-5 1.24 PURGE - PURGING..............................................
1-5 1.25 QUADRANT POWER T I LT RATI0....................................
1-5 1.26 RATED THERMAL P0WER..........................................
1-5 1.27 REACTOR TRIP SYSTEM RESPONSE TIME............................
1-5 1.28 REPORTABLE EVENT.............................................
1-5 BRAIDWOOD - UNITS 1 & 2 I
AMENDMENT NO. 86
LIMITING C6NDITIONS FOR OPERATION AND SURVEILLANCE REOUIREMENTS SECTION PA_G1 TABLE 3.8-2a MOTOR OPERATED VALVES THERMAL OVERLOAD PROTECTION DEVICES (UNIT 1)....................................
3/4 8-40 TABLE 3.8-2b MOTOR OPERATED VALVES THERMAL OVERLOAD PROTECTION DEVICES (UNIT 2)....................................
3/4 8-44 3/4 9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION......................................
3/4 9-1 3/4.9.2 INSTRUMENTATION..........................................
3/4 9-2 3/4.9.3 DECAY TIME...............................................
3/4 9-3 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS........................
3/4 9-4 3/4.9.5 COMMUNICATIONS...........................................
3/4 9-6 3/4.9.6 REFUELING MACHINE........................................
3/4 9-7 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE FACILITY...............
3/4 9-8 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION High Water Level.........................................
3/4 9-9 Low Water Level..........................................
3/4 9-10 3/4.9.9 CONTAINMENT PURGE ISOLATION SYSTEM.......................
3/4 9-11 3/4.9.10 WATER LEVEL - REACTOR VESSEL.............................
3/4 9-12 3/4.9.11 WATER LEVEL / BORON CONCENTRATION - STORAGE P00L...........
3/4 9-13 3/4.9.12 FUEL HANDLING BUILDING EXHAUST FILTER PLENUMS............
3/4 9-14 BRAIDWOOD - UNITS 1 & 2 XII AMENDMENT NO. 86
. = -.
Bl,SES SECTION ggi 3/4.9.6 REFUELING MACHINE.........................................
3 3/4 9-2 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE FACILITY................
B 3/4 9-2 3/4.9.9 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION.............
B 3/4 9-2 3/4.9.9 CONTAINMENT PURGE ISOLATION SYSTEM........................
B 3/4 9-3 3/4.9.10 WATER LEVEL - REACTOR VESSEL..............................
B 3/4 9-3 3/4.9.11 WATER LEVEL / BORON CONCENTRATION - STORAGE P00L............
B 3/4 9-4 3/4.9.12 FUEL HANDL!llG BUILDING EXHAUST FILTER PLENUM..............
B 3/4 9-3 3/4.10 SPECIAL TEST EXCEPTIONS l
3/4.10.1 SHUTDOWN MARGIN...........................................
B 3/4 10-1 4
3/4.10.2 GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION LIMITS....
B 3/4 10-1 3/4.10.3 PHYSICS TESTS.............................................
B 3/4 10-1 3/4.10.4 REACTOR COOLANT L00PS.....................................
B 3/4 10-1 3/4.10.5 POSITION INDICATION SYSTEM - SHUTD0WN.....................
B 3/4 10-1 3/4.11 RADIOACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS Liquid Holdup Tanks..................................
B 3/4 11-1 3/4.11.2 GASEOUS EFFLUENTS Explosive Gas Mirture.................................
B 3/4 11-2 Gas Decay Tanks.......................................
B 3/4 11-2 BRAIDWOOD - UNITS 1 & 2 XVII AMENDMENT N0. 86
DESIGN FEATURES SECTION EAE 5.1 SITE 5.1.1 EXCLUSION AREA..............................................
5-1 5.1.2 L OW PO PUL AT I ON 2 0N E.........................................
5-1 5.1.3 MAP DEFINING UNRESTRICTED AREAS AND SITE BOUNDARY FOR RADIOACTIVE GASEOUS AND LIQUID EFFLUENTS................
5-1 FIGURE 5.1-1 EXCLUSION AREA AND UNRESTRICTED AREA FOR RADIDACTIVE GASEOUS AND LIQUID EFFLUENTS.............
5-2 FIGURE 5.1 2 LOW POPULATION Z0NE..................................
5-3 5.2 CONTAINMENT 5.2.1 CONFIGURATION...............................................
5-1 5.2.2 DESIGN PRESSURE AND TEMPERATURE.............................
5-1 5.3 REACTOR CORE 5.3.1 FUEL ASSEMBLIES.............................................
5-4 5.3.2 CONTROL ROD ASSEMBLIES......................................
5-4 5.4 REACTOR COOLANT SYSTEM 5.4.1 DESIGN PRESSURE AND TEMPERATURE.............................
5-4 5.4.2 V0LUME......................................................
5-4 5.5 METEOROLOGICAL TOWER LOCAT10N.................................
5-4 5.6 FUEL STORAGE 5.6.1 CRITICALITY.................................................
5-5 FIGURE 5.6-1 MINIMUM BURNUP VERSUS INITIAL ENRICHMENT FOR REGION 2 - ALL CELL CONFIGURAT10N STORAGE.......
5-5b FIGURE 5.6-2 MINIMUM BURNUP VERSUS INITIAL ENRICHMENT FOR REGION 2 0VT-OF-4 CHECKERBOARD CONFIGURATION.......................................
5-5c FIGURE 5.6-3 MINIMUM BURNUP VERSUS INITIAL ENRICHMENT FOR REGION 2 00T-0F-4 CHECKERBOARD CONFIGURATION.......................................
5-5d 5.6.2 DRAINAGE........................................
5-5 5.6.3 CAPACITY....................................................
5-5 5.7 C OM PON ENT C YC L I C OR T RANS I ENT L I M I T...........................
5-5 TABLE 5.7-1 COMPONENT CYCLIC OR TRANSIENT LIMITS..................
5-6 BRAIDWOOD - UNITS 1 & 2 XVIII AMENDMENT NO. 86
ADMINISTRATIVE CONTROLS SECTION EAgE 6.7 SAFETY LIMIT VICLATI0N........................................
6-15 6.8_ PROCEDURES.AND.PR0 GRAMS.......................................
6-16 6.9 REPORTING REQUIREMENTS........................................
6-20 6.9.1 ROUTINE REP 0RTS.............................................
6-20 Startup Report..............................................
6-20 An n u al Re po rt s..............................................
6-20 Annual Radiological Environmental Operating Report..........
6-22 Annual Radioactive Effluent Release Report..................
6-22 Honthly Operating Report....................................
6-22 Operating Limits Report.....................................
6-22
{
6.9.2 SPECIAL REP 0RTS.............................................
6-23 6.10 RECORD RETENTION.............................................
6-23 6.11 RADIATION PROTECTION PROGRAM................................
6-24 6.12 HIGH RADIATION AREA..........................................
6-25 6.13 PROCESS CONTROL PROGRAM (FCP)................................
6-26 6.14 0FFSITE DOSE CALCULATION MANUAL (00CM).......................
6-26 l
BRAIDWOOD - UNITS 1 & 2 XX AMENDNENT NO. 86
1.0 -DEFINITIONS 3 -.
The defined tems of this section appear in capitalized type and are applicable
.throughout these Technical Specifications.
J-I ACTION 1.1 ACTION shall be that part of a Technical Specification which prescribes
-remedial measures required under designated conditions.
il ACTUATION LOGIC TEST-1.2 An ACTUATION LOGIC TEST shall_be the application of various simulated 3
input combinations in conjunction with each possible interlock logic state and verification of the required logic output.
The ACTUATION LOGIC TEST shall i
include a continuity check, as a minimum, of output devices.
4 ANALOG CHANNEL OPERATIONAL TEST 1.3 A1 ANALOG CHANNEL OPERATIONAL TEST shal! be the injection of a simulated signal into the channel as close to the sensor as practicable to verify 0PERA8Il.ITY of alarm, interlock and/or trip functions. The ANALOG CHANNEL OPERATIONAL TEST shall-include adjustments, as necessary, of the alarm, inter-lock and/or. Trip Setpoints such that the Setpoints are within the required range and accuracy.
AXIAL FLUX DIFFERENCE 1.4 AXIAL FLUX DIFFERENCE shall be the difference in normalized flux signals between the top and bottom halves of a two section excore neutron detector.
. CHANNEL CALIRRATION
- 1. 5 A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel such that it responds within the required range and accuracy to known values of input.
The CHANNEL CALIBRATION shall encompass the entire channel including the sensors and alarm, interlock and/or trip functions and may be performed by any series of sequential, overlapping, or total channel steps such that the entire channel is calibrated.
CHANNEL CHECK
- 1. 6 A CHANNEL CHECK shall be the qualitative assessment of channel behavior during operation by observation.
This determination shall include, where possible, comparison of the channel indication and/or status with other 4
indications and/or status derived from independent instrument channels measuring the same parameter.
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DEFINITIONS
[0NTAINMENT INTEGRITY 1.7 CONTAINMENT INTEGRITY shall exist when:
a.
All penetrations required to be closed during accident conditions are either:
1)
Capable of being closed by an OPERABLE containment automatic isolation valve system, or 2)
Closed by manual valves, blind flanges, or deactivated automatic valves secured in their closed positions, except as provided in Table 3.6-1 of Specification 3.6.3.
b.
All equipment hatches are closed and sealed, i
c.
Each air lock is in compliance with the requirements of Specification 3.6.1.3, d.
The containment leakage rates are within the limits of Specification 3.6.1.2, and e.
The sealing mechanism associated with each penetration (e.g., welds, bellows, or 0-rings) is OPERABLE.
CONTROLLED LEAKAGE 1.8 CONTROLLED LEAKAGE shall be that seal water flow supplied to the reactor coolant pump seals.
CORE ALTERATION 1.9 CORE ALTERATION shall be the movement or manipulation of any component within the reactor vessel with the vessel head removed and fuel in the vessel.
Suspension of CORE ALTERATION shall not preclude completion of movement of a component to a safe conservative position.
DIGITAL CHANNEL OPERATIONAL TEST 1.10 A DIGITAL CHANNEL OPERATIONAL TEST shall consist of exercising the digital computer hardware using data base manipulation and injecting simulated process data to verify OPERABILITY of alarm and/or trip functions.
DOSE EQUIVALENT I-131 1.11 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcurie / gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131,1-132,1-133, I-134, and I-135 actually present.
The thyroid dose conversion factors used for this calculation shall-be those listed in Table III of TID-14844, " Calculation of Distance Factors for Power and Test Reactor Sites."
BRAIDWOOD - UNITS 1 & 2 1-2 AMENDMENT N0. 85
x REFUELING OPERATIONS 3/4.9.11 WATER LEVEL / BORON CONCENTRATION - STORAGE POOL i
LIMITING CONDITION FOR OPERATION 3.9.11 At least 23 feet of water shall be maintained over the top of irradiated fuel assemblies seated in the storage racks.
The dissolved boron concentration of the water in the storage pool shall be maintained at greater than or equal to 2000 ppe.
l APPLICABILIII:
Whenever irradiated fuel assemblies are in the storage pool.
ACTION:
a.
With the water level requirements of the above specification not satisfied, suspend all movement of fuel assemblies and crane operations with loads in the fuel storage areas and restore the water level to within its limit within 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 requirements of the above specification not s&tisfied, suspend all movement of fuel assemblies and crane operations with loads in the fuel storage areas and immediately take action to restore the dissolved boron concentration to within its limit as soon as possible.
l c.
The provisions of Specification 3.0.3 are not applicable.
i i
SURVEILLANCE REQUIREMENTS l
4.9.11 The water level in the 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.ll.a Boron concentration in the storage pool shall be determined to be greater than or equal to 2000 ppm at least once per 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.
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BRAIDWOOD - UNITS 1 & 2 3/4 9-13 AMENDMENT NO. 85
REFUEllNG OPERATIONS
~'
' BASES 3/4.9.S RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION (Continued)
The requirement to have two RHR loops OPERABLE when there is less than 23 feet of water above the reactor vessel flange ensuras that a single failure of the operating RHR loop will not result in a complete lo:;s of RHR capability.
With the reactor vessel head removed and at least 23 feet of water above the reactor vessel flange, a large heat sink is available for core cooling. Thus, in the event of a failure of the operating RHR loop, adequate time is provided to initiate emergency procedures to cool the core.
3/4.9.9 CONTAINMENT PURGE ISOLATION SYSTf3 The OPERABILITY of this system ensures that the containment surge penetrations will be automatically isolated upon detection of hig1 radiation levels within the containment. The OPERABILITY of this system is required to restrict the release of radioactive material from the containment atmosphere to the environment.
3/4.9.10 WATER LEVEL - REACTOR VESSEL 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.
3/4.9.11 WATER LEVEL / BORON CONCENTRATION - STORAGE POOL 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 restrictions on soluble boron concentration in the storage pool water ensure the spent fuel rack k will be maintained less than or equal to 0.95 m
i j
with a 95-percent confidence level.
3/4.9.12 FUEL HANDLING BUILDING EXHAUST FILTER PLENUM The limitations on the fuel Handling Building Exhaust filter Plenum ensure that all radioactive material released from an irradiated fuel assembly will be filtered through the HEPA filters and charcoal adsorber prior to dis-i charge to the atmosphere. The OPERABIlliiY of this system and the resulting iodine removal capacity are consistent with the assumptions of the safety analyses. ANSI N510-1980 will be used as a procedural guide for surveillance testing.
BRAIDWOOD - UNITS _1 & 2 B 3/4 9-1 AMEliDMENT NO. 86 l
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PUBLIC FACILITIES AND INSTITUTIONS WITHIN 1.125 MILES OF THE SITE BRAIDWOOD - UNITS 1 & 2 5-3
DESIGN FEATURES i
l 5.3 REACTOR CORE i
FUEL ASSEMBLIES 5.3.1 The core shall contain 193 fuel assemblies with each fuel assembly I
containing 264 fuel rods clad with Zircaloy-4 or ZIRLO, except that limited substitut' on of fuel rods by filler rods consisting of Zircaloy-4,- ZIRLO, or stainless steel or by vacancies may be made if justified by a cycle specific i
reload analysis.
Each fuel rod shall have a nominal active fuel length of.
j 4
i 144 inches. The initial core loading shall have a maximum enrichment of less i
than 3.20 weight percent U-235. Reload fuel shall be similar in physical
' design to the initial core loading or previous cycle loading. The enrichment of any reload fuel design shall be detemined to be acceptable for storage in j
either the spent fuel pool or the new fuel vault.
Such acceptance criteria t
i shall be based on the results of the '8yron and Braidwood Spent Fuel Rack l
Criticality Analysis Using Soluble Boron Credit,' May 1997, CAC-97-162 and
-' Criticality Anasysis of the Byron /Braidwood Fresh Fuel Racks," June-1989.
CONTROL ROD ASSEM8 LIES s
5.3.2 The core shall contain 53 full-length and no part-length control rod assemblies. The full-length control rod assemblies shall contain a nominal 142 inches of absorber material. 1.11 control rods shall be hafnium, silver-indium-cadmium, or a mixture of both types. All controi rods shall be clad with stainless steel tubing.
i i.4 REACTOR COOLANT SX112 c
DESIGN PRESSURE AND TEMPERATURE 5,4.1 The Reactor Coolant System is designed and shall be maintained:
a.
In accordance with the Code requirements specified in Section 5.2 of the UFSAR, with allowance for normal degradation pursuant to the l-applicable Surveillance Requirements, i
b.
For a pressure of 2485 psig, and 4
c.
For a totperature of 650'F, except for the pressurizer which is 680*F.
VOLUME 5.4.2 The total water and steam volume of the Reactor Coolant System is
+
12,257 cubic feet at a nominal T, of 586.4*F.
i 5.5 METEOROLOGICAL TOWER LOCATION 5.5.1 The meteorological tower shall be located as shown on Figure 5.1-1.
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BRAIDWOOD - UNITS 1 & 2 5-4 Amendment No. 86 i
DESIGN FEATURES
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5.6 FUEL STORAGE CRITICALITY l
5.6.1.1 The spent fuel storage racks are designed and shall be maintained
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with:
a.
Fuel assemblies having a maximum initial U-235 enrichment of 5.0 weight percent; b.
A k,,, < l.0 if fully flooded with unborated water, which includes an allowance for uncertainties as described in WCAP-14416-NP-A,
" Westinghouse Spent Fuel Rack Criticality Analysis with Credit for Soluble Boron " Revision 1, November 1996; A k ' des an allowance for uncertainties as described in WCAP-14 5 0.95 if fully flooded with water borated to 550 ppm, which c.
inclu NP-A, " Westinghouse Spent Fuel Rack Criticality Analysis with Credit for Soluble Boron," Revision 1, November 1996; d.
A nominal 10.32 inch north-south and 10.42 inch east-west center-to-center distance between fuel assemblies placed in the Region I racks; e.
New or spent assemblies with sufficient Integral Fuel Burnable Absorbers present in each fuel assembly, as rescribed in the " Byron and Braidwood Spent Fuel Rack Criticality An 'vsis Using Soluble Boron Credit," May 1997, CAC-97-162, which m;y be allowed unrestricted storage in the Region I racks; f.
A nominal 9.03 inch center-to-center distance between fuel assemblies placed in the Region 2 racks; g.
New or spent fuel assemblies with a combination of discharge burnup, initial enrichment, and decay time in the acceptable region of Figures 5.6-1, 5.6-2, or 5.6-3, as appi tcable, which may be stored in the Region 2 racks in the applicable checkerboard configuration, as described in the " Byron and Braidwood Spent Fuel Rack Criticality Analysis Using Soluble Boron Credit," May 1997 CAC-97-162; and h.
Interface requirements within and between adjacent racks as described in the " Byron and Braidwood Spent Fuel Rack Criticality Analysis Using Soluble Boron Credit," May 1997, CAC-97-162.
5.6.1.2 The k,,
for new fuel for the first core loading stored dry in the spent fuel stora,ge racks shall not exceed 0.98 when aqueous foam moderation is assumed.
DRAINAGE 5.6.2 The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 423 feet 0 inches.
BRAIDWOOD - UNITS 1 & 2 5-5 Amendment No. 36
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1.0 2.0 3.0 4.0 5.0 Initial U-235 Enrichment (w/c)
Note:
The use oflinear interpolation between the minimum burnups is acceptable.
FIGURE 5.6-1 h0NIMUM BURNUP VERSUS INITIAL ENRICHMENT FOR REGION 2 ALL CELL CONFIGURATION STORAGE BRAIDWOOD - UNITS 1 & 2 5-Sb AENN NO. 86 i
i 1
. - - ~. -. - -. -..
~
J 45000 s
4 1
J 40000 e teano j
/
ss f,
i
/
/. se I
^ 35000
)
) I ss :eme,
/
/
rf se seen, j
1
) ) fi r/
1
/
///I ACCEPTABLE
/ //s 30000
/. fJ//
) f rir
/- f/ff
) / FA
/ //r/
g 25000
> // />
////f
/ /> r//
M
///A f
// W f
20000
///A r
l F/M l
/ F/r rn v 1
M
/ fd I 15000
/h V b
J'f2
/A 7
. rm A V 10000 y
r A
(
A l
ACC M ABM 5000 s
F A
/
Q I
\\
1.0 2.0 3.0 4.0 5.0 2nitial U-235 Enrichment (w/o)
Note:
The use oflinear interpolation between the minimum burnups is acceptable.
FIGURE 5.6-2 MINIMUM BURNUP VERSUS INITIAL ENRICHMENT FOR REGION 2 3 OUT-OF-4 CHECKERBOARD CONFIGURATION BRAIDWOOD - UNITS 1 & 2 M
AMENDMENT NO. 86 l
..u--
mm-m ww.-
rr-m,wva------vam-.-rr--a s-e---
u-.em-+woww+--w-m,rw%e----
yw y--w++r wwm3.y
%,--,---p w.
%q wywp g,%+,9
-my y
l l
5000 i
)
}
r 400C-f
~
t f
2
{
\\
)
ACCEPTABLE
/
3000
}
i
)
)
i k
/
F
)
(
j 2000 g
.=
.s
/
g f
/
NOT ACCEPTABLE i
100C
/
4
/
r
)
l f
0 i
4.0 4.2 4.4 4.6 4.8 5.0 Initial U-235 Enrichment (w/o)
Note:
The use of knear interpolation between the minimum bumups is acceptable, t
FIGURE 5.6-3 MINIMUM BURNUP VERSUS INITIAL ENRICHMENT FOR REGION 2 2-OUT-OF-4 CHECKERBOARD CONFIGURATION r
BRAIDWOOD - UNITS-1 & 2 5-5d AMENIMENT NO. 86
1 ADMINISTRATIVE CONTROLS SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the Regional Administrator of the NRC Regional Office within the time period specified for each report.
6.10 RECORD RETENTION In addition to the applicable record retention requirements of Title 10, Code of Federal Regulations, the following records shall be retained for a*
1 east the minimum period indicated.
6.10.1 The following records shall be retained for at least 5 years:
a.
Records and logs of unit operation covering time interval at each power level; b.
Records and logs of principal maintenance activities, inspections, repair and replacement of principal items of equipment related to nuclear safety; c.
All REPORTABLE EVENTS; d.
Records of surveillance activities, inspections, and calibratione.
required by these Technical Specifications; e.
Recon > of changes made to the procedures required by Specification 6.8; f.
Records of radioactive shipments; g.
Records of sealed wrce and fission detector leak tests and results; and h.
Records of annual physical inventory of all sealed source material of record.
6.10.2 The following records shall be retained for the duration of the unit Operating License:
a.
Records and drawing changes reflecting unit design modifications made to systems and equipment described in the Final Safety Analysis Report; b.
Records of new and irradiated fuel inventory, fuel transfers and assembly burnup histories; SRAIDWOOD - UNITS 1 & 2 6-23 AMENDMENT NO. 86
_