ML20203D358
| ML20203D358 | |
| Person / Time | |
|---|---|
| Site: | Byron  |
| Issue date: | 12/04/1997 |
| From: | Dick G Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20203D363 | List: |
| References | |
| NPF-37-A-094, NPF-66-A-094 NUDOCS 9712160174 | |
| Download: ML20203D358 (20) | |
Text
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'a UNITED STATES i
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NUCLEAR REOULATORY COMMISSION e
WAsHINcToN D.C. 3066Hlo01 o
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COMMONWEALTH EDISON COMPANY DOCKET NO. STN 50-454 BYRON STATION. UNIT NO 1 AMENL; MENT TO FACILIT( OPERATING LICENSE Amendment No. 94 License No. NPF-37 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Commonwealih 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) cad the Commission's rules and regulations set forth ) 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 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. NPF-37 is hereby amended to read as follows:
9712160174 971204 PDR ADOCK 0500 4
P
. (2) Technical Soecifications The Technical Specifications contained in Appendix A as revised through Amendment No.94 and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, 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 of the date of its issuance and shall be implemented within 30 days.
FOR THE NUCLEAR REGULATORY COMMISSION Geo
. Dick, Senior Project Manager Project Directorate 111-2 Division of Reactor Projects - lil/IV Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of issuance: December 4, 1997
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i UNITED STATES g
j NUCLEAR REGULATORY COMMISSION o,
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WASHINGTON, D.C. 206 2 0001
%,..... f COMMONWEALTH EDISCN COMPANY DOCKET NO. STN SMSS BYRON STATION. UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 94 License No. NPF-66 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Cor monwealth Edison Company (the licensee) dated June 30,1997, as supplemerited on September 25,1997, complies with the standaros 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:
B.
The facility will operate in confom11ty 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 autilorized 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 publir; 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. NPF-66 is hereby amended to read as fo!!ows:
1 2
(2)
Technical Soecifications The Technical Specifications contained in Appendix A (NUREG-1113), as revised through Amendment No.94 cnd revised by Attachmer,t 2 to NPF-66, and the Environmental Protection Plan contained in Appeadix B, both of which were attached to License No. NPF-37, dated February 14,1985, are hereby incorporated into this license. Attachment 2 contains a revision to Appendix A which is 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 of the date of its issuance and shall be implemented within 30 days.
FOR THE NUCLEAR REGULATORY COMMISSION Georg Dick, Senior Project Manager Project Directorate ill 2 Division of Reactor Projects - lil/IV Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of issuance: December 4, 1997 l
e.
ATTACHMENT TO LICENSE AMENDMENT NOS 94 AND 9u FACILITY OPERATING LICENSE NOS. NPF-37 AND NPF-66 DOCKET NOS. STN 50-454 AND STN 50-455 Revise the Appendix A Technical Specifications by removing the pages identified below and inserting the attached pages. The revised pages are identified by the captioned amendment number and contain marginal lines indicating the area of change.
Pages indicated by an asterisk are provided for convenience only.
Remove Paaes insert Paoes l
I Xil Xil XVil XVil XVll!
XVill XX XX 1-1*
1-1*
1-2 1-2 3/4 9-13 3/4 9-13 8 3/4 9-3 B 3/4 9-3 5-4 5-4 55 5-5 5-Sb 5-5b 5-Sc 5-5d 6-23 6-23 i
wo l
INDEX DEFINITIONS SECTION EAEE 1.0 DEFINITIONS 1.1 ACTI0N........................................................
1-1 1.2 -ACTUATION LOGIC TEST..........................................
1-1 1.3 ANALOG CHANNEL OPERATIONAL TEST...............................
1 1.4 AXIAL FLUX DIFFERENCE.........................................
1-1 1.5 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 ALT E RAT I ON...............................................
1-2 l
1.10 DIGITAL CHANNEL OPERATIONAL TEST.............................
1-2 1.11 DOSE EQUIVALENT I-131........................................
1-2a 1.12 E-AVERAGE DISINTEGRATION ENERGY..............................
1-3 1.13 ENGINEERED SAFETY FEATURES RESPONSE TIME.....................
1-3
.14 FREQUENCY N0TATION...........................................
1-3 1.15 IDENTIFIED LEAKAGE...........................................
1-3 1.15.a L,..........................................................
1-3 1.16 MASTER RELAY 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 - M0DE......................................
1-4 1.20.a P,..........................................................
1-4 1.21 PHYSICS TESTS................................................
1-4 1.22 PRESSURE BOUNDARY LEAXAGE....................................
1-4 1.23 PROCESS CONTROL PR0 GRAM......................................
1-5 1.24 P U RG E - PU hG I N G..............................................
1-5 1.25 QUADRANT POWER TILT RAT10....................................
1-5 1.c6 RATED TH E RMAL P0WER..........................................
1-5 1.27 REACTOR TRI P SYSTEM RESPONSE TIME............................
1-5 1.28 REPORTABLE EVENT.............................................
1-5 BYRON - UNITS 1 & 2 I
AMENDMENT NO. 9,
LIMITING CONDITIONS _FOR OPERATION AND SURVEILLANCE REOUTREMENTS 1EC.UQM E2GE Motordperated Valves Thermal Overload Protectio,
. Devices................................................
3/4 8-39 TABLE 3.8-2a MOTOR OPERATED VALVES THERMAL OVERLOAD PROTECTION DEVICES (UNIT 1)....................................
3/48-40 TABLE 3.8-2b MOTOR OPERATED VALVES THERMAL OVERLOAD PROTECTION DEVICES (UNIT 2)....................................
3/4B-44 3 /4. 9 REFUELING OPERATIONS 3/4.C.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 Leve1.........................................
3/4 9-9 Low Water Level..........................................
3/4 9-10 3/4.9.9 CONTAINMENT PURGE ISOLATION SYSTEM.......................
3/49-11 3/4.9.10 WATER LEVEL - REACTOR VESSEL.............................
3/49-12 3/4.9.11 WATER LEVEL / BORON CONCENTRATION - STORAGE P00L...........
3/49-13 3/4.9.12 FUEL HANDLING BUILDING EXHAUST FILTER PLENUMS............
3/4 9-14 BYRON - UNITS 1 & 2 XII AMENDMENT NO. 94
BASES 5ECTION EA_Q1 3/4.9.6 REFUELING MACHINE.........................................
B 3/4 9-2 3/4.9.7 CRANE 1 RAVEL - SPENT FUEL STORAGE FACILITY................
B 3/4 9-2 3/4.9.8 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 YESSEL..............................
B 3/4 9-3 3/4.9.11 WATER LEVEL / BORON CONCENTRATION - STORAGE P00L............
B 3/4 9-3 3/4.9.12 FUEL HANDLING BUILDING EXHAUST FILTER PLENUM..............
B 3/4 9-3 3/4.10 SPECIAL TEST EXCEPTION 9 3/4.10.1 SHUTDOWN MARGIN...........................................
B 3/410-1 3/4.10.2 GROUP HEIGHf, INSERTION, AND POWER DISTRIBUTION LIMITS....
B 3/410-1 3/4.10.3 PHYSICS TESTS.............................................
B 3/410-1 3/4.10.4 REACTOR COOLANT L00PS.....................................
B 3/410-1 1
3/4.10. 5 POSITION INDICATION SYSTEM - SHUTD0WN.....................
B 3/410-1 3/4.11 RADI0 ACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS Liquid Holdup Tanks..................................
B 3/411-1 3/4.11.2 GASEOUS EFFLUENTS Explosive Gas Mixture.................................
B 3/411-2 Gas Decay Tanks...................................
B3/411-2 1
l i
BYRON - UNITS 1 & 2 XVII AMEHDHENT NO. 94
DESIGN FEATURES SECTION EAGE 5.1 SITE 5.1.1 EXCLUSION AREA..............................................
5-1 5.1.2 LOW POPULATION Z0NE.........................................
5-1 5.1.3 MAP DEFINING UNRESTRICTED AREAS AND SITE BOUNDARY FOR RADI0 ACTIVE GASEQUS AND LIQUID EFFLUENTS..............
5-1 FIGURE 5.1-1 EXCLUSION AREA AND UNRESTRICTED AREA FOR RADIOACTIVE GASEOUS AND LIQUID EFFLUENTS.............
5-2 FIGURE 5.1-2 LOW POPULATION 20NE..................................
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 R00 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 METEOROLOG1(Al TOWER LOCATION.................................
5-4 5.6 FDEL STORAGE 5.6.1 CRITICALITY...................,.............................
5-5 FIGURE 5.6-1 MINIMUM BURNUP VERSUS INITIAL ENRICHMENT FOR REGION 2 - ALL CELL CONFIGURATION STORAGE...........
5-5b FIGURE 5.6-2 MINIMUM BURNUP VERSUS INITIAL ENRICHMENT FOR REGION 2 00T-0F-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 COMPONENT CYCLIC OR TRANSIENT LIMIT...........................
5-5 TABLE 5.7-1 COMPONENT CYCLIC OR TRANSIENT LIMITS..................
5-6 BYRON - UNITS 1 & 2 XVIII AMENDMENT N0. 94
g]NISTRAT-IVECONTROLS SECTION EAqE 6.7 SAFETY LIMIT V10LATION........................................
6-15 6.8 PROCEDURES AND PR0SRAMS.......................................
6-16
- k. 9 REPORTING RE0t?"EMENTS........................................
6-20 6.9.1 ROUTINE REP 0RTS.............................................
6-20 Startup Report.......................................
6-?9 Annual Reports..............................................
6-20 Annual Radiological Environmental Operating Report..........
6-22 Annual Radioactive Ef fl uent Release Report..................
6-22 Monthly Operating Report....................................
6-22 Operating Limits Report.....................................
6-22 3
6.9.2 SPECIAL REP 0RTS.............................................
6-23 1.10 RECORD RETENTION.............................................
6-23 6.11 RADIATION PROTECTION PR0 GRAM.................................
6-24 6.12 HIGH RADIATION AREA..........................................
6-25 L
6.13 PROCESS CONTROL PROGRAM (PCP)................................
6-26 6.14 0FFSITE DOSE CfLCULATION MANUAL (00CM).......................
6-26 L
K BYRON - UNITS 1 & 2 XX AMENDMENT NO. 94 a_- - - _. ---- - -- - --
~~
= '
- 1. 0 DEFINITIONS 1L.
The defined terms of this section appear in capitalized type and are applicable j
throughout these Technical Specifications.
ACTION 1.1 ACTION shall.be that part of a Technical Specification which prescribes remedial measures required under designated conditions.
3 ACTUATION LOGIC TEST 1.2 An ACTUATION LOGIC TEST shall be the application of various simulated input combinations in conjunction with each possible interlock logic state and verification of the required logic output.
The ACTUATION LOGIC TEST shall include a continuity check, as a minimum, of output devices.
ANALOG CHANNEL OPERATIONAL TEST 1.3 An ANALOG CHANNEL OPERATIONAL TEST shall be the injection of a simulated si n al into the channel as close to the sensor as practicable to verify
_ Li BILITY of alars, interlock and/or trip functions.
The ANALOG CHANNEL Oh
'IONAL TEST shall include adjustments, as necessary, of the alarm, inter-lock and/or Trip Setpoints such that the Setpoints are within the required c
i range and accuracy.
AXIAL FLUX DIFFERENCE 1.4 AXIAL FLUX DIFFERENCE shall be the difference in normalized flux sienals between the top and bottom halves of a two section excore neutron detector.
CHANNEL CALIBRATION J
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.
i i
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 indications and/or status derived from independent instrument channels measuring the same parameter.
BYRON - UNITS 1 & 2 1-1 L
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DEFINITIONS-CONTAINMENT INTEGRITY 1.7 CCNTAINMENT INTEGRITY shall exist when:
a.
All penetrations required to be closed during accident conditions
(*e 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, 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.
s I
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 4
process data to verify OPERABILITY of alarm and/or trip functions.
BYRON - UNITS 1 & 2 1-2 AMENDMENT NO. 94 7
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REFUELING OPERATIONS 3/4.9.11 WATER LEVEL / BORON CONCENTRATION - STORAGE POOL LINITING 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 4
1 -
concentration.of the water in the storage pool shall be maintained at greater i
than er equal to 2000 ppe.
I APPLICABILITY: 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 />.
j b.
With the boron concentration requirements of the above specification not satisfied, 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 t
c.
The provisions of Specification 3.0.3 are not applicable.
SURVEILLANCE RE0VIREMENTS 4.9.11 The water level in the storage pool 'shall be determined to be at least 2
its minimum required depth at least once per 7 days when irradiated fuel assemblies are in the fuel storage pool.
4.9.11.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 />.
l;
~ BYRON - UNITS 1 & 2 3/4 9-13 AMENDMENT NO. 94
REFUELING' OPERATIONS
^
BASES 3/4.9.8 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 ensures that a single failure of the operating RHR loop will not result in a complete loss of RHR capability.
Hith the reactor vessel head removed and at least 23 feet of water above the ceactor 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 SYSTEM The OPERABILITY of this system ensures that the containment purge penetrations will be automatically isolated upon detection of high 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.
143 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 u soluble boron concentration in the storage pool water ensure the spent fuel rack k will be maintained less than or equal to 0.h witha95-percentconfidence,Tevel, t
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-charge to the atmosphere. The OPERABILITY 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.
BYRON - UNITS 1 & 2 B 3/4 9-3 AMENDMENT NO. 94
DESIGN FEATURES
~
5.3 REACTOR CORE FUEL ASSEMBLIES 5.3.1 The core shall contain 193 fuel assemblies with each fuel assembly containing 264 fuel rods clad with Zircaloy-4 or ZIRLO, except that limited substitution 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 reload analysis.
Each fuel rod shall have a nominal active fuel length of 144 inches. The initial core loading shall have a maximum enrichment of less 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 determined to be acceptable for storage in either the spent fuel pool or the new fuel vault.
Such acceptance criteria shall be based on the results of the " Byron and Braidwood Spent Fuel Rack Criticality Analysis Using Soluble Boron Credit," May 1997, CAC-97-162 and
" Criticality Analysis of the Byron /Braidwood Fresh Fuel Racks," June 1989.
j CONTROL R00 ASSEMBLIES 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. All control rods shall be hafnium, silver-indium-cadmium, or a mixture of both types. All control rods shall be clad with stainless steel tubing.
5.4 REACTOR COOLANT SYSTEM 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 applicable Surveillance Requirements, b.
For a pressure of 2485 psig, and c.
For a temperattre of 650*F, except for the pressurizer which is 680*F.
VOLUME 5.4.2 Tne total water and steam volume of the Reactor Coolant System 1s 12,257 cubic feet at a nominal T,y of 588.4*F.
5.5 METEOROLOGICAL TOWER LOCATION 5.5.1 The meteorological tower shall be located as shown on Figure 5.1-1.
BYRON - UNITS 1 & 2 5-4 Amendment No. 94
Q{jl2LfEATURES
- 5.6 FUEL STORAGE CRITICALITY 5.6.1.1-The spent fuel storage racks are designed and shall be maintained 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; c.
Ak
< 0.95 if fully flooded with water borated to 550 ppm, which inclNdes 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; 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, cs described in the " Byron and Braidwood Spent Fuel Rack Criticality Analysis Using Soluble Boron Credit," May 1997, CAC-97-162, which may 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 applicable, 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 Boror. 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 storage 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 2 inches.
BYRON - UNITS 1 & 2 5-5 Amendment No. 94
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Note:
The use oflinear iaterpolation between the minimum burnups is acceptable.
FIGURE 5.6-1 MINIMUM BURNUP VERSUS INITIAL ENRICHMENT FOR REGION 2 ALL CELL CONFIGURATION STORAGE BYRON - UNITS 1 & 2 5-Sb AMENDPEhT NO. 94
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FIGURE 5.6-2 MINIMUM BURNUP VERSUS INITIAL ENRICHMENT FOR REGION 2 3-OUT-OFJ CHECKERBOARD CONFIGURATION BTRON - UN113 1 & 2 5-Sc AMEh M T NO.94
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Note:
The use of hnear interpolation between the minimum burnups is acceptable.
FIGURE 5.6-3 MINIMUM BURNUP VERSUS INITIAL ENRICHMENT FOR REGION 2 2-OUT-OF-4 CHECKERBOARD CONFIGURATION 3
Brnow - UNITS'1 a 2 5-9 m m, g
=
, ADMINISTRAT'VE CONTROLS l
o 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 CodeofFederalRegulations,thefollowingrecordsshallberetainedforaf least the minimum period indicated.
6.10.1 The following records shall be retained for at least 5 years:
a.
Records and logs of unit operatica 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 calibrations requiredbytheseTechnicalSpecifications; e.
Records of changes made to the procedures required by Specification 6.8;
)
f.
Records of radioactive shipments; g.
Records of sealed source 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:
i 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; 4
BYRON - UNITS 1 & 2 6-23 AMENDMENT NO.94
.