ML20217K446

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Proposed Tech Specs Re Boron Credit in SFP
ML20217K446
Person / Time
Site: Byron, Braidwood  Constellation icon.png
Issue date: 10/21/1997
From:
COMMONWEALTH EDISON CO.
To:
Shared Package
ML20217K444 List:
References
NUDOCS 9710280132
Download: ML20217K446 (14)
v • d • e


Text

. _ . . . _ . _ _ _ _ _ _

ATTACHMENT B 1 MARKED UP PAGES FOR SUPPLEMENT TO PROPOSED CHANGES TO APPENDIX A, TECHNICAL f"ECIFICATIONS, OF FACILITY OPERATINO LICENSES i NPF-37 and NPF-66 i

BORON CREDIT IN THE SPENT FUEL POOL f BYRON STATION UNITS 1 & 2 REVISED PAGES:

Table of Contents I

TableofContents XX i 1-2 5-4 6-23 f

a 1

9710290132 971021 PDR ADOCK 05000454 p PDR

s' .

y DEFINITIONS SECTION' EMig 1.0 DEFINITIONS 1.1 ACTI0N........................................................ 1-1 1.2 ACTUATION LOGIC TES1........................................... 1-1 1.3 ANALOG CHANNEL OPERATIONAL TEST............................... 1-1 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 CONTPOLLED LEAKAGE............................................ 1-2 1.9 CORE ALTERATION............................................'... 1-2 1."9h' YRON AND BRAI

' '& ebt A w u 5........................................... -2 1.10' DICITAL CHANNEL ;0PERATIONAL TEST. . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-2 1.11 DOSE EQUIVALENT I-131........................................ 1-2a 1.12 E-AVERAGE DISINTEGRATION ENERGY.............................. 1-3 3 1.13 ENGINEERED SAFETY FEATURES RESPONSE TIME..................... 1-3 1.14 FREQUENCY N0TATION........................................... 1-3 1.15 IDENTIFIED LEAKAGE........................................... 1-3 1.15.a L .......................................................... 1-3 1.16 MASTER Rl:.AY TEST............................................ 1-3 1.17 MEMBER (S) 0F THE PUBLIC...................................... 1-3 1.18 0FFS ITE DOSE CALCULATION MNUAL. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-4 1.19 OPERABLE - OPERABI LIT Y . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-4 1.19.a OPERATING LIMITS REP 0AT..................................... 1-4 1.20 OPERATIONAL MODE - M00E...................................... 1-4 1.20.a P,.......................................................... 1-4 l 1.2i PuYSics TESTS................................................ 1-4 1.22 PRESSURE B0UNDARY. LEAKAGE.................................... 1-4 1.23 PROCESS LONTROL PR0 GRAM...................................... 1-5 1.'24 PURGE - PURGING................. ............................ 1-5 1.25 QUADRANT POWER TILT RATI0.................................... 1-5 1.26 RATED THERMAL P0WER.......................................... 1-5 1.27 REACTOR TRIP SYSTEM RESPONSE TIME............................ 1-5 1.25 REPORTABLE EVENT... .. ...................................... 1-5 BYRON - UNITS 1 & 2 I AMEN 0 MENT N0.-et-R -

= 3 ADMINISTRATIVE CONTROLS

, y, .

SECTION fAGE

  • 6.7- SAFETY LIMIT VIOLAT10N........................................ 6-15 6.8 PROCEDURES AND PR0 GRAMS....................................... 6-16 i

. 6. 9 REPORTING RE0UIRENENTS........................................ 6-20 6.9.1 ROUTINE REP 0RTS............................................. 6-20 Startup Report.............................................. 6-20 An n u a l Re po r t s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-20 Annual Radiological Environmental Operating Report. . . . . . . . . . 6 ,

Annual Radioactive Effluent Release Report. . . . . . . . . . . . . . . . . . 6-22 Monthly Operating Report....................................

6-22 Operating Limits Report................

.................... 6-22 -1 Crin caii v f- 9* ie af Byron and Braim.ood h6\ek. statta, r.2 6 Tw i e g e M a c k s . . . . . . . 7. . . . . . . . . . . . . . . . . . . .

6-23

-6.9.2- SPECIAL REP 0RTS............................................. 6-23 6.10 RECORD RETENT10N............................................. 6-23 6.11 RADIATION PROTECTION PR0 GRAM................................. 6-24 6.12 HIGH RADIATION AREA.......................................... 6 6.13 PROCESS CONTROL PROGRAM (PCP)................................ '6-26 6.14 0FFSITE DOSE CALCULATION MANUAL (00CM)....................... 6-26 BYRON - UNITS 1-& 2 XX AMENOMENT NO. w. 2 t j

. - . .. - - - - - ~

~; .

y <

s ., e i i

,, iDEFINITIONS _

_ e ~

b-3

  1. t CONTAINMENT INTEGRITY +

V 1.7- CONTAINMENT INTEGRITY shall exist when:  ; '

'9 ,

a.

Allpenetrationsrequiredtobeclosed'duringasidentconditions, are either: -

1) Capable of being closed by an OPERABLE containmant automatic isolation valve system, or -
2) Closed by manual valves, blind flanges, or deactivated automatic valves secured in their clased 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 3.6.1.2, and leakage rates are within the limits of Specification L-e.

The sealing bellows, mechanism or 0-rings) associatad with each penetration (e.g., welds, is OPERABLE.

CONTROLLED LEAKAG_E

. 1.8 = CONTROLLED coolant pump seals. LEAKAGE shall be that seal water flow supplied to the reactor CORE--ALTERATION

1. 9 CORE ALTERATION shall be the movement:or manipulation of any. component within the reactor vessel with the vessel he.ad removed and fuel in the vessel.

Suspension of CORE ALTERATION shall not preclude completion of movement of a component to a safe conservative position.

hNALYSIS OF BYRON AND BRAIDWOOD STATION FUEL STORAGE RA J

1.9.a The CRITICALITY 7 N AN p Gu M ATION FUEL STORAGE.

RACKS, is:a document that provi 2 -

- storage. These limit -2 lowable fuel enrichment for Specificati termined and sube 4

t0 .

Plant operation within these limits ccordances.

with d in-incLh uGi1*Soecifications.

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.

BYRON - UNIT 1 1-2 AMENDMENT NO. a49-(L.e 1

  • O DESIGN FEATURES 5.3 REACTOR CORE FUEL. ASSEMBLIES 4

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 enrichaeat of less than 3.20 weight percent U-235. Reload fuel shall be siellar 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 G;3Gi.iTY J."^1V5is GF siV;iG te, "-

196 x

~

CONTROL ROD ASSEMBLIES 5.3.2 The core shall contain 53 full-length and no part-length contiol rod assemblies. The full-length control- rod assemblies shall contain a nominal 142 inches of absorber mattrial. All control rods shall be hafnium, silver-indium-cadmium, or a mixture of both types. All centrol rods shall be clad with stainless _ steel tubing.

5.4 REACTOR COOLANT SYSTEM l

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 Sectior 5.2 of the UFSAR, with allowance for normal degradation pursuant to the l applicable Surveillance Requirements,
b. For a pressure of 2485 psig, and
c. For a temperature of 650*F, except for the pressurizer which is 680*F.

VOLUME 5.4.2 The tott.1 water and steam volume of the Reactor Coolant System is 12,257 cubic feet at a nominal T, of 588.4*F.

5.5 METEOROLOGICAL TOWER LOCATION 5.5.1 The meteorological tower shall be located as shown on Fi ura .1-1 s'N -

h 9kron ond ~3ra'sdWood h Ofb^ 'M S b sd& hon C"St>" %ieJ, 3

CAMN, onct " cvacalihof +he. 6 rm/ d

( k g <I W June . yQ ^

W BYRON - UNITS 1 & 2 I

5-4 Amendment No.4fW

ADMINISTRATIVE CONTROLS W LITY ANALYSIS OF BYRON AND BRAIDWD0D STATION FUEL STORAGE RACKS 6.9.1.10 Fue ichment limits for storage shall be establi and L documented in the LITY ANALYSIS OF BYRON A E BRAI STATION FUEL l STORAGE RACKS. The ana 1 methods used to de the maximum fuel enrichments shall be those p si reviewe approved by the NRC in -

! " CRITICALITY ANALYSIS OF BYRON AND STATION FUEL STORAGE RACKS." The l fuel enrichment limits for stora temined so that all applicable limits (e.g., subcriticalit he safety a s are met.

~

q) The CRITICAL LYSIS OF BYRON AND BRAIDWOOD STA EL STORAGE g) RACKS rep .ettf' 1 be provided upon issuance of any changes, to NRC Doc ntrol Desk, with copies to the ?)gional Administrator a nt Inspector. .

SPECIAL REPORTS l

6.9.2 Special reports shall be suheitted to the Regional Administrator of the NRC Regional Office within the time period specified for each report.

6.10 RECORD RETEN' ION In addition to the applicable record retention requirements of Title 10, Code of Federal Regulations, the following records shall be retained for at 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 operation covering time interval at each power level;
b. Records and logs of principal maintenance activities, inspections, repair and replacement of principal items of equipner.t 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 fissiori detector leak tests and results; aM
h. Records of annual physical inventory uf all sealed source material of record.

6.10.2 The following records shall be retained for the duration of the unit Operating License:

a. Reccrds 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; BYRON - UNITS 1 & 2 6-23 AMENDMENT NO. 30* -

g INSERT B 56.1.1 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,g < 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. A k.g < 0.95 if fully flooded with water borated to 550 ppm, 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;
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 1 racks;
e. New or spent assemblies with suflicient Integral Fuel Burnable Absorbers present in each fuel assembly, as 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, 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 Baron 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.

I

ATTACHMENT B-2 hWlKED UP PAGES FOR SUPPLEMENT TO PROPOSED CHANGES TO APPENDIX A, TECHNICAL SPECIFICATIONS, OF FACILITY OPERATING LICENSES NPF-72 and NPF-77 BORON CREDIT IN TIIE SPENT FUEL POOL BRAIDWOOD STATION UNITS 1 & 2 REVISED PAGES:

Table of Contents I Table of Contents XX l-2

5-4 6-23 6 . .

.;. g DEFINITIONS SECTION E&EE 1.0 DEFINITIONS 1.1 ACTI0N........................................................ 1-1 s

1.2 AC TUAT I 0ff LOG I C T E ST . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-1 1.3 ANALOG CHANNEL OPERATIONAL TEST............................... 1-1 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 ALTERATION............................................... 1-2 1.9.a CRIT!tAttTY .*?*.tvu s OF BYRON AND BRAIDWOOD STATIOK FUEL h RTORAGE "% XS...............................................

1.10 DIGITAL CHANNEL OPERATIONAL TEST.............................

1-2__

1-2 1

1.11 DOSE EQUIVALENT I-131........................................ 1-2 1.12 E-AVERAGE DISINTEGRATION ENERGY.............................. 1-3 1.13 ENGINEERED SAFETY FEATURES RESPONSE TINE..................... 1-3 1.14 FREQUENCY NOTATION........................................... 1-3 1.15 IDENTIFIED LEAKAGE........................................... 1-3 1.15.a L,.......................................................... 1-3 l 1.16 MASTER RELAY TEST............................................ 1-3 1.17 NENBER(S) 0F THE PUBLIC...................................... 1-3 1.18 0FFSITE DOSE CALCULATION MANUAL.............................. 1-4 1.19 OPERABLE - OPERABILITY....................................... 1-4 1.19.a OPERATING LINITS REPDRT..................................... 1-4 1.20 OPERATIONAL MODE - M00E...................................... 1-4 1.20.a P,.......................................................... 1-4 1.21 PHYSICS TESTS................................................ 1-4 1.22 PRESSU"F BOUNDARY LEAKAGE.................................... 1-4 1.23 PRGCESS CONTROL PR0 GRAM...................................... 1-5 1.24 PURGE - PURGING.............................................. 1-5 1.25 QUADRANT POWER TILT RATI0.................................... 1-5 1.26 P.ATED THERMAL P0WER.......................................... 1-5 1.27 REACTOR TRI P SYSTEM RESPONSE TINE. . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-5 1.28 REPORTABLE EVENT............................................. 1-5 BRAIDWOOD - UNITS 1 & 2 1 Amendment No.

t

___ __ __J

.* AIMINI S IMAI I YE LUM I MUD =

SECTION PEI 6.7 SAFETY LIMIT VIOLATION........................................ 6-15 6.8 PROCEDURES AND PROGR6%....................................... 6-16 i

(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 Monthly Operating Report.................................... 6-22 Operating Limits Report..................................... 6-22 Kffilii., M1,vsh of Byron and Braidwood h S t a tim,Je:1 St o .p Rac d . . . . . . . . . . . . . . . . . . . . . . . . . . _ .

6.9.2 SPECIAL REP 0RTS.............................................

6-23 0-zs

~ '

6.10 RECORD RETENTlQH............................................. 6-23 6.11 RADIATION PROTECTION PR0hRAM................................. 6-24 t

s 6.12 HIGH RADIATION AREA.......................................... 6-25 6.13 PROCESS CONTROL PROGRAM (PCP)................................ 6-26 6.14 0FFSITE DOSE CALCULATION MANUAL (0DCM)....................... 6-26 BRAIDWOOD - UNITS 1 & 2 XX AMENDMENT NO. 59

4

.,. - DEFINITIONS CONTAINMENT-INTEGRITY 1.7. CONTAINNENT 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,
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 raactor 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.

CTITicAt4ILANALYSIS OF BYRON AND BRAIDWOOD STATION FUEL STORAGE

1. 9a The CRITICALI BYRON AND BRAI N FUEL-STORAGE RACKS, is a document that provides - r- allowable fuel enrichment for storage. These limits s ermined aD1 accordance with

$ Specification . . Plant operation within these limits d in Ci pectfications.

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 simulatad 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, I-132, I-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."

e BRAIDWOOD UNITS 1 & 2 1-2 AMENDMENT NO.

7 , ----- m , , m_ ,

5.3 REACTOR CORE FUEL ASSEMBLIES

~

5.3.I' The core shall contain 193 fuel essemblies with each fuel assembly 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, ZlRLO, 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 l 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 criteribs shall be based on the results of the C TY ANALYSIS 1F-BYRGN ANL BRAIDWO0tMYATM TZL ';T^ZC 00X;. g[

CONTROL ROD ASSEMBLIES 5.3.2 The core shall contain 53 full-length and no part-length control rod assemblies. The full-1cngth 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 1

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 temperature 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 588.4 *F.

m 5.5 METEOROLOGICAL TOWER LOCATION 5.5.1 The meteorological tower shall be located as shown on Figure 5.1-1.

Q QlY l 2nY RAk. MO by Nlp lh bd W 0 hk h ^'hl0 ard ver,+iald fuel AmW Janc " y halys6

-U 1989- h*a-.f6pn/Brmdwoel P Amendment No.JM

,?

,ADMINI5TRATIVE CONTROLS IRITICALITY ANALYSIS OF BYRON AND BRAIDWOOD STATION FUEL STORAGE RACKS 6.9.1.10 -eririchment limits for storage shall be establi and documentedintheblQCALITYANALYSISOFBYRONANDBRAI STATION FUEL STORAGE RACKS. The analyti e the maximum fuel enrichments shall be thosepreviously (al methods used to deteapproved by the NRC in reviewed

" CRITICALITY ANALYSIS OF BYRON AN TATION FUEL STORAGE RACKS." The fuel enrichment limits for storage s determined so that all applicable p limits (e.g., suberiticality) o e safety a is are met.

J y The CRITICALITY RACKS report sh SIS OF BYRON AND BRAIDWOOD ST e provided upon issuance of any changes, FUEL STORAGE e NRC l c ntfoi Desk, with copies to the Regional Administrator an e l

Docume Residg nspector.

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 d

In addition to the applicable record retention requirements of Title 10, Code of Federal Regulations, the following records shall be retained for at f 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 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 REPORTA8LE EVENTS;
d. Records of surveillance activities, inspections, and calibrations required by these Technical Specifications;
e. Records of changes made to the procedures rc. quired 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 'nventory 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; BRAIDWOOD - UNITS 1 & 2 6-23 AMENDMENT NO.

m J

q, INSERT B 5.6.1.1 The spent fuel rtorage racks are designed and shall be maintained with:

a. ruel assemblies having a maximum initial U-235 enrichment of 5.0 weight percent;
b. A k.a < 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. A kon < 0.95 if fully flooded witn water borated to 550 ppm, 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;
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 1 racks;
e. New or spent assemblies with sufficient Integral Fuel Burnable Absorbers present in each fuel assembly, as described in the " Byron and Braidwood Spent Fuel Rack Criticality Andysis Using Soluble Boron Credit," May 1997, CAC-97-162, which may be allowed unrestricted storage in the Region 1 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, 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.

f

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