ML052800352

Tech Spec Pages for Amendment 236 Conversion of TS Section 4.0 to STS Format

ML052800352
Person / Time
Site:	Fort Calhoun
Issue date:	10/03/2005
From:	Division of Nuclear Materials Safety IV
To:
Wang A, NRR/DLPM, 415-1445
Shared Package
ML052850414	List: ML051680267 ML052800352
References
TAC MC6652
Download: ML052800352 (6)

Text

TECHNICAL SPECIFICATION TABLE OF CONTENTS (Continued) 2.13 DELETED 2.14 Engineered Safety Features System Initiation Instrumentation Settings 2.15 Instrumentation and Control Systems 2.16 River Level 2.17 Miscellaneous Radioactive Material Sources 2.18 DELETED 2.19 DELETED 2.20 Steam Generator Coolant Radioactivity 2.21 Post-Accident Monitoring Instrumentation 2.22 Toxic Gas Monitors 3.0 SURVEILLANCE REQUIREMENTS 3.1 Instrumentation and Control 3.2 Equipment and Sampling Tests 3.3 Reactor Coolant System and Other Components Subject to ASME Xi Boiler and Pressure Vessel Code Inspection and Testing Surveillance 3.4 DELETED 3.5 Containment Test 3.6 Safety Injection and Containment Cooling Systems Tests 3.7 Emergency Power System Periodic Tests 3.8 Main Steam Isolation Valves 3.9 Auxiliary Feedwater System 3.10 Reactor Core Parameters 3.11 DELETED 3.12 Radioactive Waste Disposal System 3.13 Radioactive Material Sources Surveillance 3.14 DELETED 3.15 DELETED 3.16 Residual Heat Removal System Integrity Testing 3.17 Steam Generator Tubes 4.0 DESIGN FEATURES 4.1 Site 4.2 Reactor Core 4.3 Fuel Storage TOC - Page 2 Amendment No. 11,27,32,38,43,16,54,60,81,86, 93,97,101,122,436,452,460,176,483, 211, 230, 236

TECHNICAL SPECIFICATION TABLE OF CONTENTS (Continued) 5.0 ADMINISTRATIVE CONTROLS 5.1 Responsibility 5.2 Organization 5.3 Facility Staff Qualifications 5.4 Training 5.5 Review and Audit 5.5.1 Plant Review Committee (PRC) 5.5.2 Safety Audit and Review Committee (SARO) 5.6 Reportable Event Action 5.7 Safety Limit Violation 5.8 Procedures 5.9 Reporting Requirements 5.9.1 Routine Reports 5.9.2 Reportable Events 5.9.3 Special Reports 5.9.4 Unique Reporting Requirements 5.9.5 Core Operating Limits Report 5.9.6 RCS Pressure-Temperature Limits Report (PTLR) 5.10 Record Retention 5.11 Radiation Protection Program 5.12 DELETED 5.13 Secondary Water Chemistry 5.14 Systems Integrity 5.15 Post-Accident Radiological Sampling and Monitoring 5.16 Radiological Effluents and Environmental Monitoring Programs 5.16.1 Radioactive Effluent Controls Program 5.16.2 Radiological Environmental Monitoring Program 5.17 Offsite Dose Calculation Manual (ODCM) 5.18 Process Control Program (PCP) 5.19 Containment Leakage Rate Testing Program 5.20 Technical Specification (TS) Bases Control Program 5.21 Containment Tendon Testing Program 6.0 INTERIM SPECIAL TECHNICAL SPECIFICATIONS 6.1 DELETED 6.2 DELETED 6.3 DELETED 6.4 DELETED TOC - Page 3 Amendment No. 32,34,43,51,55,57, 73,60,86,89,93,99,141,152,157,184,185, 221, 236

TECHNICAL SPECIFICATIONS 4.0 DESIGN FEATURES 4.1 Site The site for Fort Calhoun Station Unit No. 1 is in Washington County, Nebraska, on the west bank of the Missouri River and approximately nineteen miles north, northwest of the city of Omaha, Nebraska. The exclusion area, as defined in 10 CFR Part 100, Section 100.3(a), consists of approximately 1242 acres. The exclusion area boundary extent includes approximately 660 acres in Washington County, Nebraska, owned by the Omaha Public Power District (OPPD), and 582 acres in Harrison County, Iowa, on the east bank of the river directly opposite the facility, on which the District retains perpetual easement rights. The minimum exclusion area boundary point is located approximately at the 187.0 degree radial from the outer wall of the containment building and at a distance of 910 meters.

4.2 Reactor Core 4.2.1 Fuel Assemblies The reactor shall contain 133 fuel assemblies. Each assembly shall consist of a matrix of zircaloy or ZIRLO fuel rods with an initial composition of natural, depleted, or slightly enriched uranium dioxide (UO2) as fuel material. Limited substitutions of zirconium alloy or stainless steel filler rods for fuel rods, in accordance with approved applications of fuel rod configurations, may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff approved codes and methods and shown by tests or analyses to comply with all fuel safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in nonlimiting core regions.

4.2.2 Control Element Assemblies The reactor core shall contain 49 control element assemblies (CEAs). The control material shall be silver indium cadmium, boron carbide, or hafnium metal as approved by the NRC.

4.3 Fuel Storage 4.3.1 Criticality 4.3.1.1 The spent fuel storage racks are designed and shall be maintained with:

a. Fuel assemblies having a maximum U-235 enrichment of 4.5 weight percent,

b. keff 50.95 if fully flooded with unborated water, which includes an allowance for uncertainties as described in Section 9.5 of the USAR, 4.0 - Page 1 Amendment Nos. 20,36,409, 4-78, 236

TECHNICAL SPECIFICATIONS 4.0 DESIGN FEATURES (Continued)

c. A nominal 8.6 inch center to center distance between fuel assemblies placed in Region 2, the high density fuel storage racks,

d. A nominal 9.8 inches (East-West) by 10.3 inches (North South) center to center distances between fuel assemblies placed in Region 1,the low density fuel storage racks,

e. New or partially spent fuel assemblies with a discharge burnup in the uacceptable domain" of Figure 2-10 for "Region 2 Unrestricted" may be allowed unrestricted storage in any of the Region 2 fuel storage racks in compliance with Reference (1),

f. Partially spent fuel assemblies with a discharge burnup between the uacceptable domain" and "Peripheral Cells" of Figure 2-10 may be allowed unrestricted storage in the peripheral cells of the Region 2 fuel storage racks in compliance with Reference (1),

g. New or partially spent fuel assemblies with a discharge burnup in the "unacceptable domain" of Figure 2-10 will be stored in Region 1 in compliance with Reference (1).

4.3.1.2 The new fuel storage racks are designed and shall be maintained with:

a. Fuel assemblies having a maximum U-235 enrichment of 5.0 weight percent,

b. keff ' 0.95 if fully flooded with unborated water, which includes an allowance for uncertainties as described in Reference (2).

c. ke S 0.98 if moderated by aqueous foam, which includes an allowance for uncertainties as described in Reference (2).

d. A nominal 16 inch center to center distance between fuel assemblies placed in the storage racks.

References:

(1) Letter from R. Wharton (NRC) to T. Patterson (OPPD), Amendment 174 to Facility Operating License No. DPR-40, (TAC NO. M94789) Dated July 30, 1996, NRC-96-0126.

(2) Ft. Calhoun USAR, Reference 9.5-1 4.0 - Page 2 Amendment No. 236

TECHNICAL SPECIFICATIONS 4.0 DESIGN FEATURES (Continued) 4.3.2 Drainage The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 23 ft.

4.3.3 Capacity The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 1083 fuel assemblies.

4.0 - Page 3 Amendment No. 6-AT.

60n109,n47-8, 236

TECHNICAL SPECIFICATIONS TABLE 1-1 RPS LIMiTING SAFETY SYSTEM SETTINGS No. Reactor Trip Trip Setpoints 1 High Power Level (A) <109.0% of Rated Power 4-Pump Operation 2 Low Reactor Coolant Flow (B)(F) 4-Pump Operation >95% of 4 Pump Flow 3 Low Steam Generator Water Level 31.2% of Scale 4 Low Steam Generator Pressure (C) >500 psia 5 High Pressurizer Pressure <2400 psia 6 Thermal Margin/Low Pressure (B)(F) 1750 psia to 2400 psia (depending on the reactor coolant temperature as shown in the Thermal Margin/Low Pressure 4 Pump Operation Figure provided in the COLR) 7 High Containment Pressure (D) <5 psig 8 Axial Power Distribution (E) (as shown in the Axial Power Distribution for 4 Pump Operation Figure provided in the COLR) 9 Steam Generator Differential Pressure <135 psid 1.0- Page 11 Amendment No. 7,32,47,77,92,14,170, 910, 236