ML031350762

Technical Specifications (5 Pages), S107051

ML031350762
Person / Time
Site:	Columbia
Issue date:	05/12/2003
From:	Brian Benney NRC/NRR/DLPM/LPD4
To:	Parrish J Energy Northwest
Benney B, NRR/DLPM, 415-3764
Shared Package
ml031350003	List: ML031340744 ML031350762
References
TAC MB6319
Download: ML031350762 (5)

Text

Design Features 4.0 4.0 DESIGN FEATURES 4.1 Site Location 4.1.1 Site and Exclusion Area Boundaries The site area shall include the area enclosed by the exclusion area plus the plant property lines that fall outside the exclusion area, as shown in Figure 4.1-1.

The exclusion area boundary is a circle with its center at the reactor and a radius of 1950 meters.

4.1.2 Low PoDulation Zone The low population zone is all-the la'nd within a circle with its center at the reactor and a radius of 4827 meters.

4.2 Reactor Core 4.2.1 Fuel Assemblies The reactor shall 'contain 764 fuel assemblies.

Each assembly shall consist of a matrix of Zircaloy clad fuel rods with an initial composition'of depleted, natural, or slightly enriched uranium' dioxide (UO) as fuel material and water rods or channels.

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 desig-ns that have been analyzed with applicable NRC taff approved codes and methods and shown by tests or analyses to 'comply with;all' safety design bases.

A limited number of lead fuel'assemblies that have not completed representative testing-may be'placed in nonlimiting core regions.

4.2.2 Control Rod Assemblies The reactor core shall contain 185 cruciform shaped control rod assemblies.

The control material shall be boron carbide and hafnium metal as approved by the NRC.

(continued)

Columbia Generating Station 4 0- 1 Amendment No. 49. 169,185

Design Features Design eatures 4.0 4.0 DESIGN FEATURES (continued) 4.3 Fuel Storage 4.3.1 Criticalitv 4.3.1.1 The spent fuel storage racks are designed and shall be maintained with:

a. keff < 0.95 if fully flooded with unborated water, which includes an allowance for uncertainties as described in Section 9.1.2 of the FSAR; and

b. A nominal 6.5 inch center to center distance between fuel assemblies placed in the storage racks.

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

a. ke,f

< 0.95 if fully flooded with unborated water, which includes an allowance for uncertainties as described in Section 9.1.1 of the FSAR; and

b. A maximum of 60 new fuel assemblies stored in the new fuel storage racks, arranged.in-6 spatially separated zones.

Within a storage zone, the nominal center-to-center distance between cells for storing fuel assemblies is 14 inches.

The nominal center-to-center distance between cells for storing fuel assemblies in adjacent zones is 37 inches.

Design features relied upon to spatially limit the placement of fuel bundles within the new fuel vault are required to be installed prior to placement of new fuel bundles in the vault.

4.3.2 Drainage The spent fuel-storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 83 ft 1.25 inches.

4.3.3 Capacity The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more.than 2658 fuel.

assemblies.

Columbia Generating Station 4.0- 2 Amendment No. 149.i63 169l

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR)

(continued)

b.

The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC. specifically those described in the following documents:

1.

XN-NF-81-58(P)(A),

RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Model." Exxon Nuclear Company

2.

XN-NF-85-67(P)(A), "Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel," Exxon Nuclear Company

3.

EMF-85-74(P) Supplement 1(P)(A) and Supplement 2(P)(A),

"RODEX2A (BWR) Fuel Rod Thermal-Mechanical Evaluation Model," Siemens Power Corporation

4.

ANF-89-98(P)(A)., "Generic Mechanical Design Criteria for BWR Fuel Designs." Advanced Nuclear Fuels Corporation

5.

XN-NF-80-19(P)(A) Volume 1, "Exxon Nuclear Methodology for Boiling Water Reactors - Neutronic Methods for Design and Analysis." Exxon Nuclear Company

6.

XN-NF-80-19(P)(A) Volume 4. "Exxon Nuclear Methodology for Boiling Water Reactors: Application of the ENC Methodology to BWR Reloads." Exxon Nuclear Company

7.

EMF-2158(P)(A), "Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4/MICROBURN-B2," Siemens Power Corporation

8.

XN-NF-80-19(P)(A) Volume 3,:'Exxon Nuclear Methodology for Boiling Water Reactors, THERMEX: Thermal Limits Methodology Summary Description," Exxon Nuclear Company

9.

XN-NF-84-105(P)(A) Volume 1 "XCOBRA-T: A Computer Code for BWR Transient Thermal-Hydraulic Core Analysis,"

Exxon Nuclear Company (continued)

Columbia Generating Station 5.6-3 Amendment No. 149.1'4.158,169,1'1,185

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR)

(continued)

10.

ANF-524(P)(A), "ANF Critical Power Methodology for Boiling-Water Reactors," Advanced Nuclear Fuels Corporation

11.

ANF-913(P)(A) Volume 1 "COTRANSA2: A Computer Program for Boiling Water Reactor Transient Analysis," Advanced Nuclear Fuels Corporation

12.

ANF-1358(P)(A) "The Loss of Feedwater Heating Transient in Boiling Water Reactors," Advanced Nuclear Fuels Corporation

13.

EMF-2209(P)(A), "SPCB Critical Power Correlation."

Siemens Power Corporation

14.

EMF-2245(P)(A), "Application of Siemens Power Corporation's Critical Power Correlations to Co-Resident Fuel," Siemens Power Corporation

15.

EMF-2361(P)(A), "EXEM BWR-2000 ECCS Evaluation Model."

-Framatome ANP Richland

16.

EMF-2292(P)(A), "ATRIUMTM -10: Appendix K Spray Heat Transfer Coefficients," Siemens Power Corporation

17.

EMF-CC-074(P)(A) Volume 4, BWR Stability Analysis-Assessment of STAIF with Input from MICROBURN-B2,"

Siemens Power Corporation

18.

CENPD-300-P-A, "Reference Safety Report for Boiling Water Reactor Reload Fuel," ABB Combustion Engineering Nuclear Operations

19.

NEDO-32465-A, "B'WR Owners' Group Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology and Reload Applications" (continued) 5.6-4 Amendment No. 149n-4r4-+/-E58-4-69--t7s,185 Columbia Generating Station

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR)

(continued)

c.

The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits.

core thermal hydraulic limits., Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM. transient analysis limits, and accident analysis limits) of the safety analysis are met.

d.

The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

5.6.6 Post Accident Monitorina (PAM) Instrumentation Report When a report is required by Condition B or F of LCO 3.3.3.1, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days.

The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

41~~~~~~~~~~~~~~~~~~~~~~~~~~~

Amendment No.

49.1547-j-&S,183 Columbia2 Generating Station

5. 6-5