ML040720488

TS Pages

ML040720488
Person / Time
Site:	Susquehanna
Issue date:	03/09/2004
From:	NRC/NRR/DLPM
To:
References
TAC MB9902
Download: ML040720488 (4)

Text

SLs 2.0 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs 2.1.1.1 With the reactor steam dome pressure < 785 psig or core flow < 10 million Ibm/hr:

THERMAL POWER shall be < 25% RTP.

2.1.1.2 With the reactor steam dome pressure > 785 psig and core flow > 10 million Ibm/hr:

MCPR shall be > 1.08 for two recirculation loop operation or 2 1.10 for single recirculation loop operation.

2.1.1.3 Reactor vessel water level shall be greater than the top of active irradiated fuel.

2.1.2 Reactor Coolant System Pressure SL Reactor steam dome pressure shall be < 1325 psig.

2.2 SL Violations With any SL violation, the following actions shall be completed within 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />s:

2.2.1 Restore compliance with all SLs; and 2.2.2 Insert all insertable control rods.

SUSQUEHANNA - UNIT 1 TS /2.0-1 Amendment 1,94 216

Design Features 4.0 4.0 DESIGN FEATURES 4.1 Site Location 4.1.1 Exclusion Area Boundaries The exclusion area shall be as shown in Figure 4.1-1.

4.1.2 Low Population Zone The low population zone shall be as shown in Figure 4.1-2.

4.2 Reactor Core 4.2.1 Fuel Assemblies The reactor shall contain 764 fuel assemblies. Each assembly shall consist of a matrix of Zircalloy fuel rods with an initial composition of depleted, natural, or slightly enriched uranium dioxide (UO2) as fuel material, and water rods or water channels. Limited substitutions of zirconium alloy filler rods for I

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 NRC staff approved codes and methods and have been shown by tests or analyses to comply with all safety design bases. A limited number of lead use 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/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:

(continued)

SUSQUEHANNA - UNIT 1 TS f4.0-1 Amendment 1 216

Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5.6.5 COLR (continued)

(102% of 3441 MWt), remains the initial power level for the bounding licensing analysis.

Future revisions of approved analytical methods listed in this Technical Specification that are currently referenced to 102% of rated thermal power (3510 MWt) shall include reference that the licensed RTP is actually 3489 MWt. The revisions shall document that the licensing analysis performed at 3510 MWt bounds operation at the RTP of 3489 MWt so long as the LEFMVTM system is used as the feedwater flow measurement input into the core thermal power calculation.

The approved analytical methods are described in the following documents, the approved version(s) of which are specified in the COLR.

1. PL-NF-90-001-A, Application of Reactor Analysis Methods for BWR Design and Analysis."

2. XN-NF-80-19(P)(A), nExxon Nuclear Methodology for Boiling Water Reactors," Exxon Nuclear Company, Inc.

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

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

5. NE-092-001A, "Licensing Topical Report for Power Uprate With Increased Core Flow," Pennsylvania Power & Light Company.

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

(continued)

SUSQUEHANNA - UNIT 1 TS /5.0-22 emengm t17 t 216 886, >4, 409

PPL Rev. 1 Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 COLR (continued)

7. ANF-91-048(P)(A), "Advanced Nuclear Fuels Corporation Methodology I for Boiling Water Reactors EXEM BWR Evaluation Model."

8. XN-NF-79-71 (P)(A), Exxon Nuclear Plant Transient Methodology for l Boiling Water Reactors."

9. EMF-1997(P)(A), "ANFB-10 Critical PowerCorrelation." l

10. Caldon, Inc., 'TOPICAL REPORT: Improving Thermal Power Accuracy l and Plant Safety While Increasing Operating Power Level Using the LEFMVTM System," Engineering Report - 80P.

11. Caldon, Inc., "Supplement to Topical Report ER-80P: Basis for a Power l Uprate with the LEFMVTM or LEFM CheckPlusTM System," Engineering Report ER -160P.

12. EMF-85-74(P), "RODEX 2A (BWR) Fuel Rod Thermal-Mechanical l Evaluation Model."

13. EMF-CC-074(P)(A), Volume 4, UBWR Stability Analysis: Assessment of l STAIF with Input from MICROBURN-B2."

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

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.

(continued)

SUSQUEHANNA - UNIT 1 TS / 5.0-23 A endment 8 216

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