ML050600237

Technical Specification Pages Re Minimum Critical Power Ratio Safety Limit

ML050600237
Person / Time
Site:	Susquehanna
Issue date:	02/28/2005
From:	NRC/NRR/DLPM
To:
References
Download: ML050600237 (4)

Text

PPL Rev. 0 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 2 785 psig and core flow

> 10 million Ibm/hr:

MCPR shall be 2 1.09 for two recirculation loop operation or > 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 5 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.

SUSUQUEHANNA - UNIT 2 TS 2.0-1 Amendmqnt1 VI 104, 14 101, 194

PPL Rev. 2 Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 COLR (continued) core thermal power level may not exceed the originally approved RTP of 3441 MWt, but the value of 3510 MWt (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 LEFM/Tm 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.

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)(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), "Exxon Nuclear Methodology for Boiling Water Reactors," Exxon Nuclear Company.

6.

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

7.

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

8.

EMF-2292(P)(A), uATRIUM Th'-I10: Appendix K Spray Heat Transfer Coefficients," Siemens Power Corporation.

9.

XN-NF-84-105(P)(A), uXCOBRA-T: A Computer Code for BWR Transient Thermal-Hydraulic Core Analysis," Exxon Nuclear Company.

(continued)

SUSQUEHANNA - UNIT 2 TS / 5.0-22 Arpendment 10

. tA9, 1,1X4 194

PPL Rev. 2 Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 COLR (continued)

10.

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

11.

ANF-913(P)(A), UCOTRANSA2: A Computer Program for Boiling Water Reactor Transient Analyses," 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-1997(P)(A), uANFB-10 Critical Power Correlation", Siemens Power Corporation.

15.

EMF-CC-074(P)(A), "BWR Stability Analysis - Assessment of STAIF with Input from MICROBURN-B2," Siemens Power Corporation.

16.

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

17.

Caldon, Inc., "TOPICAL REPORT: Improving Thermal Power Accuracy and Plant Safety while Increasing Operating Power Level Using the LEFTMBtm System," Engineering Report - 80P.

18.

Caldon, Inc., "Supplement to Topical Report ER-80P: Basis for a Power Uprate with the LEFM4tNI or LEFM CheckPlusT m System," Engineering Report ER-1 60P.

19.

NEDO-32465-A, "BWROG Reactor Core Stability Detect and Suppress Solutions Licensing Basis Methodology for Reload Applications."

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.

SUSQUEHANNA - UNIT 2 TS / 5.0-23 Amendm t 1 1

194

PPL Rev. 2 Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.6 EDG Failures Report If an individual emergency diesel generator (EDG) experiences four or more valid failures in the last 25 demands, these failures and any nonvalid failures experienced by that EDG in that time period shall be reported within 30 days.

Reports on EDG failures shall include the information recommended in Regulatory Guide 1.9, Revision 3, Regulatory Position C.4.

PAM 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.

5.6.7 SUSQUEHANNA - UNIT 2 TS /5.0-23a Amendm nt 10 10,174, 1R9