View source for RS-05-117, Revision of License Amendment Request for Activation of the Trip Outputs of the Oscillation Power Range Monitor System

{{Adams
| number = ML052640024
| issue date = 09/20/2005
| title = Revision of License Amendment Request for Activation of the Trip Outputs of the Oscillation Power Range Monitor System
| author name = Simpson P
| author affiliation = Exelon Generation Co, LLC, Exelon Nuclear
| addressee name = 
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000373, 05000374
| license number = NPF-011, NPF-018
| contact person = 
| case reference number = RS-05-117
| document type = Letter type:RS
| page count = 5
| project = 
| stage = Request
}}

=Text=
{{#Wiki_filter:Exelon GeTieT
          "-l 4300VOI VAmpla I L 10 CFR 50.90 RS-05-117 September 20, 2005 U . S. Nuclear Regulatory Commission ATTN : Document Control Desk Washington, DC 20555-0001 LaSalle County Station, Units 1 and 2 Facility Operating License Nos. NPF-1 1 and NPF-1 8 NRC Docket Nos. 50-373 and 50-374

==Subject:==
Revision of License Amendment Request for Activation of the Trip Outputs of the Oscillation Power Range Monitor System

==Reference:==
Letter from Keith R. Jury (Exelon Generation Company, LLC) to U. S . NRC, "License Amendment Request Activation of the Trip Outputs of the Oscillation Power Range Monitor System," dated April 30, 2004 In the referenced letter, Exelon Generation Company, LLC (EGC) requested changes to the Technical Specifications (TS), Appendix A, of Facility Operating License Nos. NPF-1 1 and NPF-18 for LaSalle County Station (LCS), Units 1 and 2 . The proposed changes incorporate into the TS the Oscillation Power Range Monitor (OPRM) instrumentation .
In a teleconference on June 6, 2005, EGC and the NRC discussed the TS Surveillance Requirement (SR) to calibrate the local power range monitors (LPRMs) associated with the OPRMs. This SR was included in the sample TS provided in "Generic Topical Report for the ABB Option III Oscillation Power Range Monitor (OPRM)," CENPD-400-P, Revision 1, approved by the NRC in 1995 ; however, the SR was omitted from the referenced submittal since the LPRM calibration SR was already required as part of TS 3.3.1 .1 . As a result of this teleconference, EGC is submitting a revised TS page 3 .3.1 .3-3 to add the LPRM calibration SR to the OPRM TS 3.3 .1 .3.
EGC has reviewed the information supporting a finding of no significant hazards consideration or the environmental assessment that was previously provided to the NRC in Attachment 1 of the referenced letter. The addition of the LPRM calibration SR does not affect the bases for concluding that the proposed license amendment does not involve a significant hazards consideration.
The attachment to this letter provides revised TS pages 3.3.1 .3-3 and 5.6-3 for LCS. The addition of the LPRM calibration SR to TS 3.3.1 .3 results in an editorial change to TS page 5.6-3 in order to maintain consistency with the SR numbering . The revised TS pages supersede those provided in the referenced letter .

U . S . Nuclear Regulatory Commission September 20, 2005 Page 2 The proposed change has been reviewed by the Plant Operations Review Committee, and has received Nuclear Safety Review Board Chairman concurrence in accordance with the Quality Assurance Program .
If you have any questions, please contact Mr. David Gullott at (630) 657-2819 .
I declare under penalty of perjury that the foregoing is true and correct. Executed on the 20th day of September 2005 .
Patrick R. Simpson Manager - Licensing Attachment - LCS TS Pages 3 .3.1 .3-3 and 5.6-3

Attachment LCS TS Pages 3.3 . 1 .3-3 and 5 .6-3

OPRM Instrumentation 3 .3 .1 .3 SURVEILLANCE     REQUIREMENTS
------------------------------------- NOTE -------------------------------------
When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided the OPRM maintains trip capability .
SURVEILLANCE                                FREQUENCY SR  3 .3 .1 .3 .1  Perform CHANNEL FUNCTIONAL TEST .                184 days SR  3 .3 .1 .3 .2  Calibrate the local power range monitors .       1000 effective full power hours SR  3 .3 .1 .3 .3  ___________________NOTE-___________________
Neutron detectors are excluded .
Perform CHANNEL CALIBRATION . The setpoints   24 months for the trip function shall be as specified in the COLR .
SR  3 .3 .1 .3 .4  Perform LOGIC SYSTEM FUNCTIONAL TEST .           24 months SR  3 .3 .1 .3 .5  Verify OPRM is not bypassed when THERMAL         24 months POWER is ? 28 .6% RTP and recirculation drive flow is < 60% of rated recirculation drive flow .
SR  3 .3 .1 .3 .6  ___________________NOTE-___________________
Neutron detectors are excluded .
Verify the RPS RESPONSE TIME is within           24 months on a limits .                                         STAGGERED TEST BASIS LaSalle 1 and 2                         3 .3 .1 .3-3                     Amendment No .

Reporting Requirements 5 .6 5 .6    Reporting Requirements 5 .6 .5        CORE OPERATING LIMITS REPORT (COLR)        (continued)
: 4. The Rod Block Monitor Upscale Instrumentation Setpoint for the Rod Block Monitor-Upscale Function Allowable Value for Specification 3 .3 .2 .1 .
The OPRM setpoints for the trip function for SR 3 .3 .1 .3 .3 .
: 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 :
ANF-1125(P)(A), "ANFB Critical     Power Correlation ."
: 2. Letter, Ashok C . Thadani (NRC) to R .A . Copeland (SPQ, "Acceptance for Referencing of ULTRAFLOW" Spacer on 9x9-IX/X BWR Design," July 28, 1993 .
: 3. XN-NF-524(P)(A), "ANF Critical    Power Methodology for Boiling Water Reactors ."
ANF-913(P)(A), - COTRANSA 2 :   A Computer Program for Boiling Water Reactor Transient Analysis ."
: 5. ANF-CC-33(P)(A), "HUXY :  A Generalized MulO rod Heatup Code with 10 CFR 50, Appendix K Heatup Option ."
: 6. XN-NF-80-19(P)(A), "Advanced Nuclear Fuel     Methodology for Boiling Water Reactors ."
: 7. XN-NF-85-67(P)(A), "Generic Mechanical     Design for Exxon Nuclear Jet Pump BWR Reload Fuel ."
ANF-89-014(P)(A), "ANF Corporation Generic Mechanical Design for ANF Corporation 9x9-IX and 9x9-9X BWR Reload Fuel ."
: 9. EMF-CC-074(P)(A), Volume 4 - "BWR Stability Analysis :
Assessment of STAIF with input from MICKBURYB2 ."
10 . XN-NF-81-58(P)(A), "RODEX2 Fuel     Rod Thermal-Mechanical Response Evaluation Model ."
11 . XN-NF-84-105(P)(A), - XCOBRA-T :    A Computer Code for BWR Transient Thermal-Hydraulic Core Analysis ."
(continued)
LaSalle 1 and 2                          5 .6-3                        Amendment No . /}}