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License Amendment Request to Revise Operating License and Technical Specifications for Measurement Uncertainty Recapture Power Uprate
	ML16183A365
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Site:	Columbia
Issue date:	06/28/2016
From:	Javorik A; Energy Northwest
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
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ML16187A023	List: GO2-16-096, License Amendment Request to Revise Operating License and Technical Specifications for Measurement Uncertainty Recapture Power Uprate;
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Withhold under 10 CFR 2.390. Enclosures 7, 10, and 11 contain Proprietary Information. 5 contains Security-Related Information Alex L. Javorik Vice President, Engineering P.O. Box 968, Mail Drop PE04 Richland, WA 99352-0968 Ph. 509-377-8555 F. 509-377-2354 aljavorik@energy-northwest.com 10 CFR 50.90

-XQH

GO2-16-096 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001

Subject:

COLUMBIA GENERATING STATION, DOCKET NO. 50-397 LICENSE AMENDMENT REQUEST TO REVISE OPERATING LICENSE AND TECHNICAL SPECIFICATIONS FOR MEASUREMENT UNCERTAINTY RECAPTURE (MUR) POWER UPRATE

Reference:

Nuclear Regulatory Commission (NRC) Regulatory Issue Summary (RIS) 2002-03, Guidance on the Content of Measurement Uncertainty Recapture Power Uprate Applications, dated January 31, 2002

Dear Sir or Madam:

Pursuant to 10 CFR 50.90, Application for Amendment of License or Construction Permit, and 10 CFR 50, Appendix K, ECCS Evaluation Models, Energy Northwest hereby requests a license amendment to revise the Columbia Generating Station (Columbia) Renewed Facility Operating License (OL) NPF-21 and Technical Specifications (TS). Specifically, the proposed changes revise the OL and TS to implement an increase in rated thermal power from the current licensed thermal power (CLTP) of 3486 megawatts thermal (MWt) to 3544 MWt.

The proposed changes are based on reduced uncertainty in the feedwater flow and temperature measurement that reduces the total power level measurement uncertainty, which is achieved by utilizing Cameron International (formerly Caldon) CheckPlus' Leading Edge Flow Meter (LEFM) ultrasonic flow measurement instrumentation. The LEFM instrumentation was installed at Columbia during the spring 2015 refueling outage.

The content of this request is in accordance with the guidance contained in NRC RIS 2002-03. Energy Northwest has only proposed those OL and TS changes that are required to implement the increased power level. Additionally, Energy Northwest has reviewed the requests for additional information (RAI) from the facilities identified in Enclosure 1, Section 4.2, Precedents, and has included information within the body of the submittal to address the general topics of those requests.

This submittal contains the following

Enclosures:

License Amendment Request to Revise Operating License and Technical Specifications for Measurement Uncertainty Recapture (MUR) Power Uprate Page 2 of 3 Description and Evaluation of the Proposed Change Markup of Existing Renewed Facility Operating License and Technical Specifications New Licensee Controlled Specification and Markup of Technical Specification Bases For Information Only Revised (Clean) Renewed Facility Operating License and Technical Specification Pages Regulatory Issue Summary (RIS) 2002-03 Cross Reference Summary of Regulatory Commitments General Electric-Hitachi (GEH) Report NEDC-33853P, Safety Analysis Report for Columbia Generating Station Thermal Power Optimization, Revision 0 (Proprietary Version) Affidavits from GEH and the Electric Power Research Institute (EPRI)

Supporting the Withholding of Information in Enclosure 7 from Public Disclosure GEH Nuclear Report NEDO-33853, Safety Analysis Report for Columbia Generating Station Thermal Power Optimization, Revision 0 (Non-Proprietary Version) 0 Cameron (Caldon) Document ER-1049, Bounding Uncertainty Analysis for Thermal Power Determination at Columbia Nuclear Generating Station Using the LEFM CheckPlus System, Revision 3 (Proprietary Version) 1 Cameron (Caldon) Document ER-1074, Meter Factor Calculation and Accuracy Assessment for Columbia Nuclear Generating Station, Revision 0 (Proprietary Version) 2 Affidavits from Cameron International Corporation Supporting the Withholding of Information in Enclosures 10 and 11 from Public Disclosure 3 Columbia Calculation NE-02-15-08, Heat Balance Determination for Rated Thermal Power, Revision 0 4 LEFM Flowmeter Installation Drawings 5 Bonneville Power Administration Report, Columbia Generating Station Measurement Uncertainty Recapture Reactor Thermal Power Limit Uprate Study, dated June 22, 2016 (Security-Related Information)

Approval of the proposed amendment is requested by May 13, 2017, prior to the start of the 2017 Refueling Outage (RFO). If approved prior to the 2017 RFO, the instrument recalibrations required for implementation will occur during the 2017 RFO. If not, the amendment will be implemented within 120 days of approval.

License Amendment Request to Revise Operating License and Technical Specifications for Measurement Uncertainty Recapture (MUR) Power Uprate Page 3 of 3 In accordance with 10 CFR 50.91, "Notice for Public Comment; State Consultation," Energy Northwest is notifying the State of Washington of this amendment request by transmitting a copy of this letter and enclosures to the designated State Official.

In accordance with 10 CFR 2.390, "Public Inspections, Exemptions, Requests for Withholding,"

Columbia requests withholding from public disclosure Enclosures 7, 10, 11, and 15. Enclosure 7 contains information that is considered proprietary by GEH Nuclear Energy and the Electric Power Research Institute {EPRI). Affidavits supporting this request are provided in Enclosure 8 and a non-proprietary version of Enclosure 7 is provided in Enclosure 9. Enclosures 1O and 11 are considered proprietary by Cameron International Corporation. Affidavits supporting these requests are included in Enclosure 12. Non-proprietary versions of Enclosures 10 and 11 are not available. Enclosure 15 to this letter provides the grid study and contains information deemed by the Bonneville Power Administration to be security sensitive information related to critical infrastructure. Energy Northwest requests that Enclosure 15 be withheld from public disclosure in accordance with 10 CFR 2.390(d)(1 ). Upon removal of Enclosures 7, 10, 11, and 15, this letter is decontrolled.

Regulatory commitments associated with this submittal are identified in Enclosure 6.

If there are any questions or if additional information is needed, please contact Ms. L. L.

Williams, Licensing Supervisor, at 509-377-8148.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on this 21f'1,. day of .r~ ,2016.

Respectfully; A. L. Javorik Vice President, Engineering

Enclosures:

As stated cc: NRC Region IV Administrator NRC NRR Project Manager NRC Sr. Resident Inspector - 988C CD Sonoda - BPA 1399 (email)

WA Horin -Winston & Strawn (email)

RR Cowley- WDOH (email)

EFSECutc.wa.gov- EFSEC (email)

License Amendment Request to Revise Operating License and Technical Specifications for Measurement Uncertainty Recapture (MUR) Power Uprate Enclosure 1 Description and Evaluation of the Proposed Change

License Amendment Request to Revise Operating License and Technical Specifications for Measurement Uncertainty Recapture (MUR) Power Uprate Page 1 of 27 Description and Evaluation of the Proposed Change Contents 1.0 Summary Description 2.0 Detailed Description 2.1 Columbia Renewed Facility Operating License (OL) 2.2 Columbia TS 1.1, Definition of Rated Thermal Power (RTP) 2.3 Columbia TS 3.3.1.1, RPS Instrumentation (After Implementation of PRNM Upgrade) 2.4 Columbia TS 3.3.4.1, End of Cycle Recirculation Pump Trip (EOC-RPT)

Instrumentation 2.5 Columbia TS 3.3.6.1 Primary Containment Isolation Instrumentation 2.6 Columbia Licensee Controlled Specifications (LCS) Changes, New Section (Information Only) 2.7 TS Bases Changes (Information Only) 3.0 Technical Evaluation 3.1 Background and General Approach 3.2 LEFM Measurement and Core Thermal Power Uncertainty 3.3 Evaluation of Changes to Operating License and Technical Specifications 3.4 Additional Considerations 4.0 Regulatory Evaluation 4.1 Applicable Regulatory Requirements/Criteria 4.2 Precedents 4.3 No Significant Hazards Consideration 4.4 Conclusions 5.0 Environmental Consideration 6.0 References

License Amendment Request to Revise Operating License and Technical Specifications for Measurement Uncertainty Recapture (MUR) Power Uprate Page 2 of 27 1.0 Summary Description Pursuant to 10 CFR 50.90, Application for Amendment of License or Construction Permit, and 10 CFR 50, Appendix K, ECCS Evaluation Models, Energy Northwest hereby requests a license amendment to revise the Columbia Generating Station (Columbia) Renewed Facility Operating License (OL) No. NPF-21 and Technical Specifications (TS). Specifically, the proposed changes revise the OL and TS to implement an increase in rated thermal power (RTP) from the current licensed thermal power (CLTP) of 3486 megawatts thermal (MWt) to a measurement uncertainty recapture (MUR) thermal power of 3544 MWt. Columbia was originally licensed to 3323 MWt, and in 1995 a power uprate amendment authorized an increase in power to the CLTP of 3486 MWt.

The proposed changes are based on reduced uncertainty in the feedwater flow and temperature measurement that reduces the total power level measurement uncertainty.

This is achieved by utilizing Cameron International (formerly Caldon) CheckPlus' Leading Edge Flow Meter (LEFM) ultrasonic flow measurement instrumentation. The LEFM system was installed at Columbia during the spring 2015 refueling outage (RFO).

2.0 Detailed Description The proposed changes to the OL and TS are described below, with marked-up pages included in Enclosure 2.

A proposed new section to be added to the Licensee Controlled Specifications (LCS) and proposed changes to the TS Bases are also described below, with marked-up pages included in Enclosure 3. These changes are for information only, and do not require NRC approval.

2.1 Columbia Renewed Facility Operating License (OL)

Changes related to the value of RTP for Columbia, OL No. NPF-21, Section 2.C.(1),

Maximum Power Level, Current: The licensee is authorized to operate the facility at reactor core power levels not in excess of full power (3486 megawatts thermal).

Proposed: The licensee is authorized to operate the facility at reactor core power levels not in excess of full power (3544 megawatts thermal).

2.2 Columbia TS 1.1, Definition of Rated Thermal Power (RTP)

Current: RTP shall be a total reactor core heat transfer rate to the reactor coolant of 3486 MWt.

Proposed: RTP shall be a total reactor core heat transfer rate to the reactor coolant of 3544 MWt.

License Amendment Request to Revise Operating License and Technical Specifications for Measurement Uncertainty Recapture (MUR) Power Uprate Page 3 of 27 2.3 Columbia TS 3.3.1.1, RPS Instrumentation (After Implementation of PRNM Upgrade)

REQUIRED ACTION E.1 Current: Reduce THERMAL POWER to < 30% RTP Proposed: Reduce THERMAL POWER to < 29.5% RTP

Surveillance Requirement (SR) 3.3.1.1.12 Current: Verify Turbine Throttle Valve - Closure, and Turbine Governor Valve Fast Closure Trip Oil Pressure - Low Functions are not bypassed when THERMAL POWER is 30% RTP.

Proposed: Verify Turbine Throttle Valve - Closure, and Turbine Governor Valve Fast Closure Trip Oil Pressure - Low Functions are not bypassed when THERMAL POWER is 29.5% RTP.

TS Table 3.3.1.1-1, Reactor Protection System Instrumentation, Function 2.b Average Power Range Monitors Simulated Thermal Power - High, Allowable Value Current value: 0.63W + 64.0% RTP and 114.9% RTP(c)

Proposed value: 0.62W + 62.9% RTP and 114.9% RTP(c)

Table 3.3.1.1-1 Note (c)

Current: 0.63W + 60.8% RTP and 114.9% RTP when reset for single loop operation per LCO 3.4.1, Recirculation Loops Operating.

Proposed: 0.62W + 59.8% RTP and 114.9% RTP when reset for single loop operation per LCO 3.4.1, Recirculation Loops Operating.

Table 3.3.1.1-1 Function 8, Turbine Throttle Valve - Closure, Applicable Modes or Other Specified Conditions Current value: 30% RTP Proposed value: 29.5% RTP

Table 3.3.1.1-1 Function 9, Turbine Governor Valve Fast Closure, Trip Oil Pressure - Low, Applicable Modes or Other Specified Conditions Current value: 30% RTP Proposed value: 29.5% RTP 2.4 Columbia TS 3.3.4.1, End of Cycle Recirculation Pump Trip (EOC-RPT)

Instrumentation

APPLICABILITY Current: THERMAL POWER 30% RTP Proposed: THERMAL POWER 29.5% RTP

License Amendment Request to Revise Operating License and Technical Specifications for Measurement Uncertainty Recapture (MUR) Power Uprate Page 4 of 27

REQUIRED ACTION C.2.

Current: Reduce THERMAL POWER to < 30% RTP Proposed: Reduce THERMAL POWER to < 29.5% RTP

SR 3.3.4.1.3 Current: Verify TTV - Closure and TGV Fast Closure, Trip Oil Pressure -

Low Functions are not bypassed when THERMAL POWER is

30% RTP Proposed: Verify TTV - Closure and TGV Fast Closure, Trip Oil Pressure -

Low Functions are not bypassed when THERMAL POWER is

29.5% RTP 2.5 Columbia TS 3.3.6.1 Primary Containment Isolation Instrumentation

Table 3.3.6.1-1, Primary Containment Isolation Instrumentation, Function 1.c, Main Steam Line Flow - High, Allowable Value Current value: 124.4 psid Proposed value: 137.9 psid 2.6 Columbia Licensee Controlled Specifications (LCS) Changes, New Section (Information Only)

New LCS Section 1.3.9, LEFM Feedwater Flow Instrumentation, is added to specify the proposed requirements and bases for the LEFM system and to specify a surveillance requirement.

2.7 TS Bases Changes (Information Only)

The Bases for Sections 3.3.1.1 and 3.3.4.1 are changed to provide supporting bases discussions for the required TS changes identified in Section 2.3 through 2.5 above.

The Bases for Section 3.3.2.2, Feedwater and Main Turbine High Water Level Trip Instrumentation, are changed to incorporate the RTP value above which the Level 8 trip indirectly initiates a reactor scram from the main turbine trip.

The Bases for Section 3.7.6, Main Turbine Bypass System, are changed to reflect the bypass capacity of the system based on the revised steam flow of the main steam system.

Details of the aforementioned changes are provided in Enclosure 3.

3.0 Technical Evaluation 3.1 Background and General Approach 10 CFR 50, Appendix K, Paragraph I.A, Sources of Heat During the LOCA, requires that emergency core cooling system (ECCS) evaluation models assume that the reactor has been operating continuously at a power level at least 1.02 times the licensed power level to allow for instrumentation error. A change was made to this paragraph, which

License Amendment Request to Revise Operating License and Technical Specifications for Measurement Uncertainty Recapture (MUR) Power Uprate Page 5 of 27 became effective on July 31, 2000, that allows a lower assumed power level, provided the proposed value has been demonstrated to account for uncertainties due to power level instrumentation error. Utilization of the Cameron LEFM system at Columbia has resulted in reduced uncertainty in feedwater flow and temperature measurement that reduces the total power level measurement uncertainty. The core thermal power measurement uncertainty is described in Section 3.2.3 of this enclosure.

During the 2015 refueling outage, Columbia completed the installation of the LEFM CheckPlus system which included changes to the Plant Process Computer (PPC), the Transient Data Acquisition System (TDAS), the Plant Data Information System (PDIS) and the MONICORE core monitoring system. The LEFM CheckPlus system provides a more accurate reactor feedwater mass flow measurement. The LEFM system measures feedwater flow using ultrasonic pulses, which are digitally processed. Since installation, the LEFM system provides a more accurate feedwater flow input to the thermal heat balance calculation performed by the PPC. This calculated thermal power is used by the control room operators to monitor compliance with the OL condition for CLTP maximum power level, to determine the margins to the power distribution limits (PDL) and to calibrate the average power range monitor (APRM) neutron flux indication to represent actual reactor power. This amendment request, once approved, authorizes the changes identified in Sections 2.1 - 2.5 of this enclosure allowing an increase in RTP to the MUR thermal power of 3544 MWt.

The PPC provides indication and alerts related to the LEFM system. As discussed in the following sections of this enclosure, the PPC is also used to determine the difference between the feedwater flow indication from the LEFM system and the existing reactor feedwater flow venturi instrumentation for the purpose of data validation.

The scope and content of the evaluations performed and described in this request are in accordance with the guidance contained in NRC Regulatory Issue Summary (RIS) 2002-03, Guidance on the Content of Measurement Uncertainty Recapture Power Uprate Applications, (Reference 6.2). Enclosure 5 of this request provides a cross-reference between the contents of this application and the guidance in RIS 2002-03.

The ECCS evaluation and other plant safety analyses currently assume an uncertainty of 2% of the CLTP (3486 MWt). Energy Northwest has evaluated the effects of the proposed increase in RTP using an approach developed by General Electric-Hitachi (GEH) Nuclear Energy and approved by the NRC, which is documented in NEDC-32938P-A, Licensing Topical Report: Generic Guidelines and Evaluations for General Electric Boiling Water Reactor Thermal Power Optimization, (Reference 6.6). summarizes the results of all significant safety evaluations performed that justify increasing the licensed thermal power. Review of these analyses support the requested license power level increase to 3544 MWt.

License Amendment Request to Revise Operating License and Technical Specifications for Measurement Uncertainty Recapture (MUR) Power Uprate Page 6 of 27 3.2 LEFM Measurement and Core Thermal Power Uncertainty 3.2.1 LEFM Feedwater Flow and Temperature Measurement The ultrasonic feedwater flowmeter installed at Columbia is a Cameron LEFM CheckPlus ultrasonic multi-path, transit time flowmeter. This LEFM system will be used in lieu of the current venturi-based feedwater flow indication and resistance temperature detector (RTD) temperature indication to provide feedwater flow input for the plant thermal heat balance calculation. The currently installed feedwater flow venturis will be used if the LEFM is not functional. The LEFM system uses ultrasonic transit time principles to determine fluid velocity and sound velocity. This flow measurement method is described in Caldon topical reports ER-80P, Improving Thermal Power Accuracy and Plant Safety While Increasing Operating Power Level Using the LEFM Check' System, Revision 0 (Reference 6.7), and ER-157P, Supplement to Caldon Topical Report ER-80P: Basis for Power Uprates with an LEFM Check or an LEFM CheckPlus System, Revision 8 and Revision 8 Errata (Reference 6.8). These topical reports were approved by the NRC in documents titled, Comanche Peak Steam Electric Station, Units 1 and 2 - Review of Caldon Engineering Topical Report ER-80P, Improving Thermal Power Accuracy and Plant Safety While Increasing Power Level Using the LEFM System, (Reference 6.9) and Final Safety Evaluation for Cameron Measurement Systems Engineering Report ER-157P, Revision 8, Caldon Ultrasonics Engineering Report ER-157P, Supplement to Topical Report ER-80P: Basis for a Power Uprate with the LEFM Check or CheckPlus System, (Reference 6.10).

In References 6.9 and 6.10, the NRC established criteria for use of these topical reports in requests for license amendments. Energy Northwests response to those criteria is provided in Section 3.2.4 of this enclosure. 0 provides the analysis of the uncertainty contribution of the LEFM CheckPlus system operating in the Check Plus (normal) mode, as well as when operating in the Check (maintenance) mode, to the overall calculated thermal power uncertainty. This analysis is a bounding analysis for Columbia and was completed following the calibration of the LEFM spool pieces. Additionally, the as-built dimensions were inputs for all computations, and confirmed that the uncertainties in these dimensions lie within the bounding values used in the bounding analysis. The commissioning tests for the Columbia LEFM CheckPlus system confirmed that the time measurement uncertainties are within the bounding values used in the analysis.

The LEFM instrumentation is not safety-related. Components such as the spool pieces, system control cabinet and components, pressure transmitters, RTDs, and the power supplies are Quality Class 2, Seismic Category II. The LEFM system was designed and manufactured in accordance with Camerons Quality Assurance Program. Specific examples of quality measures undertaken in the design, manufacture, and testing of the LEFM system are provided in Reference 6.7, Section 6.4 and Table 6.1.

The LEFM CheckPlus system consists of a measurement spool piece meter in each feedwater line, two transmitter signal processing units per spool piece and two redundant central processing units (CPU). Each measurement spool piece contains 16

License Amendment Request to Revise Operating License and Technical Specifications for Measurement Uncertainty Recapture (MUR) Power Uprate Page 7 of 27 ultrasonic, multi-path, transit time transducers grouped into two planes of eight transducers each, two 4-wire RTDs, and two pressure transmitters.

The LEFM system installed at Columbia performs automatic continuous self-checking of the transducer signals and the calculation results. This testing provides verification that the digital circuits are operating correctly and the LEFM system is within its specified accuracy envelope.

The LEFM system has two operating modes as well as a fail mode. Normal operation for the LEFM system is the Check Plus mode. In this mode, both planes of transducers are in service and system operations are processed by both redundant CPUs. If the system is subjected to a failure involving a transducer or failure of one plane of operation due to a transmitter signal processing unit malfunction, the system reverts to the Check mode. The control room operators are provided a visual alarm on the PPC when the LEFM system shifts from the Check Plus mode (normal mode) to the Check mode (maintenance mode).

Check Plus Mode (normal mode):

When in the Check Plus mode, a system normal is displayed when all the feedwater flow, temperature, and header pressure signals for feedwater lines A and B are normal and operating within design limits. Calculated power level uncertainty associated with the LEFM flow measuring system in this condition is less than 0.3%.

The plant can operate at 3544 MWt as discussed in Section 3.2.3 of this enclosure.

Check Mode (maintenance mode):

When the LEFM system shifts from the Check Plus mode to the Check mode a visual alarm indicates that there has been a loss of LEFM system redundancy.

The LEFM system Check mode indicates a loss of function that causes it to operate outside that specified accuracy envelope of +/- 0.3%. Typically, this occurs due to a malfunction of a single path or plane and results in an uncertainty increase to +/- 0.5%. In the event of a failure of one path or plane that cannot be restored to full functionality (Check Plus mode) within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , power will be reduced from 3544 MWt to 3537 MWt as discussed in Section 3.2.3 of this enclosure. The plant can operate at this power level indefinitely. The operators will be provided with procedural guidance for those occasions when the LEFM system is in the Check mode.

Fail Mode:

The LEFM system's Fail mode indicates a loss of function that causes the LEFM system to operate outside the specified accuracy envelope of +/- 0.5%. In this case the power level uncertainty reverts to the 2.0% associated with the venturi flow meters and power will be reduced to 3486 MWt within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months if LEFM functionality cannot be restored.

License Amendment Request to Revise Operating License and Technical Specifications for Measurement Uncertainty Recapture (MUR) Power Uprate Page 8 of 27 The LEFM system has continuous operating online self-diagnostic processes to verify that the digital circuits are operating correctly and within the design basis uncertainty limits. These processes can identify failure conditions that will cause the LEFM to switch from the Check Plus mode to the Check mode or to the Fail mode. Validated LEFM data including calculated results, status, and signal process information is sent to the PPC at regular intervals. Calculated LEFM results are compared to venturi data and RTD instrument results as a means of further data validation.

The PPC will provide a visual alarm upon change in the LEFM system status on the operator overview display screen. This includes a change from the Check Plus mode to the Check mode or LEFM Fail mode and requires entry into the Compensatory Measures of the new LCS 1.3.9 as described in Enclosure 3. A visual alarm is provided on the operator overview display screen for sustained loss of data between the LEFM and PPC. In addition to a visual alarm, a loss of data link results in indication that entry into the Compensatory Measures of LCS 1.3.9 is required. Core thermal power calculations automatically revert to calibrated venturi output when the PPC does not have a valid LEFM signal.

When LEFM operation is governed by one of the LCS Conditions, the remaining Completion Time for the Required Compensatory Measures will be displayed (e.g., 71.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months remaining until Columbia is derated to 3486 MWt). Maximum allowed core thermal power (CTP) and indications of compliance with the maximum allowed CTP based on LEFM status are displayed.

The 72-hour Completion Time begins when the PPC screens located at the Reactor Operator and Control Room Supervisor stations begin flashing a predetermined warning. The warnings reflect the following conditions:

System status changes from Check Plus to Check mode due to:

o one LEFM feedwater flow meter in Check mode and one LEFM feedwater flow meter in the Check Plus mode, or o both LEFM feedwater flow meters in Check mode.

System status change to Fail mode due to one or both LEFM feedwater flow meters in Fail mode.

Validated loss of signal between LEFM and PPC.

Additionally, there are PPC displays that the operators can use to display detailed information about the LEFM connection status and the function of the LEFM components. This includes detailed information for transducers, the signal processing function and the CPU status. Columbia has two fully redundant PPCs. Each PPC includes redundant processes to collect data from each of the redundant CPUs in the LEFM cabinet.

Methods to determine LEFM system status and the cause of alarms are described in Cameron documentation which will be used to develop specific procedures for

License Amendment Request to Revise Operating License and Technical Specifications for Measurement Uncertainty Recapture (MUR) Power Uprate Page 9 of 27 operators and maintenance response actions. Justification for the 72-hour completion time is provided in Section 3.2.4 of this enclosure.

3.2.2 Plant Implementation The Columbia LEFM system was installed and commissioned in accordance with the appropriate Cameron installation and testing procedures. The LEFM measurement spool pieces were installed in the feedwater piping of the two feedwater lines as shown in the installation drawings provided in Enclosure 14.

The installations in feedwater lines A and B are located in straight sections of 24 inch feedwater pipe about 20 feet downstream of the existing feedwater flow venturis. Both spool pieces are located sufficiently remote from major hydraulic disturbances as required by Cameron spool piece installation specifications.

The transducers are located in the turbine building (TB) steam tunnel extension at the TB 501 foot elevation. The integrated gamma dose for 40 years of normal plant operation is 9.5E5 Rads. The material in the LEFM transducers has been exposed to gamma irradiation levels of 10 to 100 Mega Rads with negligible degradation in transducer performance. The system control cabinet is located outside the steam tunnel extension in an area with no significant gamma dose. Therefore, no radiation damage or degradation to the instruments due to the exposure levels in the plant is anticipated.

Following installation, testing included an inservice leak test, comparisons of feedwater flow and thermal power calculated by various methods, and final commissioning testing.

Final commissioning testing is described in Cameron's LEFM CheckPlus Flow Measurement System Installation and Commissioning Manual for Columbia Nuclear Power Plant (March 2014) (Reference 6.12). All testing was completed satisfactorily in July of 2015.

3.2.3 LEFM and Core Thermal Power Measurement Uncertainty and Methodology 0 provides an analysis of the uncertainty contribution of the LEFM CheckPlus system when operating in the Check Plus mode, as well as when operating in the Check mode, to the overall calculated thermal power uncertainty. At Columbia with the LEFM CheckPlus system in the Check Plus mode, calculated core thermal power uncertainty due to the LEFM system is +/- 0.276%. In the Check mode, calculated core thermal power uncertainty due to the LEFM system is +/- 0.485%. These uncertainties were calculated using the methodology described in Reference 6.8, which was approved by the NRC in Reference 6.10. These uncertainties were rounded up to 0.3% and 0.5% respectively, in the heat balance uncertainty calculation (Enclosure 13).

The measurement uncertainty recapture allows a licensed power level that maintains margin to 102% of CLTP. In Enclosure 13, 102% of 3486 MWt (3556 MWt) was used as a maximum value when determining the MUR power uprate value. The total thermal power heat balance calculation uncertainty is obtained by combining the input uncertainties as random terms except for control rod drive and reactor water cleanup flows which may have dependency due to a PPC bias, thus they are conservatively added together. This results in the following thermal power uncertainties and proposed

License Amendment Request to Revise Operating License and Technical Specifications for Measurement Uncertainty Recapture (MUR) Power Uprate Page 10 of 27 power levels. The method used in performing the above calculation is based on Energy Northwest Standard EES-4, Setpoint Methodology .

For the LEFM system operating in Check Plus mode, the heat balance calculation has an uncertainty of +/-11.649 MWt. This results in a power level of 3556 MWt -

11.649 MWt = 3544.351 MWt. The proposed power level in the Check Plus mode is rounded down to 3544 MWt. Therefore the requested increase in power is approximately 1.66% above the CLTP of 3486 MWt.

For the LEFM system operating in the Check mode, the heat balance calculation has an uncertainty of +/-18.586 MWt. This results in a power level of 3556 MWt - 18.586 MWt = 3537.414 MWt. The proposed power level in the Check (maintenance) mode is rounded down to 3537 MWt.

A revised heat balance calculation has been added to the PPC to support feedwater input from the LEFM system and the existing venturi flow nozzles.

Caldon Topical Report ER-157P, Revision 8 (Reference 6.8), states that the redundancy inherent in the two measurement planes of an LEFM CheckPlus system also makes this system more resistant to total failure when compared to the LEFM Check system. For any single component failure, continued operation at a power greater than that prior to the MUR power uprate can be justified with the LEFM system since the system with the failure is no less than an LEFM Check system.

The NRC SER (Reference 6.10) approving ER-157P, Revision 8 required licensees referencing ER-157P, Revision 8 to ensure compliance with these two limitations/conditions:

1. Continued operation at the pre-failure power level for a pre-determined time and the decrease in power that must occur following that time, are plant-specific and must be acceptably justified.

2. The only mechanical difference that potentially affects the Topical Report ER-157P, Revision 8 statement above is that the LEFM CheckPlus system has 16 transducer housing interfaces with the flowing water, whereas the LEFM Check System has 8.

Consequently, a LEFM CheckPlus system operating with a single failure that is assumed to disable one plane of transducers is not identical to an LEFM Check system. Although the effect on hydraulic behavior is expected to be negligible, this must be acceptably quantified if a licensee wishes to operate as stated. An acceptable quantification method is to establish the effect in an acceptable test configuration such as can be accomplished at the Alden Laboratory.

Cameron reports ER-1049 (Enclosure 10) and ER-1074 (Enclosure 11) identify the uncertainties associated with LEFM operation in the Check Plus mode and Check mode, including meter factor uncertainties specific to Columbia. These uncertainties were established by the calibration tests performed at Alden Research Laboratory. The impact of a failure disabling one plane of transducers on the LEFM system installed at Columbia has been quantified with an uncertainty of less than +/-0.5%. The associated increase in uncertainty from 0.3% to 0.5% results in a maximum allowable power level for this condition of 3537 MWt.

License Amendment Request to Revise Operating License and Technical Specifications for Measurement Uncertainty Recapture (MUR) Power Uprate Page 11 of 27 In the event the LEFM system is non-functional (Fail mode), the heat balance calculation will use the existing feedwater venturi flow nozzles until the LEFM system is returned to functional status. To ensure that the venturi-based heat balance calculation is consistent with the LEFM system based heat balance calculation, the venturi-based flow rate will be normalized to the pre-failure LEFM system flow rate.

The loss of the data link between the LEFM system and the PPC (beyond that associated with anticipated data flow interruptions) or a PPC failure will require reducing core thermal power to 3486 MWt within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . It is conservative to limit the power within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to this level until the LEFM system is returned to functional status. A new proposed LEFM feedwater flow instrumentation specification will be added to the LCS, as shown in Enclosure 3, to provide operators with actions to be taken when the LEFM system is not in the normal mode.

This meets the two limitations/conditions identified above.

3.2.4 Disposition of NRC Criteria for Use of LEFM Topical Reports In References 6.9 and 6.10, the NRC established criteria to be addressed by licensees incorporating the LEFM methodology into the licensing basis. The criteria are listed below, along with a discussion of how each is or will be satisfied.

Criterion 1 Discuss maintenance and calibration procedures that will be implemented with the incorporation of the LEFM, including processes and contingencies for inoperable LEFM instrumentation and the effect on thermal power measurements and plant operation.

Response to Criterion 1 Calibration and Maintenance Installation of the LEFMs included development of the necessary procedures and documents required for maintenance and calibration of the LEFM system. Plant maintenance and calibration procedures have been revised to incorporate Camerons maintenance and calibration requirements. Initial preventive maintenance scope and frequency are based on vendor recommendations. The incorporation of, and continued adherence to, these requirements will assure that the LEFM system is properly maintained and calibrated.

For instrumentation other than the LEFM system that contributes to the thermal power heat balance computation, calibration and maintenance is performed periodically using existing site procedures. Instrument channel accuracy, drift, calibration error and instrument error were evaluated and accounted for within the thermal power uncertainty calculation.

The LEFM system software and the PPC software configuration is maintained using existing Columbia procedures, which include verification and validation of changes to software configuration. Configuration of the hardware associated with the LEFM system and the instrumentation that contributes to the heat balance calculation is maintained in accordance with Columbia configuration control procedures.

License Amendment Request to Revise Operating License and Technical Specifications for Measurement Uncertainty Recapture (MUR) Power Uprate Page 12 of 27 Columbia programs and procedures addressing corrective actions, reporting deficiencies, and receiving and evaluating manufacturers deficiency reports are discussed in Section 3.2.5, Deficiencies and Corrective Actions.

LEFM Non-functionality and the Effect on Thermal Power Measurements and Plant Operations The redundancy inherent in the two measurement planes of an LEFM system as described in Enclosure 10 makes the system tolerant to component failures.

Continuously operating online self-diagnostic testing is provided to verify that the digital circuits are operating correctly and within the design basis uncertainty limits. LEFM system malfunctions result in PPC alarm messages to alert the operators if the status of the LEFM instrumentation changes. In these cases, the proposed LCS Compensatory Measures will be applied. Additionally, if the interface between the LEFM system and the PPC has failed, the LEFM will be considered non-operational and the proposed 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months allowed outage time would be entered and the LCS Compensatory Measures will be applied. As provided in Enclosure 3, the new LCS Requirements for Operation (RFO) 1.3.9, Feedwater Flow Instrumentation, will be implemented prior to raising thermal power above the CLTP (See Enclosure 6, Item 1).

The proposed LCS specification requires verification that each LEFM system meter is in the Check Plus mode every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . In addition to this confirmation of status, the PPC alarm messages described above alert the operators if the status of the LEFM instrumentation changes.

The existing feedwater flow venturi-based signals were calibrated using the LEFM system measured feedwater flow at the beginning of operating cycle 23, following the commissioning of the LEFM. The venturi calibration is revalidated and adjusted at the beginning of each cycle when the LEFM is operational at full power conditions. During the operating cycle, the input to the PPC from the venturis is also adjusted using the ratio between LEFM input and the venturi input. The ratio is calculated using 30 minute averaged feed water flow data from the LEFM and the venturi at rated conditions. The 30 minute average is satisfactory to negate the effects of bi-stable core flow as discussed in Regulatory Issue Summary 2007-21, Adherence to Licensed Power Limits. Feedwater flow input to the core thermal power calculation is provided by the existing feedwater flow venturis when LEFM data is not available. Since the feedwater flow venturis are corrected to the last validated data from the LEFM system, it is acceptable to remain at the MUR thermal power of 3544 MWt for up to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to enact LEFM system repairs. After 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , actions required by the LCS will be taken to reduce power to the appropriate level.

Since the LEFM Check Plus system has two modes of operation, LCS 1.3.9 allows for an intermediate power reduction. With one or both LEFM feedwater flow meters in the Check mode (one plane out of service on one or both meters), feedwater measurement uncertainty increases from 0.3% to 0.5%. This additional uncertainty equates to a 0.2% power reduction from MUR uprate thermal power to 3537 MWt. As noted in the LCS provided, if the LEFM system is not returned to functionality within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , power

License Amendment Request to Revise Operating License and Technical Specifications for Measurement Uncertainty Recapture (MUR) Power Uprate Page 13 of 27 will be reduced and administratively controlled to remain less than or equal to 3537 MWt. A similar allowance was approved in Reference 6.13.

The 72-hour Completion Time for the LEFM system prior to reducing to the CLTP is acceptable. As discussed above, during the 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion Time, the existing feedwater flow venturi-based signals will be corrected to the last validated data from the LEFM system. Although the feedwater flow venturi measurement signals may drift slightly during this period due to fouling of the feedwater flow venturis, such fouling results in a higher than actual indication of feedwater flow. This condition results in an overestimation of the calculated thermal heat balance power level, which is conservative, as the reactor will actually be operating below the calculated power level.

Note that the NRC has previously approved power uprate applications with Completion Times of up to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months for similar BWRs (References 6.3 through 6.5 and 6.13).

Regarding potential drift in the measurement of feedwater differential pressure across the feedwater flow venturis, industry experience for similar BWRs shows that the instrument drift associated with feedwater flow measurements are insignificant over a 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months time period. In Reference 6.7, Table A-1 provides the systematic error associated with feed flow nozzle differential pressure as approximately 1.0% over an operating cycle. Thus, over a 72-hour period, this would have an insignificant effect on the feedwater flow measurement.

A sudden de-fouling event during the 72-hour Completion Time is unlikely. Significant sudden de-fouling would be detected by a change in the balance of plant parameters. A review of recent plant operating experience has not identified any instances of sudden de-fouling events at Columbia.

Criterion 2 For plants that currently have LEFMs installed, provide an evaluation of the operational and maintenance history of the installed installation and confirmation that the installed instrumentation is representative of the LEFM system and bounds the analysis and assumptions set forth in Caldon Topical Report ER-80P.

Response to Criterion 2 The LEFMs were installed during the spring 2015 RFO. Following commissioning, the LEFM system was used to supply the feedwater flow input to the PPC core thermal power calculation and the station has remained 3486 MWt (CLTP). Since the commissioning of the LEFM, the following maintenance issues have occurred:

An error was introduced into the LEFM transmitter configuration files due to an incorrect configuration file change provided by the vendor, Cameron. Condition reports (CRs) were initiated to correct the error. A cause evaluation was performed and determined that there were weaknesses in the configuration control and a lack of rigor in validating vendor-supplied changes to the configuration file. Actions are being taken to address the causes including instituting more robust controls on software quality and configuration. Cameron has taken actions to fully review all current configuration files and provide a comparison file with explanations for any file

License Amendment Request to Revise Operating License and Technical Specifications for Measurement Uncertainty Recapture (MUR) Power Uprate Page 14 of 27 updates. This condition has been entered into the stations corrective action program.

Four out of 16 Paths have experienced degraded transducer signal quality.

Troubleshooting investigation points to loose wiring at the transducers as a likely cause. A work order was initiated to troubleshoot and correct the problem with the transducers when access to transducers is available. This condition has been entered into the stations corrective action program.

The LEFM CPU has experienced lock-ups which results in stale flow data being output to the plant computer. A CR was initiated to address this issue. Preliminary reviews by the vendor, Cameron, have identified an error in the LEFM watchdog timer configuration settings. Corrective actions are in place to resolve the configuration error. Cameron also recommends periodically rebooting the CPUs to eliminate lock-ups commonly experienced on personal computers that are continuously running. Actions to create recurring tasks to reboot the CPUs quarterly are being taken. This condition has been entered into the stations corrective action program.

These issues have been discussed with Cameron, who is working with the LEFM system engineers to assess the issues and provide resolutions. Cameron has also agreed to provide a root cause report outlining all errors experienced and root causes of these errors.

Criterion 3 Confirm that the methodology used to calculate the uncertainty of the LEFM in comparison to the current feedwater instrumentation is based on the accepted plant setpoint methodology (with regard to the development of instrument uncertainty). If an alternative approach is used, the application should be justified and applied to both venturi and ultrasonic flow measurement instrumentation installations for comparison.

Response to Criterion 3 The method used in performing this calculation is based on the accepted plant setpoint methodology Standard EES-4, Setpoint Methodology. This standard is based on the American Society of Mechanical Engineers (ASME) PTC 19.1-1985, Measurement Uncertainty, and the Instrumentation, Systems, and Automation Society (ISA)

RP67.04.02-2000, Methodologies for the Determination of Setpoints for Nuclear Safety-Related Instrumentation. The methodologies used in the heat balance determination (Enclosure 13) are discussed in Section 3.2.3 of this enclosure.

Criterion 4 For plants where the ultrasonic meter (including LEFM) was not installed with flow elements calibrated to a site-specific piping configuration (i.e., flow profiles and meter factors not representative of the plant specific installation), additional justification should be provided for its use. The justification should show that the meter installation is either independent of the plant specific flow profile for the stated accuracy, or that the installation can be shown to be equivalent to known calibrations and plant

License Amendment Request to Revise Operating License and Technical Specifications for Measurement Uncertainty Recapture (MUR) Power Uprate Page 15 of 27 configurations for the specific installation including the propagation of flow profile effects at higher Reynolds numbers. Additionally, for previously installed calibrated elements, confirm that the piping configuration remains bounding for the original LEFM installation and calibration assumptions.

Response to Criterion 4 This is not applicable to Columbia. The calibration factors for the Columbia ultrasonic LEFM flow meters were established by tests of these flow meters at Alden Research Laboratory. These tests were performed on a full-scale model of the Columbia hydraulic geometry. A discussion of the impact of the plant-specific installation factors on the feedwater measurement uncertainty is provided in Cameron Report ER-1049, Revision 3, (Enclosure 10) and Cameron Report ER-1074, Revision 0 (Enclosure 11).

The test configurations modeled the portion of piping upstream of the LEFM spool pieces and can be compared to the plant installation drawings by comparing the drawings Enclosure 11, Figures 2.1 and 2.2, to the installation drawings in Enclosure

14. There is no significant difference between the Columbia feedwater piping configuration and the test configuration used at Alden Research Laboratory.

Criterion 5 Continued operation at the pre-failure power level for a pre-determined time and the decrease in power that must occur following that time are plant-specific and must be acceptably justified.

Response to Criterion 5 Justification for continued operation at the pre-failure level for a predetermined time and the actions to be taken in the event that time is exceeded (i.e., power reduction) is provided in the response to Criterion 1 above.

Criterion 6 A CheckPlus operating with a single failure is not identical to an LEFM Check. Although the effect on hydraulic behavior is expected to be negligible, this must be acceptably quantified if a licensee wishes to operate using the degraded CheckPlus at an increased uncertainty.

Response to Criterion 6 As identified in Enclosure 10, using the total thermal power uncertainty approach documented in Reference 6.8, the uncertainty in the Columbia LEFM CheckPlus system measurement is as follows:

Total thermal power uncertainty in the LEFM Check Plus mode is +/- 0.276%.

Total thermal power uncertainty in the LEFM Check mode is +/- 0.485%.

The LEFM CheckPlus system is in Check mode when one or both LEFM system meters are in the Check mode and not in Fail. The total uncertainty of the LEFM CheckPlus system operating in the Check mode was evaluated in Enclosure 11 and resulted in the increased uncertainty stated above.

License Amendment Request to Revise Operating License and Technical Specifications for Measurement Uncertainty Recapture (MUR) Power Uprate Page 16 of 27 Criterion 7 An applicant with a comparable geometry can reference the Section 3.2.1 finding (of Reference 6.10) to support a conclusion that downstream geometry does not have a significant influence on CheckPlus calibration. However, CheckPlus test results do not apply to a Check and downstream effects with use of a CheckPlus with disabled components that make the CheckPlus comparable to a Check must be addressed. An acceptable method is to conduct applicable Alden Laboratory tests.

Response to Criterion 7 The installation configuration of the Columbia LEFM system spool pieces are described in Section 3.2.2 of this enclosure. Testing was conducted at Alden Research Laboratories as described in Enclosure 11. The hydraulic model configuration was designed as a hydraulic duplicate of the principle hydraulic features of the installation site (ALD-1160, Hydraulic Calibration Plan for Columbia Nuclear Generating Station, Revision 2, which is Reference 1 of Enclosure 11, contains the plant details). The tests conducted at the Alden Research Laboratories verified the use of an LEFM CheckPlus system with disabled components make the CheckPlus system comparable to a Check system. The testing supports that the downstream geometry does not have a significant influence on the Columbia LEFM system calibration.

Criterion 8 An applicant that requests a MUR with the upstream flow straightener configuration discussed in Section 3.2.2 (of Reference 6.10) should provide justification for claimed CheckPlus uncertainty that extends the justification provided in Reference 17 (of Reference 6.10). Since the Reference 17 evaluation does not apply to the Check, a comparable evaluation must be accomplished if a Check is to be installed downstream of a tubular flow straightener.

Response to Criterion 8 The LEFM system spool pieces at Columbia are both located in the feed water lines downstream of a 90 degree elbow. The venturi flow elements are located upstream of the 90 degree elbow. A flow straightener is located at the inlet to each of the venturi flow elements. The arrangement of the 90 degree elbow, the venturi flow element and the flow straightener were all modeled in detail during testing at Alden Research Laboratories. A full range of flow tests were performed in the normal piping configuration on both LEFM meters. Flow testing was also performed by rotating the flow straightener, which indicated that it had no significant effect in the LEFM calibration. Additional flow testing was performed with inline flow disruptions, half moon plates at various locations before the flow straightener, before the venturi and before the 90 degree elbow. These flow disruptions created significantly larger flow profile asymmetry and flow swirl than existed in the normal plant piping configuration. The testing results indicated that actual increases in the flow profile asymmetry and flow swirl cause the LEFM meter to indicate a more conservative flow. Based on the results of this testing the flow straightener located upstream of the venturi is sufficiently far

License Amendment Request to Revise Operating License and Technical Specifications for Measurement Uncertainty Recapture (MUR) Power Uprate Page 17 of 27 enough upstream of the LEFM meter that its effect does not significantly impact the operation of the LEFM in the Check Plus or Check mode.

Criterion 9 An applicant assuming large uncertainties in steam moisture content should have an engineering basis for the distribution of the uncertainties or, alternatively, should ensure that their calculations provide margin sufficient to cover the differences shown in Figure 1 of Reference 18 of Reference 6.10.

Response to Criterion 9 Columbia conservatively assumes no moisture content in the core thermal power uncertainty calculation (Enclosure 13). This approach is consistent with that described in Section 3.2.3 of Reference 6.10. Thus, this criterion is not applicable to Columbia.

3.2.5 Deficiencies and Corrective Actions Cameron has procedures to notify users of important LEFM deficiencies. Columbia also has processes for addressing manufacturer's deficiency reports. Such deficiencies are documented in Columbia's corrective action program. Deficiencies associated with the vendors processes or equipment are reported to the vendor to support corrective action.

3.2.6 Reactor Power Monitoring Energy Northwest's Policy Statement Manual provides guidance to ensure that reactor power remains within the requirements of the operating license. Plant procedures provide requirements for monitoring and controlling reactor power in compliance with TS that is consistent with the guidance proposed by the Nuclear Energy Institute (NEI) and endorsed by the NRC in Reference 6.11.

3.3 Evaluation of Changes to Operating License and Technical Specifications The proposed changes to the TS described in Section 2.0, Detailed Description, are evaluated below. The numbering of these changes corresponds to the numbering in Section 2.0.

Sections 2.1 and 2.2, Changes Related to RTP The proposed increase in RTP in the Columbia OL and TS Definitions is acceptable based on the decreased uncertainty in the core thermal power calculation due to the use of the LEFM feedwater flow measurement system and on the evaluations provided in this License Amendment Request.

Section 2.3, Changes Related to Revised Allowable Values for the Average Power Range Monitors Simulated Thermal Power - High Trip Function The proposed changes in the two-loop and single-loop Average Power Range Monitor Simulated Thermal Power - High trip functions are contained in TS 3.3.1.1 Table 3.3.1.1-1, Function 2.b. The proposed change to the Allowable Values (AVs) for the Average Power Range Monitors Simulated Thermal Power - High Trip functions are based on the approach described in Reference 6.6, Section F.4.2.1, Flow Referenced

License Amendment Request to Revise Operating License and Technical Specifications for Measurement Uncertainty Recapture (MUR) Power Uprate Page 18 of 27 APRM Trip and Alarm Setpoints. The Average Power Range Monitor Simulated Thermal Power - High trip function AVs, for both two-loop operation and single-loop operation, are unchanged in units of absolute core thermal power versus recirculation drive flow. Because these values are expressed in percent of RTP, they decrease in proportion to the MUR power uprate. The specific values are provided in Section 5.3, Technical Specification Instrument Setpoints, of Enclosure 7. The AVs were generated using approved GEH setpoint methodology. Further discussion of the setpoint methodology is found in Section 3.4.4 of this enclosure.

Section 2.3, Changes Related to Revised Allowable Values for Turbine Throttle Valve -

Closure and Turbine Governor Valve - Fast Closure, Trip Oil Pressure - Low The proposed change for the power level at which the Turbine Throttle Valve - Closure and Turbine Governor Valve - Fast Closure, Trip Oil Pressure - Low trip functions are bypassed are contained in TS 3.3.1.1, Required Action E.1, SR 3.3.1.1.12, and Table 3.3.1.1-1, Functions 8 and 9. The bypass of these trip functions is accomplished by sensing turbine first-stage pressure. Based on the guidelines in Section F.4.2.3, Turbine First-Stage Pressure Signal Setpoint, of Reference 6.6, the value at which the Turbine Throttle Valve - Closure trip and Turbine Governor Valve - Fast Closure, Trip Oil Pressure - Low trip functions are bypassed, in percent of RTP, is reduced by the ratio of the MUR power uprate increase. The value does not change with respect to absolute thermal power. The specific values are provided in Section 5.3 of Enclosure 7.

Section 2.4, Changes related to End of Cycle Recirculation Pump Trip (EOC-RPT)

Instrumentation The proposed change to the power level at which the Turbine Throttle Valve (TTV) -

Closure and Turbine Governor Valve (TGV) Fast Closure, Trip Oil Pressure - Low trip functions are bypassed are contained in TS 3.3.4.1, APPLICABILITY, Required Action C.2, and SR 3.3.4.1.3. The EOC-RPT function is automatically disabled by sensing turbine first stage pressure. Based on the guidelines in Reference 6.6, Section F.4.2.3, Turbine First-Stage Pressure Signal Setpoint, the value at which the TTV - Closure and TGV Fast Closure, Trip Oil Pressure - Low trip functions are bypassed, in percent of RTP, is reduced by the ratio of the MUR power uprate increase. The value does not change with respect to absolute thermal power. The specific values are provided in Section 5.3 of Enclosure 7.

Section 2.5, Changes related to the Primary Containment Isolation Instrumentation The proposed change to the Main Steam Line Flow - High pressure setpoint is contained in TS 3.3.6.1, Table 3.3.6.1-1, Function 1.c. As stated in Section 5.3.5, Main Steam Line High Flow Isolation, of Enclosure 7, a new setpoint as a result of the increased steam flow was calculated using the GEH setpoint methodology. A TS AV change is required to change the differential pressure setpoint at the allowable steam flow.

License Amendment Request to Revise Operating License and Technical Specifications for Measurement Uncertainty Recapture (MUR) Power Uprate Page 19 of 27 3.4 Additional Considerations 3.4.1 Summary of Analyses The following is a summary of the analyses performed in support of these proposed changes, along with the results and a reference to the sections of Enclosure 7 providing further detail.

Enclosure 7 Topic Conclusion Section Normal Plant MUR power uprate is accommodated by increasing Section 1 Operating core flow along previously established MELLLA rod Conditions lines.

Reactor Core Reactor core and fuel design is adequate for operation Section 2 and Fuel at MUR uprated conditions.

Performance Reactor Overpressure protection, fracture toughness, Section 3 Coolant and structural, and piping evaluations are acceptable.

Connected Systems Engineered Acceptable based on previous analyses at 102% of Section 4 Safety current licensed power.

Features Instrumentation Current instrumentation is acceptable. Changes to Section 5 and Control some TS values are necessary.

Electrical Minor increases in normal power system loads. Section 6 Power and Emergency power systems are unaffected. Auxiliary Auxiliary systems are acceptable.

Systems Power Power conversion systems are adequate without Section 7 Conversion modification.

Systems Radwaste and Small increases in normal operation radiation levels Section 8 Radiation and effluents. Accident consequences are bounded by Sources previous evaluations.

Reactor Safety Design basis events are bounded by previous Section 9 Performance evaluations. Special events meet acceptance criteria.

Evaluations Other All evaluation results are acceptable. Section 10 Evaluations 3.4.2 Adverse Flow Effects Industry experience has revealed that power uprate conditions can cause vibrations associated with acoustic resonance that can lead to steam dryer and main steam line

License Amendment Request to Revise Operating License and Technical Specifications for Measurement Uncertainty Recapture (MUR) Power Uprate Page 20 of 27 (MSL) valve degradation. This experience has been associated with extended power uprates (EPUs), and not with smaller uprates, such as MUR power uprates.

The generic evaluation provided in Reference 6.6, Appendix J.2.3.7 is applicable to Columbia. The requirements for the main steam isolation valves (MSIVs) remain unchanged for MUR uprate conditions. All safety and operational aspects of the MSIVs are within previous evaluations.

The stresses of the RPV internals that were affected by GEH Safety Communications were reconciled for the increase of the acoustic load to show that adequate stress margins still exist and the stresses remain within the allowable limits. All the RPV internals were shown to be within the allowable limits. The limiting stresses of all RPV internal components are summarized in Enclosure 7, Table 3-8. Therefore the RPV internal components are demonstrated to be structurally qualified for operation at MUR uprate conditions.

Based on the above, no adverse flow induced vibration effects are expected as a result of the MUR power uprate.

3.4.3 Plant Modifications The evaluations performed to support the MUR power uprate identified that no physical modifications are required to plant systems. However, software changes to PPC are required to support the interface with the LEFM system for operation above the CLTP limit of 3486 MWt.

3.4.4 Instrument Setpoint Methodology The determination of Allowable Values described in Section 2.0 of this enclosure is based on the GEH setpoint methodology. Reference 6.6 used approved GEH setpoint methodology to generate the values. Each actual trip setting is established to preclude inadvertent initiation of the protective action, while assuring adequate allowances for instrument accuracy, calibration, drift and applicable normal and accident design basis events.

Columbia previously adopted portions of Technical Specification Task Force (TSTF)

Traveler TSTF-493, Clarify Application of Setpoint Methodology for LSSS Functions, Revision 4, for the Average Power Range Monitor instrumentation in Amendment 226, which was approved on January 31, 2014. This amendment added Notes (d) and (e) to TS Table 3.3.1.1-1.

3.4.5 Grid Stability Studies The Columbia Final Safety Analysis Report (FSAR) Section 6.3.2.2 on equipment and component descriptions states the following. Regular AC power is from the main transformers [TR-N(1) and (2)] during plant operation or from the startup transformer (TR-S) (an offsite power source) when the main generator is off-line. Should regular AC power be lost, Division 1 (low-pressure core spray (LPCS) and low-pressure coolant injection (LPCI) loop A) and Division 2 (LPCI loops B and C) would be transferred to a second offsite power supply and backup transformer (TR-B). Division 3 high pressure core spray (HPCS) would be powered from its onsite standby diesel. If the backup

License Amendment Request to Revise Operating License and Technical Specifications for Measurement Uncertainty Recapture (MUR) Power Uprate Page 21 of 27 transformer were also lost, Divisions 1 and 2 would then be powered from their respective and independent onsite standby diesels. A more detailed description of the power supplies for the ECCS is contained in FSAR Section 8.3. 5 provides the grid study performed to assess the effects of the MUR power uprate. A steady-state power-flow study and a transient study were performed for specific contingencies including transfer of station service load to TR-S or TR-B offsite sources following a reactor scram. The power flow studies are comprised of a post-contingency voltage assessment and a voltage stability study. The study found no adverse effects from the additional generating capacity resulting from the uprate and that the existing system has the ability to maintain the required 1.0 p.u.(per unit) voltage at the off-site station service sources TR-S and TR-B. The transient stability studies ensure that 500-kV line faults or loss of major generation does not result in undamped conditions, voltage dip violation or frequency excursion violations in accordance with the reliability criteria. Results for the transient study cases show that each contingency was transiently stable and dynamically damped.

3.4.6 Operator Training, Human Factors, and Procedures The operator response to plant transients or accidents is unaffected by the proposed power uprate changes. When the LEFM system status shifts from the Check Plus mode to the Check mode, the control room operators are alerted with a visual alarm from the PPC. The proposed LCS Requirement for Operation provided in Enclosure 3 provides the Required Compensatory Measures and Completion Times for the identified Conditions. These are the only new operator actions associated with this license amendment request. The PPC, with displays at the Reactor Operator and Control Room Supervisor stations, will provide a visual alarm to alert the operators to changes in the LEFM system status (See Enclosure 6, Item 6). The LEFM electronics unit installed in the system control cabinet contains a display and keyboard that is used to respond to system status changes when indicated by the PPC visual alarm.

The Plant Process Computer provides LEFM status information through the PPC Overview display. The initial indication of a change in LEFM status is immediate but non-intrusive to the operators. This ensures that the operators are aware that the status of the LEFM has changed but does not require any immediate action from the operators. The PPC includes a nominal time allowance for normal and expected operational conditions such as momentary rejection of transducer data or momentary failure of the LEFM data validation check that are resolved without operator intervention.

These conditions are normally self-correcting. If the condition exists for longer than the specified time allowance, then an actual LEFM failure may exist. At this point the PPC will generate the visual alarm to notify the operator of a change in the LEFM system status. This ensures that the Human Machine Interface (HMI) of the PPC does not become an operator distraction. An audible alarm is not required since the operators routinely monitor the PPC Overview display as part of the normal duties to ensure thermal power is maintained within limits.

License Amendment Request to Revise Operating License and Technical Specifications for Measurement Uncertainty Recapture (MUR) Power Uprate Page 22 of 27 Necessary operating procedure revisions will be completed prior to implementation of the proposed MUR power uprate (See Enclosure 6, Item 2). The plant simulator will be modified for the uprated conditions and the changes will be validated in accordance with plant configuration control processes (See Enclosure 6, Item 3). Any necessary operator training will be completed prior to implementation of the proposed changes (See Enclosure 6, Item 4).

3.4.7 Plant Testing Plant testing for the MUR power uprate will be completed as described in Section 10.4, Testing of Enclosure 7 (See Enclosure 6, Item 5).

4.0 Regulatory Evaluation 4.1 Applicable Regulatory Requirements/Criteria 10 CFR 50, Appendix K, ECCS Evaluation Models, requires that emergency core cooling system evaluation models assume that the reactor has been operating continuously at a power level at least 1.02 times the licensed power level to allow for instrumentation error. A change to this paragraph, which became effective on July 31, 2000, allows a lower assumed power level, provided the proposed value has been demonstrated to account for uncertainties due to power level instrumentation error.

10 CFR 50, Appendix K does not permit licensees to utilize a lower uncertainty and increase thermal power without NRC approval. 10 CFR 50.90 requires that licensees desiring to amend an operating license file an amendment with the NRC.

RIS 2002-03, Guidance on the Content of Measurement Uncertainty Recapture Power Uprate Applications, provides criteria for the content of license amendment requests involving power uprates based on measurement uncertainty recapture.

This application is consistent with the requirements and criteria described in 10 CFR 50, Appendix K, 10 CFR 50.90, and the guidelines of RIS 2002-03.

4.2 Precedents The following facilities have recently received NRC approval for power uprates based on use of the LEFM system.

Facility Amendment No(s). Approval Date Accession No.

LaSalle, Units 1 and 198/185 September 16, 2010 ML101830361 2*

Limerick, Units 1 201/163 April 8, 2011 ML110691095 and 2*

Fermi 2* February 10, 2014 ML13364A131 196 Correction March 14, 2014 ML14066A410 Shearon Harris* 139 May 30, 2012 ML11356A096

CheckPlus system

License Amendment Request to Revise Operating License and Technical Specifications for Measurement Uncertainty Recapture (MUR) Power Uprate Page 23 of 27 Unlike this Columbia submittal, the listed precedent submittals of LaSalle, Limerick and Fermi also included a request that included TSTF-493, Clarify Application of Setpoint Methodology for LSSS Functions, Revision 4. Columbia incorporated portions of TSTF-493 as discussed in Section 3.4.4, Instrument Setpoint Methodology.

Similar to the approved Shearon Harris submittal, Columbia is also proposing use of the Check mode allowing the use of an increased uncertainty allowing operation at power level greater than the CLTP, but less than MUR uprated power as discussed in Section 3.2.1 of this enclosure.

4.3 No Significant Hazards Consideration In accordance with 10 CFR 50.90, Application for Amendment of License, Construction Permit, or Early Site Permit and 10 CFR 50, Appendix K, ECCS Evaluation Models, Energy Northwest requests an amendment to Columbia Generating Station (Columbia)

Renewed Facility Operating License (OL) NPF-21. Specifically, the proposed changes revise the OL and Technical Specifications (TS) to implement an increase of approximately 1.66% in RTP from 3486 megawatts thermal (MWt) to 3544 MWt. These changes are based on increased feedwater measurement accuracy, which was achieved by utilizing Cameron International (formerly Caldon) CheckPlus Leading Edge Flow Meter (LEFM) ultrasonic flow measurement instrumentation.

According to 10 CFR 50.92, Issuance of Amendment, paragraph (c), a proposed amendment to an operating license does not involve a significant hazard if operation of the facility in accordance with the proposed amendment would not:

(1) Involve a significant increase in the probability or consequences of any accident previously evaluated; or (2) Create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) Involve a significant reduction in a margin of safety.

Energy Northwest has evaluated the proposed changes, using the criteria in 10 CFR 50.92, and has determined that the proposed changes do not involve a significant hazards consideration. The following information is provided to support a finding of no significant hazards consideration.

1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No The proposed change will increase the Columbia Generating Station rated thermal power from 3486 MWt to 3544 MWt. The reviews and evaluations performed to support the proposed uprated power conditions included all structures, systems and components that would be affected by the proposed changes. The reviews and evaluations determined that these structures, systems, and components are capable of performing their design function at the proposed uprated RTP of 3544 MWt. All accident mitigation systems will function as designed, and all performance requirements

License Amendment Request to Revise Operating License and Technical Specifications for Measurement Uncertainty Recapture (MUR) Power Uprate Page 24 of 27 for these systems have been evaluated and were found acceptable. Thus, the proposed changes do not create any new accident initiators or increase the probability of an accident previously evaluated.

The primary loop components (e.g., reactor vessel, reactor internals, control rod drive housings, piping and supports, and recirculation pumps) remain within their applicable structural limits and will continue to perform their intended design functions. Thus, there is no increase in the probability of a structural failure of these components.

The nuclear steam supply systems will continue to perform their intended design functions during normal and accident conditions. The balance of plant systems and components continue to meet their applicable structural limits and will continue to perform their intended design functions. Thus, there is no increase in the probability of a failure of these components. The safety relief valves and containment isolation valves meet design sizing requirements at the uprated power level. Because the integrity of the plant will not be affected by operation at the uprated condition, Energy Northwest has concluded that all structures, systems, and components required to mitigate a transient remain capable of fulfilling their intended functions.

The current safety analyses remain applicable, since they were performed at power levels that bound operation at a core power of 3544 MWt. The results demonstrate that acceptance criteria of the applicable analyses continue to be met at the uprated conditions. As such, all applicable accident analyses continue to comply with the relevant event acceptance criteria. The analyses performed to assess the effects of mass and energy releases remain valid. The source terms used to assess radiological consequences have been reviewed and determined to bound operation at the uprated condition.

Power level is an input assumption to equipment design and accident analyses, but it is not a transient or accident initiator. Accident initiators are not affected by power uprate, and plant safety barrier challenges are not created by the proposed changes.

Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No No new accident scenarios, failure mechanisms, or limiting single failures are introduced as a result of the proposed changes. The proposed changes do not adversely affect any current system interfaces or create any new interfaces that could result in an accident or malfunction of a different kind than previously evaluated. All structures, systems and components previously required for the mitigation of a transient remain capable of fulfilling their intended design functions. The proposed changes have no adverse effects on any safety-related system or component and do not challenge the performance or integrity of any safety-related system.

Plant operation at a RTP of 3544 MWt does not create any new accident initiators or precursors. Credible malfunctions are bounded by the current accident analysis of

License Amendment Request to Revise Operating License and Technical Specifications for Measurement Uncertainty Recapture (MUR) Power Uprate Page 25 of 27 record or recent evaluations demonstrate that applicable criteria are still met with the proposed changes. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No The margins of safety associated with the power uprate are those pertaining to core thermal power. Operation at the uprated power condition does not involve a significant reduction in a margin of safety. Analyses of the primary fission product barriers have concluded that relevant design criteria remain satisfied, both from the standpoint of the integrity of the primary fission product barrier, and from the standpoint of compliance with the required acceptance criteria. As appropriate, all evaluations have been performed using methods that have either been reviewed or approved by the Nuclear Regulatory Commission, or that are in compliance with regulatory review guidance and standards.

Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

4.4 Conclusions Based on the above evaluation, Energy Northwest concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92, paragraph (c), and accordingly, a finding of no significant hazards consideration is justified.

In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or the health and safety of the public.

5.0 Environmental Consideration 10 CFR 51.22, Criterion for Categorical Exclusion; Identification of Licensing and Regulatory Actions Eligible for Categorical Exclusions or Otherwise Not Requiring Environmental Review, addresses requirements for submitting environmental assessments as part of licensing actions. 10 CFR 51.22, paragraph (c)(9) states that a categorical exclusion applies for Part 50 license amendments that meet the following criteria:

i. No significant hazards consideration (as defined in 10 CFR 50.92(c));

ii. No significant change in the types or significant increase in the amounts of any effluents that may be released offsite; and iii. No significant increase in individual or cumulative occupational radiation exposure.

The proposed changes do not involve a significant hazards consideration. The reviews and evaluations performed to support the proposed uprated power conditions concluded

License Amendment Request to Revise Operating License and Technical Specifications for Measurement Uncertainty Recapture (MUR) Power Uprate Page 26 of 27 that all systems will function as designed, and all performance requirements for these systems have been evaluated and found acceptable. No new accident scenarios, failure mechanisms, or limiting single failures are introduced as a result of the proposed changes. Operation at the uprated power condition does not involve a significant reduction in a margin of safety.

There is no significant change in the types or significant increase in the amounts of any effluents. Evaluations of the effects of the proposed changes on effluent sources concluded that the increase in effluents will be small, and within the current applicable permits and regulations.

There is no significant increase in individual or cumulative occupational radiation exposure. Evaluations of projected radiation exposure concluded that normal operation radiation levels increase slightly for the proposed power uprate, but that occupational exposure is controlled by the plant radiation protection program and is maintained well within values required by regulations.

Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22; paragraph (c)(9). Therefore, pursuant to 10 CFR 51.22, paragraph (b), no environmental impact statement or environmental assessment is required in connection with the proposed amendment.

6.0 References 6.1 Columbia NRC Docket No. 50-397 NRC License No. NPF-21 6.2 NRC Regulatory Issue Summary 2002-03, Guidance on the Content of Measurement Uncertainty Recapture Power Uprate Applications, dated January 31, 2002 (ML013530183) 6.3 Letter from Carl F. Lyon (USNRC) to Stewart B. Minahan (Nebraska Public Power District), Cooper Nuclear Station - Issuance of Amendment Re: Measurement Uncertainty Recapture Power Uprate (TAC No. MD7385), dated June 30, 2008 (ML081540280) 6.4 Letter from Christopher Gratton (USNRC) to Michael J. Pacilio (Exelon Nuclear),

LaSalle County Station, Units 1 and 2 - Issuance of Amendments Re:

Measurement Uncertainty Recapture Power Uprate (TAC Nos. ME3288 and ME3289), dated September 16, 2010 (ML101830361) 6.5 Letter from Peter Bamford (USNRC) to Michael J. Pacilio (Exelon Nuclear).

Limerick Generating Station, Units 1 and 2 - Issuance of Amendments Re:

Measurement Uncertainty Recapture Power Uprate and Standby Liquid Control System Changes (TAC Nos. ME3589, ME3590, ME3591, and ME3592), dated April 8, 2011 (ML110691095) 6.6 General Electric-Hitachi (GEH) Nuclear Energy Report NEDC-32938P-A, Licensing Topical Report: Generic Guidelines and Evaluations for General Electric Boiling Water Reactor Thermal Power Optimization, Revision 2, dated May 2003

License Amendment Request to Revise Operating License and Technical Specifications for Measurement Uncertainty Recapture (MUR) Power Uprate Page 27 of 27 6.7 Caldon Topical Report ER-80P, Improving Thermal Power Accuracy and Plant Safety While Increasing Operating Power Level Using the LEFM Check' System, Revision 0, dated March 1997 6.8 Caldon Topical Report ER-157P, Supplement to Caldon Topical Report ER-80P:

Basis for Power Uprates with an LEFM Check or an LEFM CheckPlus System, Revision 8, dated May 2008 6.9 Letter from John N. Hannon (USNRC) to C. Lance Terry (TU Electric),

Comanche Peak Steam Electric Station, Units 1 and 2 - Review of Caldon Engineering Topical Report ER 80P, Improving Thermal Power Accuracy and Plant Safety While Increasing Power Level Using the LEFM System, (TACS Nos.

MA2298 and MA2299), dated March 8, 1999 (ML9903190053) 6.10 Letter from Thomas B. Blount (USNRC) to Ernest Hauser (Cameron), Final Safety Evaluation for Cameron Measurement Systems Engineering Report ER-157P, Revision 8, Caldon Ultrasonics Engineering Report ER-157P, Supplement to Topical Report ER-80P: Basis for a Power Uprate With the LEFM Check or CheckPlus System, (TAC No. ME1321), dated August 16, 2010 (ML102160663) 6.11 Memorandum from Timothy Kolb (USNRC) to Mike Case (USNRC), Safety Evaluation Regarding Endorsement of NEI Guidance for Adhering to the Licensed Thermal Power Limit (TAC No. MD9233), dated October 8, 2008 (ML082690105) 6.12 Cameron Manual IB1404 LEFM CheckPlus Flow Measurement System Installation and Commissioning Manual for Columbia Nuclear Power Plant, Revision 0, dated March 2014 6.13 Letter form Araceli T. Billoch Colon (USNRC) to Chris Burton (Progress Energy Carolinas, Inc.) Shearon Harris Nuclear Power Plant, Unit 1 - Issuance of Amendment Re: Measurement Uncertainty Recapture Power Uprate (TAC NO.

ME6169), dated May 30, 2012 (ML11356A096)

License Amendment Request to Revise Operating License and Technical Specifications for Measurement Uncertainty Recapture (MUR) Power Uprate Enclosure 2 Markup of Existing Renewed Facility Operating License and Technical Specifications Renewed Facility Operating License Page 3 Technical Specifications Pages 1.1-5 3.3.1.1-10 3.3.1.1-13 3.3.1.1-15 3.3.1.1-18 3.3.4.1-1 3.3.4.1-2 3.3.4.1-3 3.3.6.1-5

(2) Pursuant to the Act and 10 CFR Part 70, to receive, possess and use at any time special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Final Safety Analysis Report, as supplemented and amended; (3) Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4) Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use in amounts as required any byproduct, source of special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (5) Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

(6) Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to store byproduct, source and special nuclear materials not intended for use at Columbia Generating Station. The materials shall be no more than 9 sealed neutron radiation sources designed for insertion into pressurized water reactors and no more than 40 sealed beta radiation sources designed for use in area radiation monitors. The total inventory shall not exceed 24 microcuries of strontium-90, 20 microcuries of uranium-235, 30 curies of plutonium-238, and 3 curies of americium-241.

C. This renewed license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1) Maximum Power Level The licensee is authorized to operate the facility at reactor core power levels not in excess of full power (34863544 megawatts thermal).

Renewed License No. NPF-21 Amendment No. 225

Definitions 1.1 1.1 Definitions PHYSICS TESTS (continued)

c. Otherwise approved by the Nuclear Regulatory Commission.

RATED THERMAL POWER RTP shall be a total reactor core heat transfer rate to the (RTP) reactor coolant of 3486 3544 MWt.

REACTOR PROTECTION The RPS RESPONSE TIME shall be that time interval from SYSTEM (RPS) RESPONSE when the monitored parameter exceeds its RPS trip setpoint at TIME the channel sensor until de-energization of the scram pilot valve solenoids. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured.

SHUTDOWN MARGIN (SDM) SDM shall be the amount of reactivity by which the reactor is subcritical or would be subcritical throughout the operating cycle assuming that:

a. The reactor is xenon free;

b. The moderator temperature is 68°F, corresponding to the most reactive state; and

c. All control rods are fully inserted except for the single control rod of highest reactivity worth, which is assumed to be fully withdrawn. With control rods not capable of being fully inserted, the reactivity worth of these control rods must be accounted for in the determination of SDM.

STAGGERED TEST BASIS A STAGGERED TEST BASIS shall consist of the testing of one of the systems, subsystems, channels, or other designated components during the interval specified by the Surveillance Frequency, so that all systems, subsystems, channels, or other designated components are tested during n Surveillance Frequency intervals, where n is the total number of systems, subsystems, channels, or other designated components in the associated function.

THERMAL POWER THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.

Columbia Generating Station 1.1-5 Amendment No.169 225, 228

RPS Instrumentation (After Implementation of PRNM Upgrade) 3.3.1.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. One or more Functions C.1 Restore RPS trip capability. 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months with RPS trip capability not maintained.

D. Required Action and D.1 Enter the Condition Immediately associated Completion referenced in Time of Condition A, B, Table 3.3.1.1-1 for the or C not met. channel.

E. As required by Required E.1 Reduce THERMAL 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Action D.1 and POWER to referenced in < 3029.5% RTP.

Table 3.3.1.1-1.

F. As required by Required F.1 Be in MODE 2. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Action D.1 and referenced in Table 3.3.1.1-1.

G. As required by Required G.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Action D.1 and referenced in Table 3.3.1.1-1.

H. As required by Required H.1 Initiate action to fully insert Immediately Action D.1 and all insertable control rods in referenced in core cells containing one or Table 3.3.1.1-1. more fuel assemblies.

Columbia Generating Station 3.3.1.1-10 Amendment No. 169 225 226

RPS Instrumentation (After Implementation of PRNM Upgrade) 3.3.1.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.1.1.8 Perform CHANNEL FUNCTIONAL TEST. 92 days SR 3.3.1.1.9 Deleted.

SR 3.3.1.1.10 ------------------------------NOTES-----------------------------

1. Neutron detectors are excluded.

2. For Function 1, not required to be performed when entering MODE 2 from MODE 1 until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after entering MODE 2.

3. For Functions 2.b and 2.f, the recirculation flow transmitters that feed the APRMs are included.

Perform CHANNEL CALIBRATION. 18 months for Functions 1, 3, 4, 6, 7, and 9 through 11 AND 24 months for Functions 2, 5, and 8 SR 3.3.1.1.11 Deleted.

SR 3.3.1.1.12 Verify Turbine Throttle Valve - Closure, and Turbine 18 months Governor Valve Fast Closure Trip Oil Pressure -

Low Functions are not bypassed when THERMAL POWER is 29.530% RTP.

SR 3.3.1.1.13 Perform CHANNEL FUNCTIONAL TEST. 24 months Columbia Generating Station 3.3.1.1-13 Amendment No. 179 225 226

RPS Instrumentation (After Implementation of PRNM Upgrade) 3.3.1.1 Table 3.3.1.1-1 (page 1 of 4)

Reactor Protection System Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION D.1 REQUIREMENTS VALUE

1. Intermediate Range Monitors

a. Neutron Flux - High 2 3 G SR 3.3.1.1.1 122/125 SR 3.3.1.1.3 divisions of full SR 3.3.1.1.5 scale SR 3.3.1.1.6 SR 3.3.1.1.10 SR 3.3.1.1.14 (a) 5 3 H SR 3.3.1.1.1 122/125 SR 3.3.1.1.4 divisions of full SR 3.3.1.1.10 scale SR 3.3.1.1.14

b. Inop 2 3 G SR 3.3.1.1.3 NA SR 3.3.1.1.14 (a) 5 3 H SR 3.3.1.1.4 NA SR 3.3.1.1.14

2. Average Power Range Monitors (b)

a. Neutron Flux - High 2 3 G SR 3.3.1.1.1 20% RTP (Setdown) SR 3.3.1.1.6 SR 3.3.1.1.7 (d),(e)

SR 3.3.1.1.10 SR 3.3.1.1.16 (b)

b. Simulated Thermal 1 3 F SR 3.3.1.1.1 0.632W + 642.09%

Power - High SR 3.3.1.1.2 RTP and 114.9%

(c)

SR 3.3.1.1.7 RTP (d),(e)

SR 3.3.1.1.10 SR 3.3.1.1.16 (a) With any control rod withdrawn from a core cell containing one or more fuel assemblies.

(b) Each APRM/OPRM channel provides inputs to both trip systems.

(c) 0.632W + 6059.8% RTP and 114.9% RTP when reset for single loop operation per LCO 3.4.1, Recirculation Loops Operating.

(d) If the as-found channel setpoint is outside its predefined as-found tolerance, then the channel shall be evaluated to verify that it is functioning as required before returning the channel to service.

(e) The instrument channel setpoint shall be reset to a value that is within the as-left tolerance around the Limiting Trip Setpoint (LTSP) at the completion of the surveillance; otherwise, the channel shall be declared inoperable.

Setpoints more conservative than the LTSP are acceptable provided that the as-found and as-left tolerances apply to the actual setpoint implemented in the surveillance procedures (Nominal Trip Setpoint) to confirm channel performance. The LTSP and the methodologies used to determine the as-found and as-left tolerances are specified in the Licensee Controlled Specifications.

Columbia Generating Station 3.3.1.1-15 Amendment No. 169 225 226

RPS Instrumentation (After Implementation of PRNM Upgrade) 3.3.1.1 Table 3.3.1.1-1 (page 4 of 4)

Reactor Protection System Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION D.1 REQUIREMENTS VALUE

8. Turbine Throttle Valve - 29.530% 4 E SR 3.3.1.1.8 7% closed Closure RTP SR 3.3.1.1.10 SR 3.3.1.1.12 SR 3.3.1.1.14 SR 3.3.1.1.15

9. Turbine Governor Valve 3029.5% 2 E SR 3.3.1.1.8 1000 psig Fast Closure, Trip Oil RTP SR 3.3.1.1.10 Pressure - Low SR 3.3.1.1.12 SR 3.3.1.1.14 SR 3.3.1.1.15

10. Reactor Mode Switch - 1,2 2 G SR 3.3.1.1.13 NA Shutdown Position SR 3.3.1.1.14 (a) 5 2 H SR 3.3.1.1.13 SR 3.3.1.1.14 NA

11. Manual Scram 1,2 2 G SR 3.3.1.1.4 NA SR 3.3.1.1.14 (a) 5 2 H SR 3.3.1.1.4 NA SR 3.3.1.1.14 (a) With any control rod withdrawn from a core cell containing one or more fuel assemblies.

Columbia Generating Station 3.3.1.1-18 Amendment No. 225 226 232

EOC-RPT Instrumentation 3.3.4.1 3.3 INSTRUMENTATION 3.3.4.1 End of Cycle Recirculation Pump Trip (EOC-RPT) Instrumentation LCO 3.3.4.1 a. Two channels per trip system for each EOC-RPT instrumentation Function listed below shall be OPERABLE:

1. Turbine Throttle Valve (TTV) - Closure; and

2. Turbine Governor Valve (TGV) Fast Closure, Trip Oil Pressure

- Low.

OR

b. LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR)," limits for inoperable EOC-RPT as specified in the COLR are made applicable.

APPLICABILITY: THERMAL POWER 3029.5% RTP.

ACTIONS

NOTE-----------------------------------------------------------

Separate Condition entry is allowed for each channel.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A.1 Restore channel to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months channels inoperable. OPERABLE status.

OR A.2 ---------------NOTE--------------

Not applicable if inoperable channel is the result of an inoperable breaker.

Place channel in trip. 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Columbia Generating Station 3.3.4.1-1 Amendment No.149,169 225

EOC-RPT Instrumentation 3.3.4.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. One or more Functions B.1 Restore EOC-RPT trip 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months with EOC-RPT trip capability.

capability not maintained. OR AND B.2 Apply the MCPR limit for 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months inoperable EOC-RPT as MCPR limit for specified in the COLR.

inoperable EOC-RPT not made applicable.

C. Required Action and C.1 Remove the associated 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months associated Completion recirculation pump from Time not met. service.

OR C.2 Reduce THERMAL 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months POWER to

< 3029.5% RTP.

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 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months provided the associated Function maintains EOC-RPT trip capability.

SURVEILLANCE FREQUENCY SR 3.3.4.1.1 Perform CHANNEL FUNCTIONAL TEST. 92 days Columbia Generating Station 3.3.4.1-2 Amendment No. 149,169 225

EOC-RPT Instrumentation 3.3.4.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.4.1.2.a Perform CHANNEL CALIBRATION. The Allowable 24 months Value shall be:

TTV - Closure: 7% closed.

SR 3.3.4.1.2.b Perform CHANNEL CALIBRATION. The Allowable 18 months Value shall be:

TGV Fast Closure, Trip Oil Pressure - Low:

1000 psig.

SR 3.3.4.1.3 Verify TTV - Closure and TGV Fast Closure, Trip 18 months Oil Pressure - Low Functions are not bypassed when THERMAL POWER is 3029.5% RTP.

SR 3.3.4.1.4 Perform LOGIC SYSTEM FUNCTIONAL TEST, 24 months including breaker actuation.

SR 3.3.4.1.5 -------------------------------NOTE------------------------------

Breaker arc suppression time may be assumed from the most recent performance of SR 3.3.4.1.6.

Verify the EOC-RPT SYSTEM RESPONSE TIME is 24 months on a within limits. STAGGERED TEST BASIS SR 3.3.4.1.6 Determine RPT breaker arc suppression time. 60 months Columbia Generating Station 3.3.4.1-3 Amendment No. 149,169 225

Primary Containment Isolation Instrumentation 3.3.6.1 Table 3.3.6.1-1 (page 1 of 6)

Primary Containment Isolation Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION C.1 REQUIREMENTS VALUE

1. Main Steam Line Isolation

a. Reactor Vessel 1, 2, 3 2 D SR 3.3.6.1.1 -142.3 inches Water Level - Low SR 3.3.6.1.2 Low Low, Level 1 SR 3.3.6.1.4 SR 3.3.6.1.6 SR 3.3.6.1.7

b. Main Steam Line 1 2 E SR 3.3.6.1.2 804 psig Pressure - Low SR 3.3.6.1.4 SR 3.3.6.1.6 SR 3.3.6.1.7

c. Main Steam Line 1, 2, 3 2 per D SR 3.3.6.1.1 13724.94 psid Flow - High MSL SR 3.3.6.1.2 SR 3.3.6.1.4 SR 3.3.6.1.6 SR 3.3.6.1.7 (a) (a)

d. Condenser Vacuum 1, 2 , 3 2 D SR 3.3.6.1.2 7.2 inches

- Low SR 3.3.6.1.4 Hg vacuum SR 3.3.6.1.6

e. Main Steam Tunnel 1, 2, 3 2 D SR 3.3.6.1.3 170°F Temperature - High SR 3.3.6.1.4 SR 3.3.6.1.6

f. Main Steam Tunnel 1,2,3 2 D SR 3.3.6.1.3 90°F Differential SR 3.3.6.1.4 Temperature - High SR 3.3.6.1.6

g. Manual Initiation 1, 2, 3 4 G SR 3.3.6.1.6 NA

2. Primary Containment Isolation

a. Reactor Vessel 1, 2, 3 2 F SR 3.3.6.1.1 9.5 inches Water Level - Low, SR 3.3.6.1.2 Level 3 SR 3.3.6.1.4 SR 3.3.6.1.6 (a) With any turbine throttle valve not closed.

Columbia Generating Station 3.3.6.1-5 Amendment No. 214,219 225

License Amendment Request to Revise Operating License and Technical Specifications for Measurement Uncertainty Recapture (MUR) Power Uprate Enclosure 3 New Licensee Controlled Specification and Markup of Technical Specification Bases For Information Only Licensee Controlled Specifications Pages 1.3.9-1 1.3.9-2 B 1.3.9-1 through B 1.3.9-5 Technical Specifications Bases Pages B 3.3.1.1-21 B 3.3.1.1-22 B 3.3.1.1-35 B 3.3.2.2-1 B 3.3.4.1-2 B 3.3.4.1-3 B 3.3.4.1-4 B 3.3.4.1-6 B 3.3.4.1-7 B 3.7.6-1

LEFM Feedwater Flow Instrumentation 1.3.9 1.3 INSTRUMENTATION 1.3.9 Leading Edge Flow Meter (LEFM) Feedwater Flow Instrumentation RFO 1.3.9 The LEFM Feedwater Flow Instrumentation System shall be OPERABLE.

APPLICABILITY: THERMAL POWER > 3486 MWt.

COMPENSATORY MEASURES REQUIRED CONDITION COMPENSATORY MEASURE COMPLETION TIME A. Loss of LEFM Meter A.1 Restore LEFM Meter Status 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Status indication. indication.

B. Required Compensatory B.1 Reduce power to Immediately Measure and associated 3486 MWt.

Completion Time of Condition A not met.

C. One or more LEFM C.1 --------------NOTE--------------

feedwater flow meters If current THERMAL not in the Check Plus POWER is < 3537 MWt, the Mode. maximum permissible THERMAL POWER is 3537 MWt.

Return both LEFM 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months feedwater flow meters to the Check Plus Mode.

Columbia Generating Station 1.3.9-1 Revision xx

LEFM Feedwater Flow Instrumentation 1.3.9 COMPENSATORY MEASURES (continued)

REQUIRED CONDITION COMPENSATORY MEASURE COMPLETION TIME D. Required Compensatory D.1 Reduce THERMAL Immediately Measure and associated POWER to 3537 MWt.

Completion Time of Condition C not met. AND D.2 --------------NOTE--------------

Not applicable if one or more LEFM feedwater flow meters are in the Fail Mode.

Verify both LEFM feedwater Once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months flowmeters are in the Check Mode or one LEFM feedwater flowmeter is in the Check Mode and one LEFM feedwater flowmeter is in the Check Plus Mode.

E. One or more LEFM E.1 Reduce THERMAL 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months feedwater flow meters in POWER to 3486 MWt.

the Fail Mode.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 1.3.9.1 -----------------------------NOTE-------------------------------

LEFM feedwater flow meter status is monitored by the plant process computer, which will alarm when the LEFM System is determined to be not in the Check Plus Mode.

Verify each LEFM feedwater flow meter is in the 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Check Plus Mode.

Columbia Generating Station 1.3.9-2 Revision xx

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LEFM Feedwater Flow Instrumentation B 1.3.9 BASES APPLICABILITY THERMAL POWER > 3486 MWt.

COMPENSATORY A.1 MEASURES When the PPC data link fails (total loss of communication between both PPC CPUs and both LEFM CPUs) the LEFM meter status indication must be restored. On loss of data link, the PPC initiates automatic actions to restore the connection and core thermal power calculations revert to using calibrated venturi inputs.

A Completion Time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months from the point that a valid loss of signal is confirmed is reasonable because the feedwater flow venturis are periodically calibrated using the LEFM instrumentation and subsequent venturi drift is small over the Completion Time.

If the cause of the loss of data link between the PPC and LEFM is determined to be due to actual LEFM failure, then Condition E should be entered immediately.

B.1 With Required Compensatory Measure A.1 not met, Required Compensatory Measure B.1 requires that THERMAL POWER be immediately reduced to less than or equal to 3486 MWt. In this Condition, THERMAL POWER uncertainty increases to 2% of 3486 MWt based upon the accuracy of the feedwater flow venturis (Reference 1).

Therefore, THERMAL POWER is reduced to 3486 MWt to ensure that the initial conditions of the safety analyses remain valid. At this point, 3486 MWt is the new maximum THERMAL POWER limit.

C.1, D.1, D.2 and E.1 With one or more LEFM feedwater flow meters not in the Check Plus Mode, Required Compensatory Measure C.1 requires that the affected LEFM feedwater flow meter(s) be restored to the Check Plus Mode.

Columbia Generating Station B 1.3.9-2 Revision xx

LEFM Feedwater Flow Instrumentation B 1.3.9 BASES COMPENSATORY MEASURES (continued)

If both LEFM feedwater flow meters are in the Check Mode or one LEFM feedwater flow meter is in the Check Mode and one LEFM feedwater flow meter is in the Check Plus Mode, the allowed Completion Time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is reasonable since the LEFM feedwater flow instrumentation remains functional in this Condition and allows time for maintenance on the LEFM instrumentation system.

If one or more LEFM feedwater flow meters are in the Fail Mode, the allowed Completion Time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is reasonable because the feedwater flow venturis are periodically calibrated using the LEFM instrumentation and subsequent venturi drift is small over the Completion Time. Note that if one or more LEFM feedwater flow meters are in the Fail Mode, Condition E is also entered concurrently. Required Compensatory Measure E.1 requires reduction in THERMAL POWER within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . Thus, either the LEFM feedwater flow meters are restored per Required Compensatory Measure C.1 or THERMAL POWER is reduced per Required Compensatory Measure E.1 within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months of the LEFM feedwater flow meters in the Fail Mode.

Required Compensatory Measure C.1 is modified by a Note that limits the maximum permissible THERMAL POWER to less than or equal to 3537 MWt. This note addresses the situation when one or more LEFM feedwater flow meters are not in the Check Plus Mode at reduced power levels. This note prohibits returning to RATED THERMAL POWER while in this Condition.

Conditions C and D are structured to ensure that actions are taken within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months of the initial occurrence of an LEFM feedwater flowmeter not in the Check Plus Mode.

Thus, with Required Compensatory Measure C.1 not met, Required Compensatory Measure D.1 requires that THERMAL POWER be immediately reduced. If only Conditions C and D are met and Condition E is not met, then the LEFM feedwater flow meters are either both in the Check Mode or one is in the Check Mode and one is in the Check Plus Mode. Neither LEFM feedwater flow meter is in the Fail Mode. In this case, power must be reduced to less than or equal to 3537 MWt.

With both LEFM feedwater flow meters in the Check Mode or with one LEFM feedwater flow meter in the Check Mode and one LEFM feedwater flow meter in the Check Plus Mode, LEFM uncertainty increases from 0.3% to 0.5% (Reference 2). Therefore, THERMAL POWER must be reduced to 3537 MWt (Reference 3) to ensure that the initial conditions of the safety analyses remain valid. At this point, 3537 MWt is the new maximum THERMAL POWER limit.

Columbia Generating Station B 1.3.9-3 Revision xx

LEFM Feedwater Flow Instrumentation B 1.3.9 BASES COMPENSATORY MEASURES (continued)

Required Compensatory Measure D.2 requires that the LEFM feedwater flow meters be verified to be in either the Check Mode or the Check Plus Mode on a periodic basis. This frequency is reasonable because the LEFM System performs online self-diagnostics to verify that the system operation is within design basis uncertainty limits. Any out-of-specification condition will result in a self-diagnostic alarm condition, either for "alert" status (i.e., increased flow measurement uncertainty) or "failure" status. Required Compensatory Measure D.2 is modified by a Note stating the action is not applicable with one or more LEFM feedwater flow meters in the Fail Mode.

Required Compensatory Measure E.1 requires that with one or more LEFM feedwater flow meters in the Fail Mode, THERMAL POWER must be reduced to less than or equal to 3486 MWt. When one or more LEFM feedwater flow meters are in the Fail Mode, LEFM flow uncertainty cannot be guaranteed (Reference 2). Therefore, THERMAL POWER must be reduced to 3486 MWt (Reference 3) to ensure that the initial conditions of the safety analyses remain valid. At this point, 3486 MWt is the new maximum THERMAL POWER limit.

The allowed Completion Time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is measured from the time that one or more LEFM feedwater flow meters enter the Fail Mode. Note that the allowed Completion Time of Condition C may already be expired if entering the Fail Mode from the Check Mode.

Within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months of this Condition being met, THERMAL POWER must be reduced to 3486 MWt. The allowed Completion Time is reasonable to allow time for maintenance on the LEFM instrumentation system and because the feedwater flow venturis are periodically calibrated using the LEFM instrumentation and subsequent venturi drift is small over the Completion Time.

Columbia Generating Station B 1.3.9-4 Revision xx

LEFM Feedwater Flow Instrumentation B 1.3.9 BASES SURVEILLANCE SR 1.3.9.1 REQUIREMENTS Each LEFM feedwater flow meter must be verified to be in the Check Plus Mode as indicated by the PPC once every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . This frequency is reasonable because the LEFM System performs online self-diagnostics to verify that the system operation is within design basis uncertainty limits.

Any out-of-specification condition will result in a self-diagnostic alarm condition, either for "alert" status (i.e., increased flow measurement uncertainty) or "failure" status. Additionally, if the communications link between the LEFM System and the plant computer fails (i.e., LEFM CPU Link A and B failed), the LEFM flow meter is considered inoperable.

This SR is modified by a Note which states that the LEFM feedwater flow meter status is monitored by the PPC, which will alarm when the LEFM System is determined to be not in the Check Plus Mode.

REFERENCES 1. Nuclear Regulatory Commission (NRC) Regulatory Issue Summary (RIS) 2002-03, "Guidance on the Content of Measurement Uncertainty Recapture Power Uprate Applications," dated January 31, 2002.

2. Cameron (Caldon) document ER-1049, "Bounding Uncertainty Analysis for Thermal Power Determination at Columbia Nuclear Generating Station Using the LEFM ¥+ System, " Revision 3 (Proprietary), dated December 2015.

3. Heat balance calculation NE-02-15-08, Rev 0, Heat Balance Determination for Rated Thermal Power.

Columbia Generating Station B 1.3.9-5 Revision xx

RPS Instrumentation B 3.3.1.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)

This Function must be enabled at THERMAL POWER 3029.5% RTP.

This is accomplished automatically by pressure switches sensing turbine first stage pressure; therefore, opening the turbine bypass valves may affect this Function.

The Turbine Throttle Valve - Closure Allowable Value is selected to detect imminent TTV closure thereby reducing the severity of the subsequent pressure transient.

Eight channels of Turbine Throttle Valve - Closure Function, with four channels in each trip system, are required to be OPERABLE to ensure that no single instrument failure will preclude a scram from this Function if any three TTVs should close. This Function is required, consistent with analysis assumptions, whenever THERMAL POWER is 3029.5% RTP.

This Function is not required when THERMAL POWER is

< 3029.5% RTP since the Reactor Vessel Steam Dome Pressure - High and the Average Power Range Monitor Neutron Flux - High Functions are adequate to maintain the necessary safety margins.

9. Turbine Governor Valve Fast Closure, Trip Oil Pressure - Low Fast closure of the TGVs results in the loss of a heat sink that produces reactor pressure, neutron flux, and heat flux transients that must be limited. Therefore, a reactor scram is initiated on TGV fast closure in anticipation of the transients that would result from the closure of these valves. The Turbine Governor Valve Fast Closure, Trip Oil Pressure -

Low Function is the primary scram signal for the generator load rejection event analyzed in Reference 5. For this event, the reactor scram reduces the amount of energy required to be absorbed and, along with the actions of the EOC-RPT System, ensures that the MCPR SL is not exceeded.

Turbine Governor Valve Fast Closure, Trip Oil Pressure - Low signals are initiated by the digital-electro hydraulic fluid pressure at each governor valve. There is one pressure switch associated with each governor valve, the signal from each switch being assigned to a separate RPS logic channel. This Function must be enabled at THERMAL POWER 3029.5% RTP. This is normally accomplished automatically by pressure switches sensing turbine first stage pressure; therefore, opening the turbine bypass valves may affect this Function. The basis for the setpoint of this automatic bypass is identical to that described for the Turbine Throttle Valve - Closure Function.

Columbia Generating Station B 3.3.1.1-21 Revision 87

RPS Instrumentation B 3.3.1.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)

The Turbine Governor Valve Fast Closure, Trip Oil Pressure - Low Allowable Value is selected high enough to detect imminent TGV fast closure.

Four channels of Turbine Governor Valve Fast Closure, Trip Oil Pressure

- Low Function, with two channels in each trip system arranged in a one-out-of-two logic, are required to be OPERABLE to ensure that no single instrument failure will preclude a scram from this Function on a valid signal. This Function is required, consistent with the analysis assumptions, whenever THERMAL POWER is 3029.5% RTP. This Function is not required when THERMAL POWER is < 3029.5% RTP since the Reactor Vessel Steam Dome Pressure - High and the Average Power Range Monitor Neutron Flux - High Functions are adequate to maintain the necessary safety margins.

10. Reactor Mode Switch - Shutdown Position The Reactor Mode Switch - Shutdown Position Function provides signals, via the manual scram logic channels, that are redundant to the automatic protective instrumentation channels and provide manual reactor trip capability. This Function was not specifically credited in the accident analysis, but it is retained for the overall redundancy and diversity of the RPS as required by the NRC approved licensing basis.

The reactor mode switch is a single switch with four channels (one from each of the four independent banks of contacts), each of which inputs into one of the RPS logic channels.

There is no Allowable Value for this Function since the channels are mechanically actuated based solely on reactor mode switch position.

Four channels of Reactor Mode Switch - Shutdown Position Function, with two channels in each trip system, are available and required to be OPERABLE. The Reactor Mode - Switch Shutdown Position Function is required to be OPERABLE in MODES 1 and 2, and in MODE 5 with any control rod withdrawn from a core cell containing one or more fuel assemblies, since these are the MODES and other specified conditions when control rods are withdrawn.

Columbia Generating Station B 3.3.1.1-22 Revision 87

RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.3.1.1.11 - Not Used SR 3.3.1.1.12 This SR ensures that scrams initiated from the Turbine Throttle Valve -

Closure and Turbine Governor Valve Fast Closure, Trip Oil Pressure -

Low Functions will not be inadvertently bypassed when THERMAL POWER is 3029.5% RTP. This involves calibration of the bypass channels. Adequate margins for the instrument setpoint methodology are incorporated into the Allowable Value and the actual setpoint. Because main turbine bypass flow can affect this setpoint nonconservatively (THERMAL POWER is derived from turbine first stage pressure), the main turbine bypass valves must remain closed during an in-service calibration at THERMAL POWER 3029.5% RTP to ensure that the calibration is valid.

If any bypass channel setpoint is nonconservative (i.e., the Functions are bypassed at 3029.5% RTP, either due to open main turbine bypass valve(s) or other reasons), then the affected Turbine Throttle Valve -

Closure and Turbine Governor Valve Fast Closure, Trip Oil Pressure -

Low Functions are considered inoperable. Alternatively, the bypass channel can be placed in the conservative condition (nonbypass). If placed in the nonbypass condition, this SR is met and the channel is considered OPERABLE.

The Frequency of 18 months is based on engineering judgment and reliability of the components.

SR 3.3.1.1.14 The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required trip logic for a specific channel. The functional testing of control rods, in LCO 3.1.3, "Control Rod OPERABILITY," and SDV vent and drain valves, in LCO 3.1.8, "Scram Discharge Volume (SDV) Vent and Drain Valves," overlaps this Surveillance to provide complete testing of the assumed safety function.

The 24 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance was performed with the reactor at power. Operating experience has shown that these components usually pass the Surveillance when performed at the 24 month Frequency.

Columbia Generating Station B 3.3.1.1-35 Revision 87

Feedwater and Main Turbine High Water Level Trip Instrumentation B 3.3.2.2 B 3.3 INSTRUMENTATION B 3.3.2.2 Feedwater and Main Turbine High Water Level Trip Instrumentation BASES BACKGROUND The feedwater and main turbine high water level trip instrumentation is designed to detect a potential failure of the Feedwater Level Control System that causes excessive feedwater flow.

With excessive feedwater flow, the water level in the reactor vessel rises toward the high water level, Level 8 reference point, causing the trip of the two feedwater pump turbines and the main turbine.

Reactor Vessel Water Level - High, Level 8 signals are provided by level sensors that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level in the reactor vessel (variable leg). Three channels of Reactor Vessel Water Level - High, Level 8 instrumentation are provided as input to a two-out-of-three initiation logic that trips the two feedwater pump turbines and the main turbine. The channels include electronic equipment (e.g., trip relays) that compares measured input signals with pre-established setpoints. When the setpoint is exceeded, the channel outputs a main feedwater and main turbine trip signal to the trip logic.

A trip of the feedwater pump turbines limits further increase in reactor vessel water level by limiting further addition of feedwater to the reactor vessel. A trip of the main turbine and closure of the throttle valves protects the turbine from damage due to water entering the turbine.

APPLICABLE The feedwater and main turbine high water level trip instrumentation is SAFETY assumed to be capable of providing a turbine trip in the design basis ANALYSES transient analysis for a feedwater controller failure, maximum demand event (Ref. 1). The Level 8 trip indirectly initiates a reactor scram from the main turbine trip (above 3029.5% RTP) and trips the feedwater pumps, thereby terminating the event. The reactor scram mitigates the reduction in MCPR.

Feedwater and main turbine high water level trip instrumentation satisfies Criterion 3 of Reference 2.

LCO The LCO requires three channels of the Reactor Vessel Water Level -

High, Level 8 instrumentation to be OPERABLE to ensure that no single instrument failure will prevent the feedwater pump turbines and main turbine trip on a valid Level 8 signal. Two of the three channels are needed to provide trip signals in order for the feedwater and main turbine trips to occur. Each channel must have its setpoint set within the specified Allowable Value of SR 3.3.2.2.3. The Allowable Value is set to Columbia Generating Station B 3.3.2.2-1 Revision 73

EOC-RPT Instrumentation B 3.3.4.1 BASES APPLICABLE The TTV - Closure and the TGV Fast Closure, Trip Oil Pressure - Low SAFETY Functions are designed to trip the recirculation pumps in the event of a ANALYSES, LCO, turbine trip or generator load rejection to mitigate the neutron flux, heat and APPLICABILITY flux and pressurization transients, and to increase the margin to the MCPR SL. The analytical methods and assumptions used in evaluating the turbine trip and generator load rejection, as well as other safety analyses that assume EOC-RPT, are summarized in References 2 and 3.

To mitigate pressurization transient effects, the EOC-RPT must trip the recirculation pumps after initiation of initial closure movement of either the TTVs or the TGVs. The combined effects of this trip and a scram reduce fuel bundle power more rapidly than does a scram alone, resulting in an increased margin to the MCPR SL. Alternatively, MCPR limits for an inoperable EOC-RPT as specified in the COLR are sufficient to mitigate pressurization transient effects. The EOC-RPT function is automatically disabled when THERMAL POWER, as sensed by turbine first stage pressure, is < 29.530% RTP.

EOC-RPT instrumentation satisfies Criterion 3 of Reference 4.

The OPERABILITY of the EOC-RPT is dependent on the OPERABILITY of the individual instrumentation channel Functions. Each Function must have a required number of OPERABLE channels in each trip system, with their setpoints within the specified Allowable Value of SR 3.3.4.1.2. The actual setpoint is calibrated consistent with applicable setpoint methodology assumptions. Channel OPERABILITY also includes the associated EOC-RPT breakers. Each channel (including the associated EOC-RPT breakers) must also respond within its assumed response time.

Allowable Values are specified for each EOC-RPT Function specified in the LCO. Nominal trip setpoints are specified in the setpoint calculations.

The nominal setpoints are selected to ensure the setpoints do not exceed the Allowable Value between successive CHANNEL CALIBRATIONS.

Operation with a trip setpoint less conservative than the nominal trip setpoint, but within its Allowable Value, is acceptable. A channel is inoperable if its actual trip setpoint is not within its required Allowable Value. Trip setpoints are those predetermined values of output at which an action should take place. The setpoints are compared to the actual process parameter (e.g., TGV digital-electro hydraulic (DEH) pressure),

and when the measured output value of the process parameter exceeds the setpoint, the associated device (e.g., trip relay) changes state. The analytic limits are derived from the limiting values of the process parameters obtained from the safety analysis. The Allowable Values are derived from the analytic limits, corrected for process and all instrument uncertainties, except drift and calibration. The trip setpoints are derived Columbia Generating Station B 3.3.4.1-2 Revision 73

EOC-RPT Instrumentation B 3.3.4.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) from the analytic limits, corrected for process and all instrument uncertainties, including drift and calibration. The trip setpoints derived in this manner provide adequate protection because all instrumentation uncertainties and process effects are taken into account.

The specific Applicable Safety Analysis, LCO, and Applicability discussions are listed below on a Function by Function basis.

Alternately, since this instrumentation protects against a MCPR SL violation with the instrumentation inoperable, modifications to the MCPR limits (LCO 3.2.2) may be applied to allow this LCO to be met. The MCPR penalty for the condition EOC-RPT inoperable is specified in the COLR.

Turbine Throttle Valve - Closure Closure of the TTVs and a main turbine trip result in the loss of a heat sink that produces reactor pressure, neutron flux, and heat flux transients that must be limited. Therefore, an RPT is initiated on TTV - Closure in anticipation of the transients that would result from closure of these valves. EOC-RPT decreases reactor power and aids the reactor scram in ensuring the MCPR SL is not exceeded during the worst case transient.

Closure of the TTVs is determined by measuring the position of each throttle valve. While there are two separate position switches associated with each throttle valve, only the signal from one switch for each TTV is used, with each of the four channels being assigned to a separate trip channel. The logic for the TTV - Closure Function is such that two or more TTVs must be closed to produce an EOC-RPT. This Function must be enabled at THERMAL POWER 3029.5% RTP. This is normally accomplished automatically by pressure switches sensing turbine first stage pressure; therefore, opening of the turbine bypass valves may affect this Function. Four channels of TTV - Closure, with two channels in each trip system, are available and required to be OPERABLE to ensure that no single instrument failure will preclude an EOC-RPT from this Function on a valid signal. The TTV - Closure Allowable Value is selected to detect imminent TTV closure.

This protection is required, consistent with the safety analysis assumptions, whenever THERMAL POWER is 3029.5% RTP. Below 3029.5% RTP, the Reactor Vessel Steam Dome Pressure - High and the Average Power Range Monitor (APRM) Neutron Flux - High Functions of the Reactor Protection System (RPS) are adequate to maintain the necessary safety margins.

Columbia Generating Station B 3.3.4.1-3 Revision 87

EOC-RPT Instrumentation B 3.3.4.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)

TGV Fast Closure, Trip Oil Pressure - Low Fast closure of the TGVs during a generator load rejection results in the loss of a heat sink that produces reactor pressure, neutron flux, and heat flux transients that must be limited. Therefore, an RPT is initiated on TGV Fast Closure, Trip Oil Pressure - Low in anticipation of the transients that would result from the closure of these valves. The EOC-RPT decreases reactor power and aids the reactor scram in ensuring that the MCPR SL is not exceeded during the worst case transient.

Fast closure of the TGVs is determined by measuring the DEH fluid pressure at each control valve. There is one pressure switch associated with each control valve, and the signal from each switch is assigned to a separate trip channel. The logic for the TGV Fast Closure, Trip Oil Pressure - Low Function is such that two or more TGVs must be closed (pressure switch trips) to produce an EOC-RPT. This Function must be enabled at THERMAL POWER 3029.5% RTP. This is normally accomplished automatically by pressure switches sensing turbine first stage pressure; therefore, opening of the turbine bypass valves may affect this Function. Four channels of TGV Fast Closure, Trip Oil Pressure - Low, with two channels in each trip system, are available and required to be OPERABLE to ensure that no single instrument failure will preclude an EOC-RPT from this Function on a valid signal. The TGV Fast Closure, Trip Oil Pressure - Low Allowable Value is selected high enough to detect imminent TGV fast closure.

This protection is required consistent with the analysis, whenever the THERMAL POWER is 3029.5% RTP. Below 3029.5% RTP, the Reactor Vessel Steam Dome Pressure - High and the APRM Neutron Flux - High Functions of the RPS are adequate to maintain the necessary safety margins. The turbine first stage pressure/reactor power relationship for the setpoint of the automatic enable is identical to that described for TTV closure.

ACTIONS A Note has been provided to modify the ACTIONS related to EOC-RPT instrumentation channels. Section 1.3, Completion Times, specifies that once a Condition has been entered, subsequent divisions, subsystems, components, or variables expressed in the Condition, discovered to be inoperable or not within limits, will not result in separate entry into the Condition. Section 1.3 also specifies that Required Actions of the Columbia Generating Station B 3.3.4.1-4 Revision 87

EOC-RPT Instrumentation B 3.3.4.1 BASES ACTIONS (continued) on a valid signal and both recirculation pumps can be tripped. This requires two channels of the Function, in the same trip system, to each be OPERABLE or in trip, and the associated drive motor breakers to be OPERABLE or in trip. Alternatively, Required Action B.2 requires the MCPR limit for inoperable EOC-RPT, as specified in the COLR, to be applied. This also restores the margin to MCPR assumed in the safety analysis.

The 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Completion Time is sufficient for the operator to take corrective action, and takes into account the likelihood of an event requiring actuation of the EOC-RPT instrumentation during this period. It is also consistent with the 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Completion Time provided in LCO 3.2.2, Required Action A.1, since this instrumentation's purpose is to preclude a MCPR violation.

C.1 and C.2 With any Required Action and associated Completion Time not met, THERMAL POWER must be reduced to < 3029.5% RTP within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

Alternately, the associated recirculation pump may be removed from service since this performs the intended function of the instrumentation.

The allowed Completion Time of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is reasonable, based on operating experience, to reduce THERMAL POWER to < 3029.5% RTP from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE The Surveillances are modified by a Note to indicate that when a channel REQUIREMENTS 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 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , provided the associated Function maintains EOC-RPT trip capability. Upon completion of the Surveillance, or expiration of the 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken. This Note is based on the reliability analysis (Ref. 5) assumption of the average time required to perform channel surveillance.

That analysis demonstrated that the 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months testing allowance does not significantly reduce the probability that the recirculation pumps will trip when necessary.

Columbia Generating Station B 3.3.4.1-6 Revision 73

EOC-RPT Instrumentation B 3.3.4.1 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.3.4.1.1 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the channel will perform the intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology.

The Frequency of 92 days is based on reliability analysis (Ref. 5).

SR 3.3.4.1.2 CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations consistent with the plant specific setpoint methodology.

The Frequency for SR 3.3.4.1.2.b is based upon the assumption of an 18 month calibration interval, in the determination of the magnitude of equipment drift in the setpoint analysis.

A Frequency of 24 months is assumed for SR 3.3.4.1.2.a because the TTV position switches are not susceptible to instrument drift.

SR 3.3.4.1.3 This SR ensures that an EOC-RPT initiated from the TTV - Closure and TGV Fast Closure, Trip Oil Pressure - Low Functions will not be inadvertently bypassed when THERMAL POWER is 3029.5% RTP.

This involves calibration of the bypass channels. Adequate margins for the instrument setpoint methodologies are incorporated into the actual setpoint. Because main turbine bypass flow can affect this setpoint nonconservatively (THERMAL POWER is derived from first stage pressure), the main turbine bypass valves must remain closed during an in-service calibration at THERMAL POWER 3029.5% RTP to ensure that the calibration is valid. If any bypass channel's setpoint is nonconservative (i.e., the Functions are bypassed at 3029.5% RTP either due to open main turbine bypass valves or other reasons), the affected TTV - Closure and TGV Fast Closure, Trip Oil Pressure - Low Functions are considered inoperable. Alternatively, the bypass channel can be placed in the conservative condition (nonbypass). If placed in the nonbypass condition, this SR is met and the channel considered OPERABLE.

Columbia Generating Station B 3.3.4.1-7 Revision 73

Main Turbine Bypass System B 3.7.6 B 3.7 PLANT SYSTEMS B 3.7.6 Main Turbine Bypass System BASES BACKGROUND The Main Turbine Bypass System is designed to control steam pressure when reactor steam generation exceeds turbine requirements during unit startup, sudden load reduction, and cooldown. It allows excess steam flow from the reactor to the condenser without going through the turbine.

The bypass capacity of the system is 2523.35% of the Nuclear Steam Supply System rated steam flow. Sudden load reductions within the capacity of the steam bypass can be accommodated without reactor scram. The Main Turbine Bypass System consists of a four valve manifold connected to the main steam lines between the main steam isolation valves and the turbine throttle valves. Each of these valves is sequentially operated by hydraulic cylinders. The bypass valves are controlled by the pressure regulation function of the Digital-Electro Hydraulic Control System, as discussed in the FSAR, Section 7.7.1.5 (Ref. 1). The bypass valves are normally closed, and the pressure regulator controls the turbine control valves, directing all steam flow to the turbine. If the speed governor or the load limiter restricts steam flow to the turbine, the pressure regulator controls the system pressure by opening the bypass valves. When the bypass valves open, the steam flows from the valve manifold, through connecting piping, to the pressure-reducing perforated pipes located in the condenser shell.

APPLICABLE The Main Turbine Bypass System is assumed to function during the SAFETY design basis feedwater controller failure, maximum demand event, ANALYSES described in the FSAR, Section 15.1.2 (Ref. 2). Opening the bypass valves during the pressurization event mitigates the increase in reactor vessel pressure, which affects the MCPR during the event. An inoperable Main Turbine Bypass System may result in an MCPR penalty.

The Main Turbine Bypass System satisfies Criterion 3 of Reference 3.

LCO The Main Turbine Bypass System is required to be OPERABLE to limit peak pressure in the main steam lines and maintain reactor pressure within acceptable limits during events that cause rapid pressurization, such that the Safety Limit MCPR is not exceeded. With the Main Turbine Bypass System inoperable, modifications to the MCPR limits (LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR)") may be applied to allow continued operation.

An OPERABLE Main Turbine Bypass System requires the bypass valves to open in response to increasing main steam line pressure. This response is within the assumptions of the applicable analysis (Ref. 2).

The MCPR limit for the inoperable Main Turbine Bypass System is specified in the COLR.

Columbia Generating Station B 3.7.6-1 Revision 73

License Amendment Request to Revise Operating License and Technical Specifications for Measurement Uncertainty Recapture (MUR) Power Uprate Enclosure 4 Revised (Clean) Renewed Facility Operating License and Technical Specifications Pages Renewed Facility Operating License Page 3 Technical Specifications Pages 1.1-5 3.3.1.1-10 3.3.1.1-13 3.3.1.1-15 3.3.1.1-18 3.3.4.1-1 3.3.4.1-2 3.3.4.1-3 3.3.6.1-5

(2) Pursuant to the Act and 10 CFR Part 70, to receive, possess and use at any time special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Final Safety Analysis Report, as supplemented and amended; (3) Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4) Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use in amounts as required any byproduct, source of special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (5) Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

(6) Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to store byproduct, source and special nuclear materials not intended for use at Columbia Generating Station. The materials shall be no more than 9 sealed neutron radiation sources designed for insertion into pressurized water reactors and no more than 40 sealed beta radiation sources designed for use in area radiation monitors. The total inventory shall not exceed 24 microcuries of strontium-90, 20 microcuries of uranium-235, 30 curies of plutonium-238, and 3 curies of americium-241.

C. This renewed license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1) Maximum Power Level The licensee is authorized to operate the facility at reactor core power levels not in excess of full power (3544 megawatts thermal).

Renewed License No. NPF-21 Amendment No. 225

Definitions 1.1 1.1 Definitions PHYSICS TESTS (continued)

c. Otherwise approved by the Nuclear Regulatory Commission.

RATED THERMAL POWER RTP shall be a total reactor core heat transfer rate to the (RTP) reactor coolant of 3544 MWt.

REACTOR PROTECTION The RPS RESPONSE TIME shall be that time interval from SYSTEM (RPS) RESPONSE when the monitored parameter exceeds its RPS trip setpoint at TIME the channel sensor until de-energization of the scram pilot valve solenoids. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured.

SHUTDOWN MARGIN (SDM) SDM shall be the amount of reactivity by which the reactor is subcritical or would be subcritical throughout the operating cycle assuming that:

a. The reactor is xenon free;

b. The moderator temperature is 68°F, corresponding to the most reactive state; and

c. All control rods are fully inserted except for the single control rod of highest reactivity worth, which is assumed to be fully withdrawn. With control rods not capable of being fully inserted, the reactivity worth of these control rods must be accounted for in the determination of SDM.

STAGGERED TEST BASIS A STAGGERED TEST BASIS shall consist of the testing of one of the systems, subsystems, channels, or other designated components during the interval specified by the Surveillance Frequency, so that all systems, subsystems, channels, or other designated components are tested during n Surveillance Frequency intervals, where n is the total number of systems, subsystems, channels, or other designated components in the associated function.

THERMAL POWER THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.

Columbia Generating Station 1.1-5 Amendment No.169 225, 228

RPS Instrumentation (After Implementation of PRNM Upgrade) 3.3.1.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. One or more Functions C.1 Restore RPS trip capability. 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months with RPS trip capability not maintained.

D. Required Action and D.1 Enter the Condition Immediately associated Completion referenced in Time of Condition A, B, Table 3.3.1.1-1 for the or C not met. channel.

E. As required by Required E.1 Reduce THERMAL 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Action D.1 and POWER to < 29.5% RTP.

referenced in Table 3.3.1.1-1.

F. As required by Required F.1 Be in MODE 2. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Action D.1 and referenced in Table 3.3.1.1-1.

G. As required by Required G.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Action D.1 and referenced in Table 3.3.1.1-1.

H. As required by Required H.1 Initiate action to fully insert Immediately Action D.1 and all insertable control rods in referenced in core cells containing one or Table 3.3.1.1-1. more fuel assemblies.

Columbia Generating Station 3.3.1.1-10 Amendment No. 169 225 226

RPS Instrumentation (After Implementation of PRNM Upgrade) 3.3.1.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.1.1.8 Perform CHANNEL FUNCTIONAL TEST. 92 days SR 3.3.1.1.9 Deleted.

SR 3.3.1.1.10 ------------------------------NOTES-----------------------------

1. Neutron detectors are excluded.

2. For Function 1, not required to be performed when entering MODE 2 from MODE 1 until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after entering MODE 2.

3. For Functions 2.b and 2.f, the recirculation flow transmitters that feed the APRMs are included.

Perform CHANNEL CALIBRATION. 18 months for Functions 1, 3, 4, 6, 7, and 9 through 11 AND 24 months for Functions 2, 5, and 8 SR 3.3.1.1.11 Deleted.

SR 3.3.1.1.12 Verify Turbine Throttle Valve - Closure, and Turbine 18 months Governor Valve Fast Closure Trip Oil Pressure -

Low Functions are not bypassed when THERMAL POWER is 29.5% RTP.

SR 3.3.1.1.13 Perform CHANNEL FUNCTIONAL TEST. 24 months Columbia Generating Station 3.3.1.1-13 Amendment No. 179 225 226

RPS Instrumentation (After Implementation of PRNM Upgrade) 3.3.1.1 Table 3.3.1.1-1 (page 1 of 4)

Reactor Protection System Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION D.1 REQUIREMENTS VALUE

1. Intermediate Range Monitors

a. Neutron Flux - High 2 3 G SR 3.3.1.1.1 122/125 SR 3.3.1.1.3 divisions of full SR 3.3.1.1.5 scale SR 3.3.1.1.6 SR 3.3.1.1.10 SR 3.3.1.1.14 (a) 5 3 H SR 3.3.1.1.1 122/125 SR 3.3.1.1.4 divisions of full SR 3.3.1.1.10 scale SR 3.3.1.1.14

b. Inop 2 3 G SR 3.3.1.1.3 NA SR 3.3.1.1.14 (a) 5 3 H SR 3.3.1.1.4 NA SR 3.3.1.1.14

2. Average Power Range Monitors (b)

a. Neutron Flux - High 2 3 G SR 3.3.1.1.1 20% RTP (Setdown) SR 3.3.1.1.6 SR 3.3.1.1.7 (d),(e)

SR 3.3.1.1.10 SR 3.3.1.1.16 (b)

b. Simulated Thermal 1 3 F SR 3.3.1.1.1 0.62W + 62.9% RTP (c)

Power - High SR 3.3.1.1.2 and 114.9% RTP SR 3.3.1.1.7 (d),(e)

SR 3.3.1.1.10 SR 3.3.1.1.16 (a) With any control rod withdrawn from a core cell containing one or more fuel assemblies.

(b) Each APRM/OPRM channel provides inputs to both trip systems.

(c) 0.62W + 59.8% RTP and 114.9% RTP when reset for single loop operation per LCO 3.4.1, Recirculation Loops Operating.

(d) If the as-found channel setpoint is outside its predefined as-found tolerance, then the channel shall be evaluated to verify that it is functioning as required before returning the channel to service.

(e) The instrument channel setpoint shall be reset to a value that is within the as-left tolerance around the Limiting Trip Setpoint (LTSP) at the completion of the surveillance; otherwise, the channel shall be declared inoperable.

Setpoints more conservative than the LTSP are acceptable provided that the as-found and as-left tolerances apply to the actual setpoint implemented in the surveillance procedures (Nominal Trip Setpoint) to confirm channel performance. The LTSP and the methodologies used to determine the as-found and as-left tolerances are specified in the Licensee Controlled Specifications.

Columbia Generating Station 3.3.1.1-15 Amendment No. 169 225 226

RPS Instrumentation (After Implementation of PRNM Upgrade) 3.3.1.1 Table 3.3.1.1-1 (page 4 of 4)

Reactor Protection System Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION D.1 REQUIREMENTS VALUE

8. Turbine Throttle Valve - 29.5% RTP 4 E SR 3.3.1.1.8 7% closed Closure SR 3.3.1.1.10 SR 3.3.1.1.12 SR 3.3.1.1.14 SR 3.3.1.1.15

9. Turbine Governor Valve 29.5% RTP 2 E SR 3.3.1.1.8 1000 psig Fast Closure, Trip Oil SR 3.3.1.1.10 Pressure - Low SR 3.3.1.1.12 SR 3.3.1.1.14 SR 3.3.1.1.15

10. Reactor Mode Switch - 1,2 2 G SR 3.3.1.1.13 NA Shutdown Position SR 3.3.1.1.14 (a) 5 2 H SR 3.3.1.1.13 SR 3.3.1.1.14 NA

11. Manual Scram 1,2 2 G SR 3.3.1.1.4 NA SR 3.3.1.1.14 (a) 5 2 H SR 3.3.1.1.4 NA SR 3.3.1.1.14 (a) With any control rod withdrawn from a core cell containing one or more fuel assemblies.

Columbia Generating Station 3.3.1.1-18 Amendment No. 225 226 232

EOC-RPT Instrumentation 3.3.4.1 3.3 INSTRUMENTATION 3.3.4.1 End of Cycle Recirculation Pump Trip (EOC-RPT) Instrumentation LCO 3.3.4.1 a. Two channels per trip system for each EOC-RPT instrumentation Function listed below shall be OPERABLE:

1. Turbine Throttle Valve (TTV) - Closure; and

2. Turbine Governor Valve (TGV) Fast Closure, Trip Oil Pressure

- Low.

OR

b. LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR)," limits for inoperable EOC-RPT as specified in the COLR are made applicable.

APPLICABILITY: THERMAL POWER 29.5% RTP.

ACTIONS

NOTE-----------------------------------------------------------

Separate Condition entry is allowed for each channel.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A.1 Restore channel to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months channels inoperable. OPERABLE status.

OR A.2 ---------------NOTE--------------

Not applicable if inoperable channel is the result of an inoperable breaker.

Place channel in trip. 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Columbia Generating Station 3.3.4.1-1 Amendment No.149,169 225

EOC-RPT Instrumentation 3.3.4.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. One or more Functions B.1 Restore EOC-RPT trip 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months with EOC-RPT trip capability.

capability not maintained. OR AND B.2 Apply the MCPR limit for 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months inoperable EOC-RPT as MCPR limit for specified in the COLR.

inoperable EOC-RPT not made applicable.

C. Required Action and C.1 Remove the associated 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months associated Completion recirculation pump from Time not met. service.

OR C.2 Reduce THERMAL 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months POWER to < 29.5% RTP.

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 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months provided the associated Function maintains EOC-RPT trip capability.

SURVEILLANCE FREQUENCY SR 3.3.4.1.1 Perform CHANNEL FUNCTIONAL TEST. 92 days Columbia Generating Station 3.3.4.1-2 Amendment No. 149,169 225

EOC-RPT Instrumentation 3.3.4.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.4.1.2.a Perform CHANNEL CALIBRATION. The Allowable 24 months Value shall be:

TTV - Closure: 7% closed.

SR 3.3.4.1.2.b Perform CHANNEL CALIBRATION. The Allowable 18 months Value shall be:

TGV Fast Closure, Trip Oil Pressure - Low:

1000 psig.

SR 3.3.4.1.3 Verify TTV - Closure and TGV Fast Closure, Trip 18 months Oil Pressure - Low Functions are not bypassed when THERMAL POWER is 29.5% RTP.

SR 3.3.4.1.4 Perform LOGIC SYSTEM FUNCTIONAL TEST, 24 months including breaker actuation.

SR 3.3.4.1.5 -------------------------------NOTE------------------------------

Breaker arc suppression time may be assumed from the most recent performance of SR 3.3.4.1.6.

Verify the EOC-RPT SYSTEM RESPONSE TIME is 24 months on a within limits. STAGGERED TEST BASIS SR 3.3.4.1.6 Determine RPT breaker arc suppression time. 60 months Columbia Generating Station 3.3.4.1-3 Amendment No. 149,169 225

Primary Containment Isolation Instrumentation 3.3.6.1 Table 3.3.6.1-1 (page 1 of 6)

Primary Containment Isolation Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION C.1 REQUIREMENTS VALUE

1. Main Steam Line Isolation

a. Reactor Vessel 1, 2, 3 2 D SR 3.3.6.1.1 -142.3 inches Water Level - Low SR 3.3.6.1.2 Low Low, Level 1 SR 3.3.6.1.4 SR 3.3.6.1.6 SR 3.3.6.1.7

b. Main Steam Line 1 2 E SR 3.3.6.1.2 804 psig Pressure - Low SR 3.3.6.1.4 SR 3.3.6.1.6 SR 3.3.6.1.7

c. Main Steam Line 1, 2, 3 2 per D SR 3.3.6.1.1 137.9 psid Flow - High MSL SR 3.3.6.1.2 SR 3.3.6.1.4 SR 3.3.6.1.6 SR 3.3.6.1.7 (a) (a)

d. Condenser Vacuum 1, 2 , 3 2 D SR 3.3.6.1.2 7.2 inches

- Low SR 3.3.6.1.4 Hg vacuum SR 3.3.6.1.6

e. Main Steam Tunnel 1, 2, 3 2 D SR 3.3.6.1.3 170°F Temperature - High SR 3.3.6.1.4 SR 3.3.6.1.6

f. Main Steam Tunnel 1,2,3 2 D SR 3.3.6.1.3 90°F Differential SR 3.3.6.1.4 Temperature - High SR 3.3.6.1.6

g. Manual Initiation 1, 2, 3 4 G SR 3.3.6.1.6 NA

2. Primary Containment Isolation

a. Reactor Vessel 1, 2, 3 2 F SR 3.3.6.1.1 9.5 inches Water Level - Low, SR 3.3.6.1.2 Level 3 SR 3.3.6.1.4 SR 3.3.6.1.6 (a) With any turbine throttle valve not closed.

Columbia Generating Station 3.3.6.1-5 Amendment No. 214,219 225

License Amendment Request to Revise Operating License and Technical Specifications for Measurement Uncertainty Recapture (MUR) Power Uprate Enclosure 5 Regulatory Issue Summary (RIS) 2002-03 Cross Reference

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License Amendment Request to Revise Operating License and Technical Specifications for Measurement Uncertainty Recapture (MUR) Power Uprate Enclosure 6 Summary of Regulatory Commitments

Enclosure 6 Summary of Regulatory Commitments COMMITED ONE-TIME ON-GOING COMMITMENT DATE OR ACTION COMMITMENT OUTAGE (Yes/No) (YES/NO)

Limitations regarding the operation Prior to use 1 with an inoperable LEFM system will above 3486 NO YES be included in the LCS. MWt Necessary operating procedure revisions (including Emergency Prior to use Operating Procedures and Abnormal 2 above 3486 NO YES Operating Procedures) will be MWt completed prior to implementation of the proposed LEFM power uprate.

The plant simulator will be modified for the uprated conditions and the Prior to use 3 changes will be validated in above 3486 YES NO accordance with plant configuration MWt control processes.

Operator training will be completed Prior to use 4 prior to implementation of the above 3486 YES NO proposed LEFM power uprate. MWt Plant testing for the proposed changes Prior to use 5 will be completed as described in above 3486 YES NO Section 10.4, Testing of Enclosure 7 MWt The plant process computer software will have a visual alarm at the Reactor Prior to use Operator and Control Room 6 above 3486 YES NO Supervisor station displays to signal MWt the operators to changes in the LEFM status.

License Amendment Request to Revise Operating License and Technical Specifications for Measurement Uncertainty Recapture (MUR) Power Uprate Enclosure 8 Affidavits from GEH and the Electric Power Research Institute (EPRI) Supporting the Withholding of Information in Enclosure 7 from Public Disclosure

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1'='~1211 ELECTRIC POWER

~*- RESEARCH INSTITUTE NEil WllMSHURST Vice President and Chief Nuclear Officer Ref. EPRI Project Number 669 June 14, 2016 Document Control Desk Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

Subject:

Request for Withholding of the following Proprietary Information Included in:

Columbia Generating Station, Docket NO. 50-397 License Amendment Request to Revise Operating License and Technical Specifications for Measurement Uncertainty Recapture (MUR) Power Uprate, included in NEDC 33853P "Safety Analysis Report for Columbia Generating Station Thermal Power Optimization", Class II (GEH Proprietary Information), June 2016 To Whom It May Concern:

This is a request under 10 C.F.R. §2.390(a)(4) that the U.S. Nuclear Regulatory Commission ("NRC") withhold from public disclosure the report identified in the enclosed Affidavit consisting of the proprietary information owned by Electric Power Research Institute, Inc. ("EPRI") identified in the attached report. Proprietary and non-proprietary versions of the Report and the Affidavit in support of this request are enclosed.

EPRI desires to disclose the Proprietary Information in confidence to assist the NRG review of the enclosed submittal to the NRC by Energy Northwest. The Proprietary Information is not to be divulged to anyone outside of the NRC or to any of its contractors, nor shall any copies be made of the Proprietary Information provided herein.

EPRI welcomes any discussions and/or questions relating to the information enclosed.

If you have any questions about the legal aspects of this request for withholding, please do not hesitate to contact me at (704) 595-2732. Questions on the content of the Report should be directed to Andy McGehee of EPRI at (704) 502-6440.

Attachment(s) c: Sheldon Stuchell, NRC (sheldon.stuchell@nrc.gov)

Together . .. Shaping th e Future of Electri city*

1300 West W.T. Harris Boulevard, Charlotte, NC 28262-8550 USA* 704.595.2732 *Mobile 704 .490.2653

nwilmshu rst@epri.com

EI~121 -

I ELECTRIC POWER RESEARCH INSTITUTE AFFIDAVIT RE: Request for Withholding of the Following Proprietary Information Included In:

Columbia Generating Station, Docket NO. 50-397 License Amendment Request to Revise Operating License and Technical Specifications for Measurement Uncertainty Recapture (MUR) Power Uprate, included in NEDC 33853P "Safety Analysis Report for Columbia Generating Station Thermal Power Optimization", Class fl (GEH Proprietary Information), June 2016 I, Neil Wilmshurst, being duly sworn, depose and state as follows:

I am the Vice President and Chief Nuclear Officer at Electric Power Research Institute, Inc. whose principal office is located at 1300 W WT Harris Blvd, Charlotte, NC. ("EPRI") and I have been specifically delegated responsibility for the above-listed report that contains EPRI Proprietary Information that is sought under this Affidavit to be withheld "Proprietary Information". I am authorized to apply to the U.S. Nuclear Regulatory Commission ("NRC") for the withholding of the Proprietary Information on behalf of EPRI.

EPRI Proprietary Information is identified in the above referenced report by a solid underline with highlighted text, inside double brackets. An example of such identification is as follows:

((This*sentence isaiiexample.tEJ))

Tables containing EPRI Proprietary Information are identified with double brackets before and after the object.

In eqch case the superscript notation {El refers to this affidavit and all the bases included below, which provide the reasons for the proprietary determination.

EPRI requests that the Proprietary Information be withheld from the public on the following bases:

Withholding Based Upon Privileged And Confidential Trade Secrets Or Commercial Or Financial Information (see e.g., 10 C.F.R. § 2.390(a)(4):

a. The Proprietary Information is owned by EPRI and has been held in confidence by EPRI. All entities accepting copies of the Proprietary Information do so subject to written agreements imposing an obligation upon the recipient to maintain the confidentiality of the Proprietary Information. The Proprietary Information is disclosed only to parties who agree, in writing, to preserve the confidentiality thereof.

b. EPRI considers the Proprietary Information contained therein to constitute trade secrets of EPRI. As such, EPRI holds the Information in confidence and disclosure thereof is strictly limited to individuals and entities who have agreed, in writing, to maintain the confidentiality of the Information.

c. The information sought to be withheld is considered to be proprietary for the following reasons. EPRI made a substantial economic investment to develop the Proprietary Information and, by prohibiting public disclosure, EPRI derives an economic benefit in the form of licensing royalties and other additional fees from the confidential nature of the Proprietary Information. If the Proprietary Information were publicly available to consultants and/or other businesses providing services in the electric and/or nuclear power industry, they would

be able to use the Proprietary Information for their own commercial benefit and profit and without expending the substantial economic resources required of EPRI to develop the Proprietary Information.

d. EPRl's classification of the Proprietary Information as trade secrets is justified by the Uniform Trade Secrets Act which California adopted in 1984 and a version of which has been adopted by over forty states. The California Uniform Trade Secrets Act, California Civil Code §§3426 - 3426.11, defines a "trade secret" as follows:

"'Trade secret' means information, including a formula, pattern, compilation, program device, method, technique, or process, that:

(1) Derives independent economic value, actual or potential, from not being generally known to the public or to other persons who can obtain economic value from its disclosure or use; and (2) Is the subject of efforts that are reasonable under the circumstances to maintain its secrecy."

e. The Proprietary Information contained therein are not generally known or available to the public. EPRI developed the Information only after making a determination that the Proprietary Information was not available from public sources. EPRI made a substantial investment of both money and employee hours in the development of the Proprietary Information. EPRI was required to devote these resources and effort to derive the Proprietary Information. As a result of such effort and cost, both in terms of dollars spent and dedicated employee time, the Proprietary Information is highly valuable to EPRI.

f. A public dlsclosure of the Proprietary Information would be highly likely to cause substantial harm to EPRl's competitive position and the ability of EPRI to license the Proprietary Information both domestically and internationally. The Proprietary Information can only be acquired and/or duplicated by others using an equivalent investment of time and effort.

I have read the foregoing and the matters stated herein are true and correct to the best of my knowledge, information and belief. I make this affidavit under penalty of perjury under the laws of the United States of America and under the laws of the State of North Carolina.

Executed at 1300 W WT Harris Blvd being the premises and place of business of Electric Power Research Institute, Inc.

Date: G- llo - lo I~.

~j ueL.

Neil Wilmshurst (State of North Carolina)

(County of Mecklenburg)

Subscrib.ed and sworn to (or affirmed) before me on this -1.1t-7fa.Y of ___i.~zn.<~

_ _,.1lt..-~----4.Hf.J,<<L~-. . .,J."-1Ul-"Atl-4--- - - - - -' proved to me on the basls of the person(s) who appeaie'Cfbefore me.

Signature fj__~a/t.._ J/. ~UJAt (Seal)

My Commission Expires -.day of ~ '20~~

License Amendment Request to Revise Operating License and Technical Specifications for Measurement Uncertainty Recapture (MUR) Power Uprate Enclosure 9 GEH Report NEDO-33853, Safety Analysis Report for Columbia Generating Station Thermal Power Optimization, Revision 0 (Non-Proprietary Version)

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NED0-33853 REVISION 0 NON- PROPRIETARY INFORMATION - CLASS I (PUBLIC)

ACRONYMS AND ABBREVIATIONS Term Definition ABA Amplitude Based Algorithm AC Alternating Cunent ACEC Acute Critical Eftluent Concentration ADS Automatic Depressurization System AL Analytical Limit ALARA As Low As Reasonably Achievable ANSI American National Standards Institute AOO Anticipated Operational Occtmence AP Annulus Pressurization APRM Average Power Range Monitor ARI Alternate Rod Inse1tion ART Adjusted Reference Temperature ASME American Society Of Mechanical Engineers ATWS Anticipated Transient Without Scram AV Allowable Value B&PV Boiler and Pressure Vessel BHP Brake Horsepower BOP Balance-of-Plant BSP Backup Stability Protection BWR Boiling Water Reactor BWRVIP Boiling Water Reactor Vessel and Internals Project CAC Containment Atmosphere Control CFD Condensate Filter Demineralizer CFR Code of Federal Regulations CGS Columbia Generating Station CIV Containment Isolation Valve CLTP CmTent Licensed Thermal Power CRD Control Rod Drive CRGT Control Rod Guide Tube csc Containment Spray Cooling lX

NED0-33853 REVISION 0 NON- PROPRIETARY INFORMATION - CLASS I (PUBLIC)

Term Definition css Core Suppo1i Stmcture CUF Cumulative Usage Factor DBA Design Basis Accident DC Direct CuITent DEH Digital Electro-Hydraulic DIVOM Delta Critical Power Ratio Over fuitial Minimum Critical Power Ratio Versus Oscillation Magnitude ECCS Emergency Core Cooling System EFPY Effective Full Power Years ELTRI NEDC-32424P-A, Generic Guidelines for General Electric Boiling Water Reactor Extended Power Uprate ELTR2 NEDC-32523P-A, Generic Evaluations of General Electric Boiling Water Reactor Extended Power Uprate EOC End-of-Cycle EOP Emergency Operating Procedure EPU Extended Power Uprate EQ Environmental Qualification FAC Flow Accelerated CoITosion FWTR F eedwater Temperature Reduction FIV Flow-fuduced Vibration FPC Fuel Pool Cooling FSAR Final Safety Analysis Repo1t FW Feedwater GDC General Design Criteria GE General Electric Company GEH GE-Hitachi Nuclear Energy GL Generic Letter GRA Growth Rate Algorithm GS3 GEH Simplified Stability Solution HCOM Hot Channel Oscillation Magnitude HELB High Energy Line Break HEPA High Efficiency Particulate Air x

NED0-33853 REVISION 0 NON- PROPRIETARY INFORMATION - CLASS I (PUBLIC)

Term Definition HFCL High Flow Control Line HP High Pressure HPCI High Pressure Coolant fujection HPCS High Pressure Core Spray HVAC Heating, Ventilation and Air Conditioning IASCC hndiation Assisted Stress Conosion Cracking ICF Increased Core Flow ICH&GT In-Core Housing and Guide Tube IPE Individual Plant Examination IRM Intennediate Range Monitor JR Jet Reaction Klc Stress Intensity for Crack fuitiation ksi Kips Per Square Inch kV Kilovolt LCO Limiting Condition for Operation LHGR Linear Heat Generation Rate LOCA Loss-of-Coolant-Accident LPCI Low Pressure Coolant Injection LPCS Low Pressure Core Spray LPRM Local Power Range Monitor LPSP Low Power Setpoint LTS Long-Term Solution MAPLHGR Maximum Average Planar Linear Heat Generation Rate MCPR Minimum Critical Power Ratio MELC Moderate Energy Line Crack MELLLA Maximum Extended Load Line Limit Analysis MeV Million Electron Volts Mlb Millions Of Pounds MOV Motor-Operated Valve MS Main Steam MSF Modified Shape Function Xl

NED0-33853 REVISION 0 NON- PROPRIETARY INFORMATION - CLASS I (PUBLIC)

Term Definition MSIV Main Steam Isolation Valve MSIVC Main Steam Isolation Valve Closure MSL Main Steam Line MSLB Main Steam Line Break MS LBA Main Steam Line Break Accident MVA Megavolt Amps MWe Megawatt(s)-Electric MWt Mega watt(s)-Thennal NCL Natural Circulation Line NFI New Fuel Introduction NFWT Nominal F eedwater Temperature NP DES National Pollutant Discharge Elimination System NPSH Net Positive Suction Head NRC Nuclear Regulato1y Commission NSSS Nuclear Steam Supply System NTSP Nominal Trip Setpoint ODAF Oil-Directed, Air-Forced OFS Orificed Fuel Support OLMCPR Operating Limit Minimum Critical Power Ratio OLTP Original Licensed The1mal Power oos Out-of-Service OPRM Oscillation Power Range Monitor PIF Power/Flow P-T Pressure-Temperature PBDA Period-Based Detection Algorithm PCB Polychlorinated Biphenyl PCS Pressure Control System PCT Peak Clad Temperature PDC Plant Design Change PRA Probabilistic Risk Assessment PRFO Pressure Regulator Failure Open-Maximum Steam Demand XU

NED0-33853 REVISION 0 NON- PROPRIETARY INFORMATION - CLASS I (PUBLIC)

Term Definition psi Pounds Per Square hlch ps1a Pounds Per Square hlch - Absolute psid Pounds Per Square hlch - Differential ps1g Pounds Per Square hlch - Gauge RBM Rod Block Monitor RCC Reactor Building Closed Cooling Water RCIC Reactor Core Isolation Cooling RCPB Reactor Coolant Pressure Boundaiy RFW Reactor Feedwater RFWP Reactor Feedwater Pump RFWT Reduced F eedwater Temperature RG Regulato1y Guide RHR Residual Heat Removal RIPD Reactor hltemal Pressure Difference RIS Re gulato1y Issue Summary RLB Recirculation Line Break RPT Recirculation Pump Trip RPV Reactor Pressure Vessel RRC Reactor Recirculation RRS Reactor Recirculation System RTNDT Reference Temperature of Nil-Ductility Transition RTP Rated Thermal Power RWCU Reactor Water Cleanup RWE Rod Withdrawal Eirnr RWM Rod W 01th Minimizer SAR Safety Analysis Repo1i SBO Station Blackout SC Safety Communication SCCS Successive Confirmation Count Setpoint SDC Shutdown Cooling SER Safety Evaluation Rep01t Xlll

NED0-33853 REVISION 0 NON- PROPRIETARY INFORMATION - CLASS I (PUBLIC)

Term Definition SFP Spent Fuel Pool SGTS Standby Gas Treatment System SL Safety Limit SLCS Standby Liquid Control System SLMCPR Safety Limit Minimum Critical Power Ratio SLO Single-Loop Operating SPC Suppression Pool Cooling SR Surveillance Requirement SRP Standard Review Plan SRSS Square Root Sum of Squares SRV Safety Relief Valve SRVDL Safety Relief Valve Discharge Line STP Simulated Thennal Power SW Service Water TBV Turbine Bypass Valve TCV Turbine Control Valve TFSP Turbine First-Stage Pressure T/G Turbine-Generator TIP Traversing ill-Core Probe TLO Two (Recirculation) Loop Operation TLTP TPO Licensed Thennal Power TLTR NEDC-32938P-A, The1mal Power Optimization Licensing Topical Repoti TPO The1mal Power Optimization TS Technical Specification(s)

TSAR Thennal Power Optimization Safety Analysis Repo1t TSV Turbine Stop Valve TSW Plant Service Water UHS Ultimate Heat Sink USE Upper Shelf Energy vwo Valves Wide Open Wd Recirculation Drive Flow XIV

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NED0-33853 REVISION 0 NON- PROPRIETARY INFORMATION - CLASS I (PUBLIC) 1.3.2 Reactor Performance Improvement Features The following performance improvement and equipment out-of-service (OOS) featm es cmTently licensed at CGS are acceptable at the TPO RTP level:

Performance Improvement Feature Single Loop Operation Increased Core Flow (ICF) (106.0% of rated)

MELLLA (82.7% of Rated Core Flow at TPO Licensed The1m al Power (TLTP))

Feedwater Temperatm e Reduction (FWTR), 355°F F eedwater Heater(s) OOS, 3 55°F SRV OOS (12 Valves in Service) I Automatic Depressm ization System (ADS) 2 Valves OOS Tmbine Bypass Valve (TBV) OOS Recirculation Pump Trip (RPT) OOS 1.4 BASI S FOR TPO UPRATE The safety analyses in this repo1t are based on a total the1m al power measm ement unce1tainty of 0.3%. This will bound the actual power level requested. The detailed basis value is provided in CGS plant design change (PDC) EC-14942, which addresses the improved FW flow measurement accuracy using the Caldon Leading Edge Flow Meter Check-Plus system.

1.5 S UMMARY AND CONCL USIONS This evaluation has investigated a TPO uprate to 101.66% of CLTP. The strategy for achieving this higher power is to increase core flow along the established MELLLA rod lines. The plant licensing challenges have been reviewed (Table 1-3) to demonstrate how the TPO uprate can be accommodated without a significant increase in the probability or consequences of an accident previously evaluated, without creating the possibility of a new or different kind of accident from any accident previously evaluated, and without exceeding any existing regulato1y limits or design allowable limits applicable to the plant which might cause a reduction in a margin of safety. The TPO uprate described herein involves no significant hazards consideration.

1-5

NED0-33853 REVISION 0 NON- PROPRIETARY INFORMATION - CLASS I (PUBLIC)

Table 1-1 Computer Codes For TPO Analyses Computer* Ve1'Sion or NRC Task Comments Code Revision Approved Reactor Heat ISCOR 09 Y(l) NEDE-2401 lP Rev. 0 SER Balance ODYSY 05 y NEDE-33213P-A Tuennal-Hydraulic ISCOR 09 Y(l) NEDE-2401 lP Rev. 0 SER Stability PANAC 11 Y(2) NEDE-30130-A TRACG 04 y NEDE-33147P-A Rev. 4 Reactor h1temal Pressure ISCOR 09 Y(l) NEDE-2401 lP Rev. 0 SER Differences Piping Components Flow SAP4G07 07 N(3) NED0-10909 fuduced Vibration (FIV)

The application of these codes to the CGS TPO analyses complies with the limitations, restrictions, and conditions specified in the approving NRC Safety Evaluation Repo1t (SER) where applicable for each code.

Notes for Table 1-1 :

(1) Tue ISCOR code is not approved by name. However, the SER suppo1ting approval of NEDE-24011P Revision 0 by the May 12, 1978 letter from D. G. Eisenhut (NRC) to R. Gridley (GE) finds the models and methods acceptable, and mentions the use of a digital computer code. The referenced digital computer code is ISCOR. The use of ISCOR to provide core thermal-hydraulic information in reactor internal pressure differences, transient, ATWS, stability, reactor core and fuel pe1fonnance, and LOCA applications is consistent with the approved models and methods.

(2) Tue use of PANAC Version 11 was initiated following approv<tl of Amendment 26 of GESTAR II by letter from S. A Richards (NRC) to G.A. Watford (GE),

Subject:

"Amendment 26 to GE Licensing Topical Repo1t NEDE-2401 lP-A, GESTAR II Implementing Improved GE Steady-State Methods," (TAC NO.

MA6481), November 10, 1999.

(3) Not a safety analysis code that requires NRC approval. Tue code application is reviewed and approved by GEH for "Level-2" application and is part of GEH's standar*d design process. Also, the application of this code has been used in previous power uprate submittals.

1-6

NED0-33853 REVISION 0 NON- PROPRIETARY INFORMATION - CLASS I (PUBLIC)

Table 1-2 Thermal-Hydraulic Parameter s at TPO Uprate Conditions TPORTP Para meter CLTP (101.66% of CLTP)

Themial Power (MWt) 3,486 3,544 (Percent of CmTent Licensed Power) 100.0 101.66 Steam Flow (Mlb/hr) 15 .016 15.284 (Percent of CmTent Rated) 100.0 101.8 FW Flow (Mlb/hr) 14.985 15.253 (Percent of CmTent Rated) 100.0 101.8 Dome Pressme (psia) 1,035 1,035 Dome Temperature (°F) 548.8 548.8 FW Temperatme (°F) 421.2 422. 1 Full Power Core Flow Range (Mlb/hr) 87.6 to 115.0 89.7 to 115.0 (Percent of CmTent Rated) (80.7 to 106.0) (82.7 to 106.0) 1-7

NED0-33853 REVISION 0 NON- PROPRIETARY INFORMATION - CLASS I (PUBLIC)

Table 1-3 Summary of Effect of TPO Uprate on Licensing Criteria Effect of 1.66%

Key Licensing Ciitelia E xplanation of Effect The1*mal Power* Inc1*ease LOCA challenges to foe! No increase in peak clad temperature Previous analysis accounted for~ 102% of (10 CFR 50, Appendix K) (PC1), no ch ange of maximlllll LHGR licensed power, botmding 1PO operation. No required. vessel pressure increase.

Change of operating limit < 0.01 increase. Minor increase (< 0.01) due to slightly h igher MCPR (OLMCPR) power density and increased minimum critical power ratio (MCPR) safety limit (SL) (slightly flatter radial power distribution).

Challenges to reactor pressure No increase in peak pressure. No inci*ease because previous analysis accounted vessel (RPV) overpressure for~ 102% overpower, bounding 1PO operation.

Prima1y containment pressure No increase in peak containment Previous analysis accounted for~ 102%

during a LOCA pressure. overpower, botmding 1PO operation . No vessel pressure increase. No increase in energy to the pool.

Suppression pool temperature No inci*ease in peak suppression pool Previous analysis accounted for~ 102%

during a LOCA temperature. overpower, botmding 1PO operation . No vessel pressure increase. No increase in energy to the suppression pool.

Offsite radiation release, DBAs No increase (remains within Previous analysis bounds 1PO operation. No RPV 10 CFR 50.67) . pressure increase.

Onsite radiation dose, nonual Approximately 1.66% increase . Must Slightly higher inventory of radionuclides in operation remain within 10 CFR 20 limits. steam/PW flow paths.

Heat disch arge to environment Less than 1°F temperature increase. Small (1 .66%) power increase.

Equipment qualification Remains within cmrent pressure, No change in harsh environment tenus (1PO radiation, and temperature envelopes. operating conditions bounded by previous analyses); minima l change in no1mal operating conditions.

Fracture toughness, < 2°F increase in reference temperature Small increase in neutron fluence.

10 CFR 50, Appendix G of the nil-ductility trans ition (RTNDr).

Stability No direct effect of1PO uprate because No inci*ease in maximum rod line boundary.

applicable stability regions and lines Characteristics of each reload core continue to be are extended beyond the absolute evah1ated as required for each stability option.

values associated with the ct11Tent boundaries to prese1v e MWt-core flow boundaries as applicable for each stability option .

ATWS peak vessel pressure Slight increase (30 psig), must stay Slightly increased power relative to SRV capacity.

witliin existing American Society of Mechanical Engineers (ASME) Code "Emergency" catego1y stress limit.

Vessel and NSSS equipment No change. Comply witli existing ASME Code stress limits design pressure for all categories.

1-8

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NED0-33853 REVISION 0 NON- PROPRIETARY INFORMATION - CLASS I (PUBLIC)

Legend

= Flow, lbm/hr 1035 p '

H = Enthalpy, Btu/lbm '~

0 .1 M F = Temperatw-e, F M = Moisture, %

P = Pressure, psia t t I

I

,I Main Steam Flow 15 .284E+06 # *

' 1191.0 H

0.30 M

Carryunder = 0 .25% 999 p

3544 Main Feed Flow MWt Wd= lOO % 15.434E+06 # 15.253E+06 #

529.5 H 399.7 H 399.9 H 534.4 °P Total 422.0 Of 422.1 °P Core Flow 108.5E+06

~h= 1.2 H #

l.813E+05 #

JiLL H 386.l H 409.4 °P Cleanup Demineralizer System 3.140E+04 # l.813E+05 #

50.9 H 528.3 H 80.0 °P 533.4 °P

Conditions at upstream side ofTSV Core Thermal Power 3544.0 Pump Heating 11.5 Cleanup Losses -7.6 Other System Losses - 1.l Tw-bine Cycle Use 3546.8 MWt Figure 1-2 Reactor Heat Balance-TPO Power (101.66% of CLTP), 100% Core Flow 1- 10

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NED0-33853 REVISION 0 NON- PROPRIETARY INFORMATION - CLASS I (PUBLIC)

Table 2-1 OPRM Amplitude Setpoint Versus OLMCPR Demonstration CGSTPO OPRM Amplitude Setpoint OLMCPR(2RPT) OLMCPR(SS) 1.05 1.27 1.19 1.06 1.29 1.21 1.07 1.31 1.23 1.08 1.33 1.25 1.09 1.35 1.27 1.10 1.37 1.29 1.11 1.40 1.31 1.12 1.42 1.33 1.13 1.44 1.35 1.14 1.47 1.37 1.15 1.49 1.40 Off-Rated OLMCPR Acceptance Criteria Rated Power OLMCPR at45% Flow 2-5

NED0-33853 REVISION 0 NON- PROPRIETARY INFORMATION - CLASS I (PUBLIC)

Table 2-2 BSP Region Intercepts for Nominal Feedwater Temperature Demonstration Region Boundary Intercept % TPOPower % Core Flow Region (Region I) Boundary Intercept on HFCL Scram Al -Base 63.6 40.0 Scram Region (Region I) Boundary Intercept on NCL Bl 37.1 23.8 Controlled Entry Region (Region II) Boundary Intercept on HFCL A2-Base 72.6 50.0 Controlled Entry Region (Region II) Boundary Intercept on NCL B2 Base 28.1 23.7 2-6

NED0-33853 REVISION 0 NON- PROPRIETARY INFORMATION - CLASS I (PUBLIC)

Table 2-3 BSP Region Intercepts for Reduced Feedwater Temperature Demonstration Region Boundary Intercept % TPOPower % Core Flow Scram Region (Region I) Boundary Intercept on HFCL Al 64.8 41.3 Scram Region (Region I) Boundary Intercept on NCL Bl 32.0 23.8 Controlled Entry Region (Region II) Boundary Intercept on HFCL A2-Base 72.6 50.0 Controlled Entry Region (Region II) Boundary Intercept on NCL B2-Base 28.1 23.7 2-7

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3.2.2 Reactor Vessel Structural Evaluation

((

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The TLTR provides a generic disposition for components that are not significantly affected. The following table provides the justification for confirming the TLTR generic disposition:

TLTR Generic Justification I Topic Parameter(s) CLTP vs. TPO Comparison or Requirement(s)

((

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NED0-33853 REVISION 0 NON- PROPRIETARY INFORMATION - CLASS I (PUBLIC) 3.4 FLOW-INDUCED V IBRATION The process for the reactor vessel internals vibration assessment is described in TLTR Section 5.5.1.3. An evaluation detennined the effects of flow-induced vibration (FIV) on the reactor internals at 106% rated core flow and TPO RTP of 101.66% of CLTP. The vibration levels for the TPO conditions were estimated from measmed vibration data dming staiiup tests on CGS and the NRC designated prototype plant (Tokai-2), as well as other plants. The expected vibration levels were compared with established vibration acceptance limits. The following components were evaluated for the TPO uprate:

Component(s) Process Parameter(s) TPO Evaluation Slight increase (< 4%) in FW flow at TPO RTP is FIV. Extrapolation of FW Spai*ger approximately 2% greater measmed data shows stresses than CLTP.

are within limits.

The increase in jet pump flow at TPO is negligible Slight increase (< 2%) in based on no change in core FIV. Extrapolation of Jet Pumps flow and a minor increase in measmed data shows stresses core differential pressme are within limits.

(< 0.1 psi).

No resonance at vane passing frequency range due to TPO.

Jet Pump Sensing Resonance at vane passing Clamps have been installed at Lines frequency all JPSLs to prevent resonance with VPF.

Flow at TPO RTP is Slight increase (< 4%) in Shroud approximately 2% greater FIV. The maximum stresses than CLTP. are well within limits.

Slight increase (< 4%) in Steam flow at TPO RTP is Shroud Head and FIV. Extrapolation of approximately 2% greater Separator measmed data shows stresses than CLTP.

are within limits.

CRGT and In-Core Core flow at TPO is No change.

Guide Tubes unchanged from CLTP.

The calculations for the TPO uprate conditions indicate that vibrations of all safety-related reactor internal components are within the GEH acceptance criteria. The analysis is conservative for the following reasons:

The GEH criteria of 10,000 psi peak stress intensity are more conservative than the ASME allowable peak stress intensity of 13,600 psi for service cycles~ 1011 .

Conservatively, the peak responses of the applicable modes ai*e absolute summed.

3-7

NED0-33853 REVISION 0 NON- PROPRIETARY INFORMATION - CLASS I (PUBLIC)

The maximum vibration stress amplitude of each mode is used in the absolute sum process, whereas in reality the maximum vibration amplitudes are unlikely to occur at the same time.

Therefore, it is concluded that the flow-induced vibrations for all evaluated components remain within acceptable limits.

The safety-related main steam (MS) piping has minor increased flow rates and flow velocities resulting from the IPO uprate. CGS has no safety-related the1mowells and sample probes installed in the FW system The piping components were evaluated in accordance with ASME Code N-1300 (Reference 12)

FIV analysis guidelines. The resonance separation mle in ASME Appendix N SubparagraphN-1324.l(d) of Reference 12 was used to dete1mine ifadequate separation exists between the vo1iex shedding frequencies and the natural frequencies of the piping components.

The MS piping experiences increased vibration levels, approximately propo1tional to the increase in the square of the flow velocities and also in propo1tion to any increase in fluid density. The MS piping vibration is expected to increase only by about 4% from 3.753 million pounds (Mlb)/hr per line at CLIP to 3.822 Mlb/hr per line at IPO. A MS piping FIV test program, after the implementation of the power uprate to CLTP, showed that vibration levels were within acceptance criteria and operating experience shows that there are no existing vibration problems in MS lines at CLIP operating conditions. Therefore, the MS lines vibration will remain within acceptable limits during TPO. Analytical evaluation has shown that the safety-related the1mowells and sample probes in the MS and recirculation piping systems are stmcturally adequate for the IPO operating conditions.

3.5 PIPING E VALUATION 3.5.1 Reactor Coolant Pressure Boundary Piping The methods used for the piping and pipe suppo1i evaluations are described in TLTR Appendix K. These approaches are identical to those used in the evaluation of previous BWR power uprates of up to 20% power. The effect of the TPO uprate with no nominal vessel dome pressure increase is negligible for the reactor coolant pressure bounda1y (RCPB) po1iion of all piping except for po1iions of the FW lines, MS lines, and piping connected to the FW and MS lines. The following table summarizes the evaluation of the piping inside containment.

Component(s) I Concern Process Parameter(s) TPO Evaluation Recirculation System Nominal dome pressure at IPO RIP is Negligible change identical to CLIP. in pipe stress Pipe Stresses Recirculation flow at TPO RTP is Pipe Suppo1is identical to CLTP. Negligible effect Small increase in core pressure drop of on pipe suppo1is

< 1 psi.

Recirculation fluid temperature increases

- 1°F.

3-8

NED0-33853 REVISION 0 NON- PROPRIETARY INFORMATION - CLASS I (PUBLIC)

Component(s) I Concern Process Parameter(s) TPO Evaluation MS and Attached Piping Nominal dome pressure at TPO RTP is Plant specific (Inside Containment) identical to CLTP. evaluation (e.g., SRV discharge line Steam flow at TPO RTP is ~ 2% greater performed (SRVDL) piping up to first than CLTP.

anchor, reactor core isolation Minor decrease in main steam line Minor change in cooling (RCIC) MS drain lines, pipe stress (MSL) pressure < 2 psi.

RPV head vent line piping located inside containment) Minor effect on pipe suppo1ts Pipe Stresses Pipe Suppol1s Minor increase in Flow-Accelerated the potential for Erosion/C01rnsion (F AC) FAC (FAC concerns are covered by existing piping monitoring program)

FW and Attached Piping Nominal dome pressure at TPO RTP is Plant specific (Inside Containment) identical to CLTP. evaluation FW flow at TPO RTP is ~2% greater performed Pipe Stresses than CLTP.

Pipe Suppo1ts Minor change in FW line pressure. Negligible change in pipe stress Fluid temperature remains the same.

Negligible effect on pipe suppol1s FAC Minor increase in the potential for FAC(FAC concerns are covered by existing piping monitoring program) 3-9

NED0-33853 REVISION 0 NON- PROPRIETARY INFORMATION - CLASS I (PUBLIC)

Component(s) I Concern Process Parameter(s) TPO Evaluation RPV Bottom Head Drain Line, Nominal dome pressure at TPO RTP is Negligible change RCIC Piping, High Pressure identical to CLTP. in pipe stress Core Spray (HPCS) Piping, Small increase in core pressure drop of LPCI Piping, LPCS Piping, < 1 psi. Negligible effect SLCS Piping, and Reactor on pipe suppo1ts Recirculation fluid temperature increases Water Cleanup (RWCU)

- 1°F.

Piping Pipe Stresses Pipe Suppo1ts FAC Minor increase in the potential for FAC{FAC concerns are covered by existing piping monitoring program)

For the MS and FW lines, suppo1ts, and connected lines, the methodologies as described in TLTR Section 5.5.2 and Appendix K were used to detennine the percent increases in applicable ASME Code stresses, displacements, CUFs, and pipe interface component loads (including suppo1ts) as a function of percentage increase in pressure (where applicable), temperature, and flow due to TPO conditions. The percentage increases were applied to the highest calculated stresses, displacements, and the CUF at applicable piping system node points to conservatively detennine the maximum TPO calculated stresses, displacements and usage factors. This approach is conse1vative because the TPO does not affect weight and all building filtered loads (i.e., seismic loads are not affected by the TPO). The factors were also applied to nozzle load, suppo1t loads, penetration loads, valves, pumps, heat exchangers and anchors so that these components could be evaluated for acceptability, where required. No new computer codes were used or new assumptions introduced for this evaluation.

MS and Attached Piping System Evaluation The MS piping system (inside containment) was evaluated for compliance with the ASME code stress criteria, and for the effects of thennal displacements on the piping snubbers, hangers, and stiuts. Piping interfaces with RPV nozzles, penetrations, flanges and valves were also evaluated.

Pipe Stresses The evaluation shows that the increase in flow associated with the TPO uprate does not result in load limits being exceeded for the MS piping system or for the RPV nozzles. The original design analyses have sufficient design margin between calculated stresses and ASME Code 3- 10

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NED0-33853 REVISION 0 NON- PROPRIETARY INFORMATION - CLASS I (PUBLIC) 3.8 MAIN STEAM I SOLATION VALVES The generic evaluation provided in TLTR Appendix J.2 .3.7 is applicable to CGS. The requirements for the main steam isolation valves (MSIVs) remain unchanged for TPO uprate conditions. All safety and operational aspects of the MSIVs are within previous evaluations.

3.9 REACTOR CORE I SOLATION COOLING The reactor core isolation cooling (RCIC) system provides invento1y makeup to the reactor vessel when the vessel is isolated from the nonnal high pressure makeup systems. The generic evaluation provided in TLTR Section 5.6.7 is applicable to CGS. The TPO uprate does not affect the RCIC system operation, initiation, or capability requirements.

3.10 REs m u ALllEATREMovALSYSTEM The residual heat removal (RHR) system is designed to restore and maintain the coolant invento1y in the reactor vessel and to remove sensible and decay heat from the primaiy system and containment following reactor shutdown for both n01mal and post-accident conditions. The RHR system is designed to function in several operating modes. The generic evaluation provided in TLTR Section 5.6.4 and Appendices J.2.3.1 and J.2.3.13 are applicable to CGS.

The following table summarizes the effect of the TPO on the design basis of the RHR system.

Operating Mode Key Function TPO Evaluation LPCIMode Core cooling See Section 4.2.4 Suppression Pool Cooling Normal SPC function is to Containment analyses (SPC) and Containment Spray maintain pool temperatme below have been perfo1med at Cooling (CSC) Modes the limit. 102% of CLTP.

For abno1mal events or accidents, the SPC mode maintains the long-te1m pool temperature below the design limit.

The CSC mode sprays water into the containment to reduce post-accident containment pressme and temperatme.

Shutdown Cooling (SOC) Removes sensible and decay heat The slightly higher decay Mode from the reactor primaiy system heat has a negligible dming a n01mal reactor effect on the SOC mode, shutdown. which has no safety function.

Steam Condensing Mode Decay heat removal CGS does not have a steam condensing mode of RHR.

3- 15

NED0-33853 REVISION 0 NON- PROPRIETARY INFORMATION - CLASS I (PUBLIC)

Operating Mode Key Function TPO Evaluation Fuel Pool Cooling (FPC) Supplemental FPC in the event See Section 6.3.1 Assist that the fuel pool heat load exceeds the heat removal capability of the FPC system.

The ability of the RHR system to perform required safety functions is demonstrated with analyses based on 102% of CLTP. Therefore, all safety aspects of the RHR system are within previous evaluations. The requirements for the RHR system remain unchanged for TPO uprate conditions.

3.11 REAC TOR WATER CLEANUP S YSTEM The generic evaluation of the RWCU system provided in TLTR Sections 5.6.6 and J.2.3.4 is applicable to CGS. The perfo1mance requirements of the RWCU system are negligibly affected by TPO uprate. There is no significant effect on operating temperature and pressure conditions in the high pressure po1tion of the system. RWCU flow is not changed for TPO conditions.

Steady power level changes for much larger power uprates have shown no effect on reactor water chemistiy and the perfo1mance of the RWCU system. Power transients that result in cmd bursts causing high inte1mediate loading on the system capacity are the primary source of challenge to the system, so safety and operational aspects of water chemistiy performance are not affected by the TPO.

3- 16

NED0-33853 REVISION 0 NON- PROPRIETARY INFORMATION - CLASS I (PUBLIC)

Table 3-1 CGS Upper Shelf Energy 60-Year License (51.56 EFPY)

S1.S6EFPY Internal Traverse U SE Traverse Material Heat or Heat/ Lot %Cu S1.S6 EFPY 1'T Fluence (n/cm'). Decrease (ft-lb) USE (3)

USE (1)(2)

(ft/ lb).

Plates Lower Intermediate Shell 85301-1 98 0.13 7.93E+l 7 12.5 (4) 86 Welds:

Vertical:

Lower Shell 3P4966 I 12 14-3482 ( S) (5) 98 (( )) (6) 2.64E+l 7 7.5 91 3P4966 I 12 14-3482 (I) (5) 98 (( )) (6) 2.64E+l 7 7.5 91 Lower Intermediate Shell 3P4966 I 12 14-3481 ( S) (5) 98 (( )) (6) 7.93E+l 7 9.5 89 3P4966 I 12 14-3481 (I) (5) 98 (( )) (6) 7.93E+l 7 9.5 89 Girth:

5P6756 I 0342-3447 ( S) (5) 91 (( )) (6) 3.22E+l 7 10 82 5P6756 I 0342-3447 (I) (5) 97 (( )) (6) 3.22E+l 7 10 87 5P4955 I 0342-3443 ( S) (5) 90 (( )) (6) 3.22E+l 7 8 83 5P49551 0342-3443 (I) (5) 95 (( )) (6) 3.22E+l 7 8 87 Nozzles:

N6 Nozzle Q2Q55W I 190S 70 0.11 4.36E+l 7 10 63 N6 Weld 5P6214B I 0331 (S) (5) 70 (( )) (6) 4.36E+l 7 8 (7) 64 5P62 14B I 0331 (I) (5) 70 (( ))(6) 4.36E+l 7 8 (7) 64 Nl2 Nozzle 2 19972 1 1 62 (( )) (8) 2.30E+l 7 16 52 718259 I 65363 62 0.25 (8) 2.30E+l 7 14.5 53 Nl2 Weld Inco 82 1 182 NA NA NA NA NA 3- 17

NED0-33853 REVISION 0 NON- PROPRIETARY INFORMATION - CLASS I (PUBLIC)

Table 3-1 CGS Upper Shelf Energy 60-Year License (51.56 EFPY) continued 51.56 EFPY Internal Trave rse USE 2 Traverse Material Heat or Heat/Lot %Cu 51.56 EFPY l' T Fluence (n/ cm ). Decrease (ft-lb) USE (3)

USE (1)(2) lft/lbl.

ISP Representative Materials (EPRI Proprietary Information (9))

Plate B0673-l (9)(10) NA NA NA NA NA Weld River Berxl 183 & SSP (F. H, C) 5P6756 (9)(11) 104.4 (9) 0.06 (9) 3.22E+ l7 9 95 PYl 3 & 177 5P6214B (9)(11) 90.9 (9) 0.027 (9) 4.36E+ l7 8.5 83 SSP (A. B, D, E, G, I) 5P6214B (9)(11) 9 1.5 (9) 0.01 (9) 4.36E+ l7 7.5 85 Notes:

1. USE Decrease obtained from Figure 2 in RG 1.99, Revision 2.

2. Rounded up to the nearest 0.5 value.

3. 51 .56 EFPY Transverse USE = Initial Transverse USE* [1 - (% Decrease USE / 100)].

4. Previous evaluation appears to have used "weld" line vs. "base" line in RG 1.99 Figure 2 calculation, thus resulting in a different% Decrease USE.

5. (S) = Single, (I) = Tandem

6. Best Estimate Chemistty used; Heat # and %Cu obtained from BWRVIP-135 R3 (Reference 14) (EPRI Proprietary h1formation).

7. Previous evaluation used conservative %Cu value of 0.05 in RG 1.99 Figure 2 (% decrease USE) calculation. Thus the resulting% decrease USE was higher.

8. Previous evaluation considered different %Cu value.

9. ISP Representative Material fuformation; Heat #, Initial USE and %Cu obtained from BWRVIP-135 R3 (EPRI Proprietaiy fuformation) .

10 . Representative heat# is not a match to plant material.

11. Representative heat# is a match to plant mat erial.

3- 18

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Equivalent Margin Analysis Plant Applicability Verification Form for Columbia Including Power Uprate Conditions 60-Year License (Cumulative Energy Provided in Fluence Report)

BWR/3-6 Plate Surveillance Plate USE (Heat B5301-1)

%Cu = 0.11 Unirradiated USE = 98.0 ft-lb 1st Capsule Measured USE = 99.6 ft-lb 1st Capsule Fluence = 1.55E+17 n/cm2 1st Capsule Measured % Decrease = -1.6 (Charpy Curves) 1st Capsule RG 1.99 Predicted % Decrease = 8.0 (RG 1.99, Rev. 2, Figure 2)

ISP Surveillance Plate USE (Heat B0673-1) (For information purposes only not as a matching beltline heat)

%Cu = 0.15 Unirradiated USE = 158.1 ft-lb DA 288° Capsule Measured USE = 158.8 ft-lb DA 288° Capsule Fluence = 5.09E+17 n/cm2 DA 36° Capsule Measured USE = 137 ft-lb DA 36° Capsule Fluence = 1.17E+18 n/cm2 SSP Capsule F Measured USE = 133 ft-lb SSP Capsule F Fluence = 1.87E+18 n/cm2 DA 108° Capsule Measured USE = 131.3 ft-lb DA 108° Capsule Fluence = 2.63E+18 n/cm2 DA 288° Capsule Measured % Decrease = -0.4 (Charpy Curves)

DA 288° Capsule RG 1.99 Predicted % Decrease = 12.0 (RG 1.99, Rev. 2, Figure 2)

DA 36° Capsule Measured % Decrease = 13.3 (Charpy Curves)

DA 36° Capsule RG 1.99 Predicted % Decrease = 14.5 (RG 1.99, Rev. 2, Figure 2)

SSP Capsule F Measured % Decrease = 15.9 (Charpy Curves)

SSP Capsule F RG 1.99 Predicted % Decrease = 16.5 (RG 1.99, Rev. 2, Figure 2)

DA 108° Capsule C Measured % Decrease = 17.0 (Charpy Curves)

DA 108° Capsule C RG 1.99 Predicted % Decrease = 17.5 (RG 1.99, Rev. 2, Figure 2)

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Limiting Beltline Plate USE (Heat C1337-1 and C1337-2):

CLTP (54 EFPY)

%Cu = 0.15 54 EFPY 1/4 T Fluence = 8.14E+17 n/cm2 (Cumulative Energy Provided in Fluence Report)

RG 1.99 Predicted % Decrease = 13.5 (RG 1.99, Rev. 2, Figure 2)

Adjusted % Decrease = NA (RG 1.99, Rev. 2, Position 2. 2)

TPO (51.56 EFPY)

%Cu = 0.15 51.56 with TPO EFPY 1/4 T Fluence = 7.93E+17 n/cm2 (Cumulative Energy Provided in Fluence Report)

RG 1.99 Predicted % Decrease = 13.5 (RG 1.99, Rev. 2, Figure 2)

Adjusted % Decrease = N/A (RG 1.99, Rev. 2, Position 2. 2) 13.5% < (( )) (for 54 EFPY)

Therefore, vessel plates are bounded by equivalent margin analysis.

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BWR/2-6 WELD Surveillance Weld USE (Heat 3P4966):

%Cu = 0.03 Unirradiated USE = 98.0 ft-lb 1st Capsule Measured USE = 108.0 ft-lb 1st Capsule Fluence = 1.55E+17 n/cm2 1st Capsule Measured % Decrease = -10.2 (Charpy Curves) 1st Capsule RG 1.99 Predicted % Decrease = 6.0 (RG 1.99, Rev. 2, Figure 2)

ISP Surveillance Weld USE (Heat 5P6756): Potentially use for adjustment because it is a matching be t ine heat.

%Cu = 0.06 Unirradiated USE = 104.4 ft-lb River Bend 183° Capsule Measured USE = 84.4 ft-lb River Bend 183° Capsule Fluence = 1.16E+18 n/cm2 SSP Capsule F Measured USE = 79.3 ft-lb SSP Capsule F Fluence = 1.94E+18 n/cm2 SSP Capsule H Measured USE = 84.6 ft-lb SSP Capsule H Fluence = 1.58E+18 n/cm2 SSP Capsule C Measured USE = 110.7 ft-lb SSP Capsule C Fluence = 2.93E+17 n/cm2 River Bend 183° Capsule Measured % Decrease = 19.2 (Charpy Curves)

River Bend 183° Capsule RG 1.99 Predicted % Decrease = 12.5 (RG 1.99, Rev. 2, Figure 2)

SSP Capsule F Measured % Decrease = 24.0 (Charpy Curves)

SSP Capsule F RG 1.99 Predicted % Decrease = 14.0 (RG 1.99, Rev. 2, Figure 2)

SSP Capsule H Measured % Decrease = 19.0 (Charpy Curves)

SSP Capsule H RG 1.99 Predicted % Decrease = 13.0 (RG 1.99, Rev. 2, Figure 2)

SSP Capsule C Measured % Decrease = -6.0 (Charpy Curves)

SSP Capsule C RG 1.99 Predicted % Decrease = 9.0 (RG 1.99, Rev. 2, Figure 2)

ISP Surveillance Weld USE (Heat 5P6214B): Potentially use for adjustment because it is a matching be t ine heat.

%Cu = 0.027 Unirradiated USE = 90.9 ft-lb PY1 3° Capsule Measured USE = 85.8 ft-lb PY1 3° Capsule Fluence = 3.18E+17 n/cm2 PY1 177° Capsule Measured USE = 94.9 ft-lb PY1 177° Capsule Fluence = 1.08E+18 n/cm2

%Cu = 0.01 Unirradiated USE = 91.5 ft-lb SSP Capsule D Measured USE = 89 ft-lb SSP Capsule D Fluence = 1.03E+18 n/cm2 SSP Capsule E Measured USE = 88.5 ft-lb SSP Capsule E Fluence = 1.77E+18 n/cm2 SSP Capsule G Measured USE 85.3 ft-lb SSP Capsule G Fluence 1.95E+18 n/cm2 SSP Capsule I Measured USE 87.7 ft-lb SSP Capsule I Fluence 2.75E+18 n/cm2 SSP Capsule A Measured USE 96.5 ft-lb SSP Capsule A Fluence 4.09E+17 n/cm2 SSP Capsule B Measured USE 97.4 ft-lb SSP Capsule B Fluence 5.26E+17 n/cm2

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PY1 3° Capsule Measured % Decrease = 5.6 (Charpy Curves)

PY1 3° Capsule RG 1.99 Predicted % Decrease = 7.4 (RG 1.99, Rev. 2, Figure 2)

PY1 177° Capsule Measured % Decrease = -4.4 (Charpy Curves)

PY1 177° Capsule RG 1.99 Predicted % Decrease = 9.9 (RG 1.99, Rev. 2, Figure 2)

SSP Capsule D Measured % Decrease = 2.7 (Charpy Curves)

SSP Capsule D RG 1.99 Predicted % Decrease = 8.8 (RG 1.99, Rev. 2, Figure 2)

SSP Capsule E Measured % Decrease = 3.3 (Charpy Curves)

SSP Capsule E RG 1.99 Predicted % Decrease = 10.0 (RG 1.99, Rev. 2, Figure 2)

SSP Capsule G Measured % Decrease = 6.8 (Charpy Curves)

SSP Capsule G RG 1.99 Predicted % Decrease = 10.2 (RG 1.99, Rev. 2, Figure 2)

SSP Capsule I Measured % Decrease = 4.2 (Charpy Curves)

SSP Capsule I RG 1.99 Predicted % Decrease = 11.0 (RG 1.99, Rev. 2, Figure 2)

SSP Capsule A Measured % Decrease = -5.5 (Charpy Curves)

SSP° Capsule A RG 1.99 Predicted % Decrease = 7.0 (RG 1.99, Rev. 2, Figure 2)

SSP Capsule B Measured % Decrease = -6.4 (Charpy Curves)

SSP Capsule B RG 1.99 Predicted % Decrease = 7.5 (RG 1.99, Rev. 2, Figure 2)

Limiting Beltline Weld USE (Heat 624039 / D205A27A):

TPO (51.56 EFPY)

%Cu = 0.10 51.56 with TPO EFPY 1/4T Fluence = 7.93E+17 n/cm2 (Cumulative Energy Provided in Fluence Report)

RG 1.99 Predicted % Decrease = 13.5 (RG 1.99, Rev. 2, Figure 2)

Adjusted % Decrease = 20.0 (RG 1.99, Rev. 2, Position 2. 2) 20.0% < (( )) (for 54 EFPY)

Therefore, vessel welds are bounded by equivalent margin analysis

NED0-33853 REVISION 0 NON- PROPRIETARY INFORMATION - CLASS I (PUBLIC)

Table 3-4 CGS Adjusted Reference Temperatures 60-Year License (51.56 EFPY)

Cempoat* Hut or Hut Lot (l) ..c. ,._,, cr C> ..YjoisttdCT (l)

I.oici..

RT*dt

%T Flwact 51.56 EFP1

.utTadt Gt a, lhrp*

"f 51.56 EFPY 51.56 EfPY Sloift ART "f .Jc,* 1 "f *r *r

'PLATES:

Lcnrrr Slwll Mic 21-1-1 Cl272-l 015 060 110 28 2 64E+l7 224 0 112 224 448 n8 Mk 21-1-2 Cl273-1 014 060 100 20 2 64E+l7 204 0 102 204 407 601 Mic 21-1-3 Cl273-2 0 14 060 100 4 2 64E+l7 204 0 102 204 407 447 Mk 21-1-4 Cl272-2 0 15 060 110 0 264E+17 224 0 112 224 448 448 SbtlU1 Mic 22-1-1 B5301-l 0 13 0 50 (4) 88 -20 7 93E+17 328 0 164 32 8 655 455 Mk22-1-2 Cl336-l 0 13 0 50 88 -8 793E+l7 32 8 0 164 328 65 5 S15 Mic 22-1-3 C1337-1 0 15 0 51 105 -20 793E+l7 39 1 0 170 340 731 Sl l Mic 22-1-4 Cl337-2 0 15 051 105 -20 793E+l7 391 0 170 340 731 Sl l NOZZLES:

N6 Nozzle Q2Q55W /790S 0 II 076 76 -20 4 36E+17 208 0 104 208 415 21 5 Weld 5P6214BI0331 (S)(5,6, 7) (( ]) (6) (( )) (6) 27 (( ))(5, 8) -50 (ll, 10) 4 36E+17 15 9 0 79 15 9 317 -18 3 Weld 5P6214BI0331(I)Ci.6, 7J (( ]) (6) (( )) (6) 27 (( ))(5, 8) -24 (ll, 10) 4 36E+17 15 9 0 79 15 9 317 77 Nll Nozzle 2 1997211 (ll) (( 11(12,13) (( )) ( 12, 13 (( )) 40 2 30E+17 25 5 0 128 25 5 511 911 Nozzle 718259165363 (11) 0 25 (16) 0 24 (16) 131 -20 2 30E+l7 245 0 123 24 5 491 291 Weld lnco 821182 (17)

WELDS:

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NED0-33853 REVISION 0 NON- PROPRIETARY INFORMATION - CLASS I (PUBLIC)

Table 3-6 CGS 51.56 EFPY Effects of Irradiation on RPV Circumferential Weld Properties Circumferential Weld Parameters at 51.56 EFPY CGS NRC Limitin2 Plant Overall Specific Analysis Limiting Weld Wire Parameter (Circ Welds) (1)

(3P4955) 64 EFPY 51.56 EFPY (CB&IRPV)

(CB&I RPV)

Cu% 0.10 (( ))

Ni% 0.99 (( ))

CF 134.9 37 End oflicense inside diameter 1.02 0.047 fluence (10 19 n/cm2)

RTNDT(U) (°F) -65 -16

~RTNOT Without margin (°F) (2) 135.6 10.4 Mean RTNDT (°F) 70.6 -5.6 P (FIE) NRC (3) 1.78E-05 (4)

Notes:

1. Chemistry information repo1t ed in BWRVIP-05 .

2. ~RTNDT = CF* f (o2s-0101ogf)

3. P (FIE) stands for "Conditional probability of vessel failure or probability of vessel failure assuming that the event occmr ed".

4. Although a conditional failure probability has not been calculated, the fact that the CGS values at the end of license are less than the 64 EFPY value provided by the NRC leads to the conclusion that the CGS RPV conditional failure probability is bounded by the NRC analysis, consistent with the requirements defined in GL98-05.

3-26

NED0-33853 REVISION 0 NON- PROPRIETARY INFORMATION - CLASS I (PUBLIC)

Table 3-7 CUF and P+Q Stress Range of Limiting C omponents P + Q Stress (ksi) CUF <4.S)

Allowable C urrent Component<I,6) Current TPO TPO Allowable 3 (ASME Code (3,486 (3,486 MWt) (3,545 MWtl ) (3,545 MWt)<3) (ASM E Code Limit)

Limit) MWt) 70.74/ 70.74/

Shroud Suppo1i 69.9 0 . 399 0.399 1.0 65.58(2) 65.58(2,6)

SteamD1yer 38.4 38.4 42.3 0 .064 0.064 1.0 Bracket Notes:

1. There are no changes in operating conditions from CLTP to TPO. Therefore, the CLTP evaluation remains applicable for TPO.

The components presented in this table are consistent with the CLTP SAR to demonstrate that the results remain unchanged from CLTPtoTPO.

2. Thermal bending included/thermal bending removed. P + Q stresses are acceptable per CLTP elastic-plastic analysis where applicable, which is valid for TPO conditions.

3 . ((

))

4. Limiting CUF is presented.

5. Fatigue usage factors are for a 40-year license.

6 . CLTP and TPO were (( )) Therefore, there is no change in values from CLTPtoTPO.

3-27

NED0-33853 REVISION 0 NON- PROPRIETARY INFORMATION - CLASS I (PUBLIC)

Table 3-8 Governing Stress Results for RPV Internals Service Stress CLTP TPO Allowable Item Component Location Unit Value<I) Limit<2)

Level C ategory Value 1 Shroud Suppo11<7) Legs B Pm+Pb psi 25,540 25 ,540 28,100 2 Shroud<?) Top Guide Wedge B Pm+Pb psi 12,730 12,730 21,450 lbs/

3 Core Plate Longest Beam B Buckling 1,016 1,016 1,179 CRGT 4 Top Guide Longest Beam B Pm+Pb psi 28,548 28 ,548 31 ,690 Control Rod Drive Housing CRD Housing at 5a B Pm+Pb psi 15,450 15,450 24,900 (Outside RPV Po1iion) RPV Bottom Head Control Rod Drive Housing CRD Housing at 5b B Pm+Pb psi 11,925 11,925 16,185 (Inside RPV Portion) RPV Stub Tube Control Rod Drive 5c Mechanism CRD Outer Tube B Pm+Pb psi 24,700 24,700 26,100 Control Rod Guide CRGTFlan ge 6a B Pm+Pb psi 8,189 8,189 24,000 Tube (Base)

Control Rod Guide 6b Mid-Span B Pm+Pb psi 9,100 9,100 16,000 Tube Control Rod Guide 6c Body B Buckling NIA 0.40 0.40 0.45 Tube 7 Orificed Fuel Suppo1i OFS Body B Load lbs 14 895(3) 14 895(3) 35 590<3) 3-28

NED0-33853 REVISION 0 NON- PROPRIETARY INFORMATION - CLASS I (PUBLIC)

Table 3-8 Governing Stress Results for RPV Internals (continued)

Service Stress CLTP TPO Allowable Item Component Location Unit Value<I) Limit<2)

Level Category Value Feedwater Sparger FW Pipe-to- Fatigue 0 .88(4) 0.48(S) 8 B NIA 1.0 Sparger Weld Usage 9 Jet Pump Assembly<7) Riser Brace D Pm+Pb psi 54,427 54,427 60,840 lOa Core Spray Line Elbow B Pm+Pb psi 19,890 19,890 23 ,850 lOb Core Spray Sparger Tee Junction B Pm psi 6,560 6,560 21 ,450 Access Hole Cover<7) 11 Cover D Pm+Pb psi 16,031 37,352 49,400 (Top Hat Design)

Shroud Head and Steam 12 Separator Assembly Shroud Head Bolt B Pm psi 7,909 7,909 16,900 In-Core Housing and In-Core Housing at 13 Guide Tube B Pm+Pb psi 25,160 25 ,160 25 ,400 RPV Penetration Core Differential 14 Pressure and Liquid Unknown c Pm+Pb psi 17,015 17,015(6) 36,900 Control Line Low Pressure Coolant 15 Injection Coupling Support Ring c Pm+Pb psi 27,600 27,600 31 ,400 Notes:

(1) Stresses/loads listed are for the limiting loading condition, with the least margin of safety.

(2) Allowable values are consistent with the original design basis.

(3) For OFS, the calculated and allowable loads are in the ve1tical downward direction.

(4) Based on a generic GEH FW sparger analysis report. Includes conservative assumptions.

(5) Based on a 60-year plant life and also based on a generic GEH FW sparger analysis repo1t. However, the conservatism was removed from the generic FW sparger analysis.

( 6) For the core differential and liquid control line, the calculated stress shown is based on an absolute summation of upset loads. Actual stress based on the square root sum of squares (SRSS) methodology will be less.

(7) These components are affected by GEH SCs.

3-29

NED0-33853 REVISION 0 NON- PROPRIETARY INFORMATION - CLASS I (PUBLIC) 4.0 ENGINEERED SAFETY FEATURES 4.1 CONTAINMENT SYSTEM PERFORMANCE TLTR Appendix G presents the methods, approach, and scope for the TPO uprate containment evaluation for LOCA. The cmTent containment evaluations were performed at 102% of CLTP.

Although the nominal operating conditions change slightly because of the TPO uprate, the required initial conditions for containment analysis inputs remain the same as previously documented.

The following table summan zes the effect of the TPO uprate on van ous aspects of the containment system perfonnance.

Topic Key Parameters TPO Effect Shoit Teim Pressure and Temperatme Response Gas Temperature Break Flow and Energy Pressm e Break Flow and Energy Long-Term Suppression Pool Temperatme Response Bulk Pool Decay Heat CuITent Analysis Local Temperature with Decay Heat Based on 102% ofCLTP SRV Discharge Containment Dynamic Loads Loss-of-Coolant Accident Break Flow and Energy Loads Safety-Relief Valve Loads Decay Heat Sub-compa1tment Break Flow and Energy Pressm ization (Note I)

Containment Isolation The ability of containment Section 4.1.1 provides isolation valves (CIVs) and confirmation that motor- operators to perfo1m their operated valves (MOVs) are required functions is not capable of perfo1ming design affected because the basis functions at TPO evaluations have been conditions. perfo1med at 102% of CLTP.

Note:

1. The CGS cmTent analysis of sub-compaitment pressurization is based on the maximum break flow and energy of postulated pipe breaks between the RPV wall and the biological shield wall. GEH recently issued safety information communication SC 09-01 (Reference 25) to 4- 1

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NED0-33853 REVISION 0 NON- PROPRIETARY INFORMATION - CLASS I (PUBLIC)

Table 4-1 CGS ECCS-LOCA Analysis Results for GNF2 Fuel Parameter MELLLA Analysis Limit Nominal PCT 1,356°F NI A 1

Appendix K PCT 1,637°F

-< 2' 200°F < )

1 Licensing Basis PCT 1,700°F

-< 2' 200°F < )

Maximum Local Oxidation < l .0% ~ 17% ( l )

Core-Wide Metal-Water Reaction <0.1% ~ 1.0% (l)

Note:

1. 10 CFR 50 .46 ECCS-LOCA analysis acceptance criteria 4-6

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NED0-33853 REVISION 0 NON- PROPRIETARY INFORMATION - CLASS I (PUBLIC)

Table 5-1 Analytical Limits and Allowable Values for Current and TPO Power Level Paramete1* Current TPO Justification APRM High Neutron Flux-Fixed Scram (% RTP), AL 123 No change APRM STP - High (Scram) (2) (1) 4 STP-High Scram Clamp (%RTP) < >, AV 114.9 No change TLO STP-High Scram (%RTP) <3>, AV 0.63Wd + 64.0 0.62Wd + 62.9 (4)

SLO STP-High Scram (%RTP) <3>, AV 0.63(Wd - !!,,W) + 64.0 0.62(Wd - !!,, W) + 62.9 (4) 0.63Wd + 60.8 0 .62Wd + 59.8 APRM STP - High (Rod Block) (2) (1) 4 STP Rod Block Clamp (%R TP) <>, AV 111 No change TLO STP-High Rod Block (%RTP) <3l, AV 0.63Wd + 60.1 0.62Wd + 59.1 (4)

SLO STP-High Rod Block (%RTP) <3>, AV 0.63(Wd - !!,,W) + 60.1 0.62(Wd - !!,,W) + 59.1 (4) 0 .63Wd + 56.9 0.62Wd + 56.0 TSV & TCV Closure Scram Bypass - AL (%R TP) 30 29.5 (5)

MSL High Flow Isolation - ALs:

% rated steam flow 140 No change (5) psid 127.5 145.37 Rod Wo1th Minimizer LPSP - AL (%RTP) 10 No change (5)

Notes:

(1) CGS does not have ALs for these setpoint functions.

(2) No credit is taken in any safety analysis for flow biased setpoints.

(3) Wd is% recirculation drive flow where 100% drive flow is that required to achieve 100% core flow at 100% power and !!,,W is the difference in % drive flow between the TLO and SLO recirculation drive flow at the same core flow.

(4) These changes to the AV s are based upon the methodology approved by the NRC in Reference 1.

(5) All limits scaled for an uprate of 1.66% thermal power.

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NED0-33853 REVISION 0 NON- PROPRIETARY INFORMATION - CLASS I (PUBLIC)

Table 6-1 TPO Plant Electrical Characteristics Parameter Value Generator Output (MWe) 1 22i1)

Rated Voltage (kV) 25 Power Factor 0.997 Generator Output (MVA) 1,230 CmTent Output (Amps) 28,406 Isolated Phase Bus Duct Rating: (Amps)

Main Section 30,000 Delta Section 17,300 Auxilia1y Section 1,200 Main Transfo1mers Rating (MVA) 1,276 I 1,425 Note:

(1) Reactive power will be closely monitored at 1,227 MWe to ensure the 1,230 MVA rating of the main generator is not exceeded.

6-8

NED0-33853 REVISION 0 NON- PROPRIETARY INFORMATION - CLASS I (PUBLIC)

Table 6-2 Main Generator Ratings Comparison Design Maximum Nominal Power Level MVA @ 75 psig H2 MWe @ 75 psig H2 MVAR @ 75 psig H2 Existing 1,230 1,206 273 UpratedCl ) 1,230 1,227 50 Note:

(1) Operation at the uprated condition is not expected to have any adverse effect on the operation of the main generator. Operation in this range is still within the operating boundaries specified in station design analysis and operating procedmes. Reactive power will be closely monitored at 1,227 MWe to ensm e the 1,230 MVA rating of the main generator is not exceeded.

Table 6-3 Main Transformer Rating Comparison Power Level Design MVA at 65°C MVALoading Existing 1,276 I 1,425 1,276 I 1,425 Uprated (l) 1,276 I 1,425 1,276 I 1,425 Note:

(1) Operation at the uprated condition is not expected to have any effect on the operation of the main transfo1mer except if the spare main transformer is placed into service. The generator MWe will increase and MVAR will decrease, thus MVA will remain the same.

Operation in this range is still within the operating boundaries specified in station design analysis and operating procedmes.

6-9

NED0-33853 REVISION 0 NON- PROPRIETARY INFORMATION - CLASS I (PUBLIC)

Table 6-4 Normal Auxiliary Transformer Ratings Comparison Windin2 RatedMVAat Existing MVA TPOMVA Identification<1) 65°C Loading Loading E-TR-Nl/X - Winding 26.9 26.541 26.602 E-TR-Nl/Y - Winding 17.9 12.050 12.079 E-TR-N2-X - Winding <2) 29.05 30.765 30.765 Notes:

(1) Operation at the uprated condition is not expected to have any effect on the operation of the n01mal auxiliaiy transformers excluding E-TR-Nl during the highest temperature months. There is a potential for plant downpower if transformer cooling cannot be maintained at 100% plant loading.

(2) Operation in this range is still within the operating boundaries specified in station design analysis and operating procedmes.

6-10

NED0-33853 REVISION 0 NON- PROPRIETARY INFORMATION - CLASS I (PUBLIC)

Table 6-5 Start-Up Auxiliary Power Transformer Comparison Windin2 RatedMVAat Existing MVA TPOMVA Identification(I ) 65°C Loading Loading E-TR-S/X - Winding 28.66 20.988 20.988 E-TR-SN - Winding 40 33.275 33.366 Note:

(1) Operation at the uprated condition is not expected to have any effect on the operation of the sta1t -up auxilia1y power transformer. Operation in this range is still within the operating boundaries specified in station design analysis and operating procedures.

6-11

NED0-33853 REVISION 0 NON- PROPRIETARY INFORMATION - CLASS I (PUBLIC)

Table 6-6 FPC System Parameters Parameter CLTP TPO Number ofRHR/FPC trains 1/ 2 1/ 2 RHR heat exchanger flow rate, RHR/SW 3,500 I 7,200 gpm 3,500 I 7,200 gpm Fuel pool heat exchanger flow rate, 575 I 575 gpm 575 I 575 gpm SFP/RCC Design heat removal capacity (one RHR 42.7E+6 BTU/hr 42.7E+6 BTU/hr heat exchanger)

Design heat load, (2) fuel pool heat 8.00E+6 BTU/hr 8.00E+6 BTU/hr exchangers Fuel cycle (months) 24 24 Bulk pool temperature (No1mal Operation) < 125°F < 125 °F Bulk pool temperature (During Refueling) < 150°F < 150°F 6-12

NED0-33853 REVISION 0 NON- PROPRIETARY INFORMATION - CLASS I (PUBLIC)

Table 6-7 Effluent Discharge Comparison Parameter State Limit Current TPO 1 2 3 Flow (mgd) 5.6 and 9.4 4.5 Minimal change 4

Total Residual Halogen (mg!L) 0.1 < 0.1 No change 1 2 Chromium, total (µg!L) 8.2 and 16.4 < 1.85 Minimal change Zinc, total (µg!L) 53 1 and 1072 < 27 5 Minimal change PCBs No discharge No discharge No change The 126 priority pollutants (40 CFR No detectable No detectable No change 423 Appendix A) contained in amount amount chemicals added for cooling tower maintenance, except chromium and ZlllC pH 6.5-9.0 7.4-8.46 No change Acute Toxicity No acute No acute Minimal change toxicity7 toxicity8 Notes:

1. Average monthly eftluent limit means the highest allowable average of daily discharges over a calendar month.

2. Maximum daily eftluent limit is the highest allowable daily discharge.

3. Maximum daily average flow for September 2014 - August 2015.

4. Value verified by two samples collected at least 15 minutes apa1t prior to initiation of blowdown following biofouling treatments.

5. Maximum for September 2014 - August 2015.

6. Minimum and maximum pH range for September 2014 - August 2015.

7. No acute toxicity detected in a test concentration representing the ACEC. The ACEC equals 11% eftluent.

8. Based on quaiterly acute toxicity testing completed in 2015.

6- 13

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License Amendment Request to Revise Operating License and Technical Specifications for Measurement Uncertainty Recapture (MUR) Power Uprate Enclosure 12 Affidavits from Cameron (Caldon) Corporation Supporting the Withholding of Information in Enclosures 10 and 11 from Public Disclosure

Measurement Systems Caldon" Ultrasonics Technology Center 1000 McClaren Woods Drive Coraopolis, PA 15108 Tel 724-273-9300 Fax 724-273-9301

( CAMERON www.c-a-m.com May 18, 2016 CAW 16-01 Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555 APPLICATION FOR WITHHOLDING PROPRIETARY INFORMATION FROM PUBLIC DISCLOSURE

Subject:

Cameron Engineering Report ER-1049 Rev. 3 "Bounding Uncertainty Analysis for Thermal Power Determination at Columbia Nuclear Generating Station Using the LEPM ./+M System" Gentlemen:

This application for withholding is submitted by Cameron (Holding) Corporation, a Nevada Corporation (herein called "Cameron") on behalf of its operating unit, Caldon Ultrasonics Technology Center, pursuant to the provisions of paragraph (b)(l) of Section 2.390 of the Commission's regulations. It contains trade secrets and/or commercial information proprietary to Cameron and customarily held in confidence.

The proprietary information for which withholding is being requested is identified in the subject submittal. In conformance with 10 CPR Section 2.390, Affidavit CAW 16-01 accompanies this application for withholding setting forth the basis on which the identified proprietary information may be withheld from public disclosure.

Accordingly, it is respectfully requested that the subject information, which is proprietary to Cameron, be withheld from public disclosure in accordance with 10 CPR Section 2.390 of the Commission's regulations.

Correspondence with respect to this application for withholding or the accompanying affidavit should reference CAW 16-01 and should be addressed to the undersigned.

Very truly yours, Ernest . Hauser Director of Business Development Enclosures (Only upon separation of the enclosed confidential material should this letter and affidavit be released.)

May 18, 2016 CAW 16-01 AFFIDAVIT COMMONWEALTH OF PENNSYLVANIA:

SS COUNTY OF ALLEGHENY:

Before me, the undersigned authority, personally appeared Ernest M. Hauser, who, being by me duly sworn according to law, deposes and says that he is authorized to execute this Affidavit on behalf of Cameron (Holding) Corporation, a Nevada Corporation (herein called "Cameron") on behalf of its operating unit, Caldon Ultrasonics Technology Center, and that the averments of fact set forth in this Affidavit are true and correct to the best of his knowledge, information, and belief:

~"-/ __

Director of Business Development Sworn to and subscribed before me this I~ day of h'\~ ,2016 j~ A. ~.&¥-

Notary Public 1

May 18, 2016 CAW 16-01

1. I am the Director of Business Development of Caldon Ultrasonics Technology Center, and as such, I have been specifically delegated the function of reviewing the proprietary infonnation sought to be withheld from public disclosure in connection with nuclear power plant licensing and rulemaking proceedings, and am authorized to apply for its withholding on behalf of Cameron.

2. I am making this Affidavit in confonnance with the provisions of I 0 CFR Section 2.390 of the Commission's regulations and in conjunction with the Cameron application for withholding accompanying this Affidavit.

3. I have personal knowledge of the criteria and procedures utilized by Cameron in designating infonnation as a trade secret, privileged or as confidential commercial or financial infonnation.

4. Cameron requests that the infonnation identified in paragraph S(v) below be withheld from the public on the following bases:

Trade secrets and commercial infonnation obtained from a person and privileged or confidential The material and infonnation provided herewith is so designated by Cameron, in accordance with those criteria and procedures, for the reasons set forth below.

5. Pursuant to the provisions of paragraph (b) (4) of Section 2.390 of the Commission's regulations, the following is furnished for consideration by the Commission in determining whether the infonnation sought to be withheld from public disclosure should be withheld.

(i) The information sought to be withheld from public disclosure is owned and has been held in confidence by Cameron.

(ii) The infonnation is of a type customarily held in confidence by Cameron and not customarily disclosed to the public. Cameron has a rational basis for detennining the 2

May 18, 2016 CAW 16-01 types of information customarily held in confidence by it and, in that connection utilizes a system to determine when and whether to hold certain types of information in confidence. The application of that system and the substance of that system constitutes Cameron policy and provides the rational basis required. Furthermore, the information is submitted voluntarily and need not rely on the evaluation of any rational basis.

Under that system, information is held in confidence if it falls in one or more of several types, the release of which might result in the loss of an existing or potential advantage, as follows:

(a) The information reveals the distinguishing aspects of a process (or component, structure, tool, method, etc.) where prevention of its use by any of Cameron's competitors without license from Cameron constitutes a competitive economic advantage over other companies.

(b) It consists of supporting data, including test data, relative to a process (or component, structure, tool, method, etc.), the application of which data secures a competitive economic advantage, e.g., by optimization or improved marketability.

(c) Its use by a competitor would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, and assurance of quality, or licensing a similar product.

(d) It reveals cost or price information, production capacities, budget levels, or commercial strategies of Cameron, its customer or suppliers.

(e) It reveals aspects of past, present or future Cameron or customer funded development plans and programs of potential customer value to Cameron.

(f) It contains patentable ideas, for which patent protection may be desirable.

3

May 18, 2016 CAW 16-01 The information sought to be withheld is considered to be proprietary for the reasons set forth in paragraphs (a), (b) and (c), above.

There are sound policy reasons behind the Cameron system, which include the following:

(a) The use of such information by Cameron gives Cameron a competitive advantage over its competitors. It is, therefore, withheld from disclosure to protect the Cameron competitive position.

(b) It is information that is marketable in many ways. The extent to which such information is available to competitors diminishes the Cameron ability to sell products or services involving the use of the information.

(c) Use by our competitor would put Cameron at a competitive disadvantage by reducing his expenditure of resources at our expense.

(d) Each component of proprietary information pertinent to a particular competitive advantage is potentially as valuable as the total competitive advantage. If competitors acquire components of proprietary information, any one component may be the key to the entire puzzle, thereby depriving Cameron of a competitive advantage.

(e) Unrestricted disclosure would jeopardize the position of prominence of Cameron in the world market, and thereby give a market advantage to the competition of those countries.

(t) The Cameron capacity to invest corporate assets in research and development depends upon the success in obtaining and maintaining a competitive advantage.

(iii) The information is being transmitted to the Commission in confidence, and, under the provisions of 10 CFR §§ 2. 390, it is to be received in confidence by the Commission.

4

May 18, 2016 CAW 16-01 (iv) The information sought to be protected is not available in public sources or available infonnation has not been previously employed in the same manner or method to the best of our knowledge and belief.

(v) The proprietary information sought to be withheld are the submittals titled:

Engineering Report ER-1049 Rev. 3 "Bounding Uncertainty Analysis for Thermal Power Determination at Columbia Nuclear Generating Station Using the LEFM ./+M System" It is designated therein in accordance with 10 CFR §§ 2.390(b)(l)(i)(A,B), with the reason(s) for confidential treatment noted in the submittal and further described in this affidavit. This information is voluntarily submitted for use by the NRC Staff in their review of the accuracy assessment of the proposed methodology for the LEFM CheckPlus M System used by Columbia Generating Station for flow measurement at the licensed reactor thermal power level of 3,544 MWt.

Public disclosure of this proprietary information is likely to cause substantial harm to the competitive position of Cameron because it would enhance the ability of competitors to provide similar flow and temperature measurement systems and licensing defense services for commercial power reactors without commensurate expenses. Also, public disclosure of the information would enable others to use the information to meet NRC requirements for licensing documentation without the right to use the information.

The development of the technology described in part by the information is the result of applying the results of many years of experience in an intensive Cameron effort and the expenditure of a considerable sum of money.

In order for competitors of Cameron to duplicate this information, similar products would have to be developed, similar technical programs would have to be performed, and a significant manpower effort, having the requisite talent and experience, would have to be expended for developing analytical methods and receiving NRC approval for those methods.

Further the deponent sayeth not.

5

Measurement Systems Caldone Ultrasonics Technology Center 1000 Mcclaren Woods Drive Coraopolis, PA 15108 Tel 724-273-9300 Fax 724-273-9301

@ CAMERON www.c-a-m.com May 18, 2016 CAW 16-02 Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555 APPLICATION FOR WITHHOLDING PROPRIETARY INFORMATION FROM PUBLIC DISCLOSURE

Subject:

Cameron Engineering Report ER-1074 Rev. 0 "Meter Factory Calculation and Accuracy Assessment for Columbia Nuclear Generating Station" Gentlemen:

This application for withholding is submitted by Cameron (Holding) Corporation, a Nevada Corporation (herein called "Cameron") on behalf of its operating unit, Caldon Ultrasonics Technology Center, pursuant to the provisions of paragraph (b)(l) of Section 2.390 of the Commission's regulations. It contains trade secrets and/or commercial infonnation proprietary to Cameron and customarily held in confidence.

The proprietary information for which withholding is being requested is identified in the subject submittal. In conformance with I 0 CFR Section 2.390, Affidavit CAW 16-02 accompanies this application for withholding setting forth the basis on which the identified proprietary infonnation may be withheld from public disclosure.

Accordingly, it is respectfully requested that the subject information, which is proprietary to Cameron, be withheld from public disclosure in accordance with 10 CFR Section 2.390 of the Commission's regulations.

Correspondence with respect to this application for withholding or the accompanying affidavit should reference CAW 16-02 and should be addressed to the undersigned.

Very truly yours, Enclosures (Only upon separation of the enclosed confidential material should this letter and affidavit be released.)

May 18, 2016 CAW 16-02 AFFIDAVIT COMMONWEALTH OF PENNSYLVANIA:

SS COUNTY OF ALLEGHENY:

Before me, the undersigned authority, personally appeared Ernest M. Hauser, who, being by me duly sworn according to law, deposes and says that he is authorized to execute this Affidavit on behalf of Cameron Holding Corporation, a Nevada Corporation (herein called "Cameron") on behalf of its operating unit, Caldon Ultrasonics Technology Center, and that the averments of fact set forth in this Affidavit are true and correct to the best of his knowledge, information, and belief:

E~

Director of Business Development Sworn to and subscribed before me this I0 day of h'\&.h ,2016 J,MJ\A0o A. ~- QJ}iv~

Notary Public MMONWEALTH OF PENN YLVANI NOTARIAL SEAL Francn A. l *l1, NoWy Public Coraopoll1 Bon>, AUtgheny County lty Commlaslon Expil'tl Nov. 25, 20tB 1

May 18, 2016 CAW 16-02

1. lam the Director of Business Development of Caldon Ultrasonics Technology Center, and as such, I have been specifically delegated the function of reviewing the proprietary information sought to be withheld from public disclosure in connection with nuclear power plant licensing and rulemaking proceedings, and am authorized to apply for its withholding on behalf of Cameron.

2. I am making this Affidavit in conformance with the provisions of 10 CFR Section 2.390 of the Commission's regulations and in conjunction with the Cameron application for withholding accompanying this Affidavit.

3. I have personal knowledge of the criteria and procedures utilized by Cameron in designating information as a trade secret, privileged or as confidential commercial or financial information.

4. Cameron requests that the information identified in paragraph S(v) below be withheld from the public on the following bases:

Trade secrets and commercial information obtained from a person and privileged or confidential The material and information provided herewith is so designated by Cameron, in accordance with those criteria and procedures, for the reasons set forth below.

5. Pursuant to the provisions of paragraph (b) (4) of Section 2.390 of the Commission's regulations, the following is furnished for consideration by the Commission in determining whether the information sought to be withheld from public disclosure should be withheld.

(i) The information sought to be withheld from public disclosure is owned and has been held in confidence by Cameron.

(ii) The information is of a type customarily held in confidence by Cameron and not customarily disclosed to the public. Cameron has a rational basis for determining the 2

May 18, 2016 CAW 16-02 types of information customarily held in confidence by it and, in that cormection utilizes a system to determine when and whether to hold certain types of information in confidence. The application of that system and the substance of that system constitutes Cameron policy and provides the rational basis required. Furthermore, the information is submitted voluntarily and need not rely on the evaluation of any rational basis.

Under that system, information is held in confidence if it falls in one or more of several types, the release of which might result in the loss of an existing or potential advantage, as follows:

(a) The information reveals the distinguishing aspects of a process (or component, structure, tool, method, etc.) where prevention of its use by any of Cameron's competitors without license from Cameron constitutes a competitive economic advantage over other companies.

(b) It consists of supporting data, including test data, relative to a process (or component, structure, tool, method, etc.), the application of which data secures a competitive economic advantage, e.g., by optimization or improved marketability.

(c) Its use by a competitor would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, and assurance of quality, or licensing a similar product.

(d) It reveals cost or price information, production capacities, budget levels, or commercial strategies of Cameron, its customer or suppliers.

(e) It reveals aspects of past, present or future Cameron or customer funded development plans and programs of potential customer value to Cameron.

(f) It contains patentable ideas, for which patent protection may be desirable.

3

May 18, 2016 CAW 16-02 The information sought to be withheld is considered to be proprietary for the reasons set forth in paragraphs (a), (b) and {c), above.

There are sound policy reasons behind the Cameron system, which include the following:

(a) The use of such information by Cameron gives Cameron a competitive advantage over its competitors. It is, therefore, withheld from disclosure to protect the Cameron competitive position.

(b) It is information that is marketable in many ways. The extent to which such information is available to competitors diminishes the Cameron ability to sell products or services involving the use of the information.

(c) Use by our competitor would put Cameron at a competitive disadvantage by reducing his expenditure of resources at our expense.

(d) Each component of proprietary information pertinent to a particular competitive advantage is potentially as valuable as the total competitive advantage. If competitors acquire components of proprietary information, any one component may be the key to the entire puzzle, thereby depriving Cameron of a competitive advantage.

(e) Unrestricted disclosure would jeopardize the position of prominence of Cameron in the world market, and thereby give a market advantage to the competition of those countries.

(f) The Cameron capacity to invest corporate assets in research and development depends upon the success in obtaining and maintaining a competitive advantage.

(iii) The information is being transmitted to the Commission in confidence, and, under the provisions of 10 CFR §§ 2. 390, it is to be received in confidence by the Commission.

4

May 18, 2016 CAW 16-02 (iv) The information sought to be protected is not available in public sources or available information has not been previously employed in the same manner or method to the best of our knowledge and belief.

(v) The proprietary information sought to be withheld are the submittals titled:

Cameron Engineering Report ER- 1074 Rev. 0 "Meter Factor Calculation and Accuracy Assessment for Columbia Nuclear Generating Station" It is designated therein in accordance with 10 CFR §§ 2.390(b)(l)(i)(A,B), with the reason(s) for confidential treatment noted in the submittal and further described in this affidavit. This information is voluntarily submitted for use by the NRC Staff in their review of the accuracy assessment of the proposed methodology for the LEFM CheckPlus M System used by Columbia Nuclear Generating Station for flow measurement at the licensed reactor thermal power level of 3,544 MWt.

Public disclosure of this proprietary information is likely to cause substantial harm to the competitive position of Cameron because it would enhance the ability of competitors to provide similar flow and temperature measurement systems and licensing defense services for commercial power reactors without commensurate expenses. Also, public disclosure of the information would enable others to use the information to meet NRC requirements for licensing documentation without the right to use the information.

The development of the technology described in part by the information is the result of applying the results of many years of experience in an intensive Cameron effort and the expenditure of a considerable sum of money.

In order for competitors of Cameron to duplicate this information, similar products would have to be developed, similar technical programs would have to be performed, and a significant manpower effort, having the requisite talent and experience, would have to be expended for developing analytical methods and receiving NRC approval for those methods.

Further the deponent sayeth not.

5

License Amendment Request to Revise Operating License and Technical Specifications for Measurement Uncertainty Recapture (MUR) Power Uprate Enclosure 13 Columbia Calculation NE-02-15-08, Heat Balance Determination for Rated Thermal Power, Revision 0

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Calculation No.

ENERGY NORTHWEST NE-02-15-08 CALCULATION Rev. No. Page No.

0 1 I

Quality Class II I Discipline Nuclear I Status I F or S*

[g]F Os I Initiating Documents PDC 14942 Equipment Piece No.

Plant Process Computer (PPC)

Study Calculations shall be used only for the purpose of evaluating alternate design options or assisting the engineer in performing assessments.

TITLE Heat Balance Determination for Rated Thermal Power PERFORMANCENERIFICATION RECORD Prepared By: ESFL Qualified 0 Yes Bob Goff Print Name Verified By: ESFL Qualified 0 Ye~(Veodo' 10 CFR Appeodix BJ Don Kinoshita Print Name ~@ Date Owner's Review By: ESFL Qualified 0 Yes 0 No Print Name Signature Date Approved By:

Print Name Signature Date Unverified Assumption : D Yes D No If Yes, Resolution per AR# _ _

Signature below denotes vedfication and resolution of unverified assumptions.

Print Name Signature Date INCORPORATED ENGINEERING CHANGES (CMR, PDC, MALT, etc.)

Only list design changes If Incorporated and shown as a current outstanding EC N/A REVISION NO. REVISION

SUMMARY

O Initial Issue 18645 R8

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CALCULATION 0 I 3 VERIFICATION CHECKLIST Item: YE~ NO NIA Initials Cover sheets properly completed. fQJ D D vk Incorporated engineering change (EC) is complete and accurate. ~ D D DlL Impacted output interface documents identified and updated. [QI" D D *D~

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Calculation results reasonable and correctly described in the Summary of Results and [121" D D Conclusions. DLl-Summary of Results and Conclusions includes discussion of margin . w D D DlL.

Calculated values within the instrument display range. 111" D D n ll-Listed attachments included and complete. ~ D D I) \l-w~

Computer program identified with version and revision. D D NIA Computer output included with program name, revision, run date on first page. D D ~ i-J I A Computer program verification/validation addressed. D D w ~/A Native document files located in the appropriate EN file in accordance with DES-4-19. ~ D D ' ] ) j.L.._

0 0 D 0 0 0 Calculation Checking Method Applicable pages/sections Direct Step - By - Step Check Alternate Calculation 0 Verified By: Calculation Verifier(s)

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B-PPC channel non-random drift 95/95 statistical analysis Preparer: _R_a'"""lp..._h-"'-B.....

er......g._e_r --~-=---'--=---++----- ....J.....2....

Date: _ _--5..,/_25 o_._16.__ _ __

Reviewer: _ _.D_.o......n__K __i._no.....s""""h...._i..... ~

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Purpose and Result s This section develops the uncertainty in the PPC term as a non-random drift of 1.81% to be applied as

+1.81%/-1.81%. A 95/95 methodology is utilized from ISA RP67.04.02-2000 to provide 95% confidence that this range will address 95% of the variation. Further conservatism is added by enveloping all sources of uncertainty into a non-random drift and applying it in the most limiting direction. The computer points that have this non-random drift are associated with analogue Columbia GS B-PPC points.

Input Data The test signals for PPC reference B019 were transmitted by memo, Steele to Menocal, May 24, 2016, and consist of 64,668 data obtained from Plant Process Computer point B019MV via the "eDNA" historian for the dates 1 2009 to 5-16-2016 (see AR 344042-69). Of these data, 62,277 are labeled "OK" for use. The data consist of output voltages given an input signal of 100 mV. A plot of the data is shown below where the roughly 2,400 bad data are replaced with the previous good data point for visual clarity.

The signal evidences a sinusoidal error with a slow drift of about 0.11% per year (calculated by the linear trend line fit, which gives .00033%/day). The sinusoidal pattern correlates very well with annual temperature variation, so it is a long-term impact that is inconsistent with random fluctuations as random errors are described in procedure EES-4. It is conservative to treat the full signal variation (peak to trough) as a non-random drift (or systematic uncertainty), since it will add to the total uncertainty in full in the limiting direction, instead of being a random factor within the SRSS radical. Note that the final result incorporates the annual temperature effects, the slow drift, and the random noise in a single term.

Signal variation 101 140 y = -0.00033x + 112.71 100.S 120 100 100 99.S 80 99 60 98.S 40 98 20 97.5 0 97 -20 9/9/7.008 l/22/2010 6/6/2011 10/18/2012 3/2/2014 7/15/2015 1

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Assumptions

1. The example drift from point B019 can be applied to all other B-PPC channels.

Methodology and Body of Calculation The 95/ 95 value for peak to trough variation is developed using Annex E to ISA RP67.04.02-2000. Since calibration is performed on a two year cycle, the variation is observed in rolling periods of two years (specifically, the six overlapping periods of 1/1/2009 to 1/1/2011; 1/1/2010 to 1/1/2012; 1/1/2011 to 1/1/2013; 1/1/2012 to 1/1/2014; 1/1/2013 to 1/1/2015; and 1/1/2014 to 1/1/2016). The worst 95/95 limits for those periods is discovered, which will be seen to be 1/1/2013 to 1/1/2015, with values of 99.916 and 98.106, making a total peak to trough uncertainty of 1.81%.

The methodology for a non-normally distributed data set is used. That the data is not a normal distribution can be determined in a variety of ways, one of which is to view the above data graph. This is not a signal that is centered on a mean, but rather a signal that drifts back and forth from a high to low region . A Kurtosis test result is +3 if perfectly bell-shaped, but our results are around -1, meaning fairly flat. The full data set is binned into 0.1 widths and plotted below, with a normal curve based on the data set mean and standard deviation for comparison. It is visually clear that the normal distribution is not a good match .

Data in 0.1 bin widths 7000 0.8 6000 0.7 0.6 5000 0.5 4000 0.4 3000 0.3 2000 0.2 1000 0.1 0 0 97.5 98 98.S 99 99.S 100 100.S 101 For the normal curve above, the mean is 99.16 and the standard deviation is 0.5356 (Excel function average() and stdev() applied to all good data points). This implies a 2.5% and 97.5% value of 99.16 +/-

1.96* .5356 =98.11, 100.21, or a 2.10% peak to trough value. This is higher than the final result below in part because it includes 7 years of drift (instead of our two-year period between calibration), but it still provides a reasonable order of magnitude check for our final result of 1.81%.

The ISA 67 .04.02-2000 Annex E methodology for the 95/95 limits of a non-normal data set are presented in Section E.4.2 using a minimum pass probability model, which is also described in Beggs, W.J., Statistics for Nuclear Engineers and Scientists, Part 1: Basic Statistical Inference," DOE Research and Development Report No. WAPD-TM-1292, February, 1981. An example on page 94 of Beggs verifies the approach .

2

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Equation E-4 of ISA 67.04.02-2000 is:

Pu. 1 =x/n +/- z((1 /n)(x/n)[1 - (x/n)]}112 [E-4)

For our purposes, Pu,1 is either 2.5% or 97.5%, which determines the limits for the first 95 in our 95/95 confidence. N is our sample size, which for two years is on the order of 8760 hours0.101 days 2.433 hours 0.0145 weeks 0.00333 months of data*2 =17520 data. Z is the confidence interval from a normal curve, which is 1.96 for our second 95 in the 95/95 confidence description. x/n is a term that is developed from the equation by iteration. It is defined as the fraction of data "outside the pass/fail criterion ." We will be using x/n to determine the range at which we can say, with 95/95 confidence, that we have reached the 2.5% or 97.5% value.

By examination of Equation E-4, we see that if the sample size was infinite (1/n = O}, then Pu,1 would equal x/n . That is, the 2.5% 95/95 value would be the 2.5% largest value when ranking all values from minimum to maximum. Similarly, with an infinite sample size, the 97.5% value would be the 97.5% term in our ranked list of data. With our sample size of 1/n =1/17,500, it takes a x/n value on the order of 2.3% to produce a value of Pu.1 = 2.5%, and x/n = 97 .7% to produce a Pu,1 value of 97.5%. In other words, when we rank our data, we can say with 95% confidence that our observed range from 2.3% to 97.7%

will encompass all future data.

Mathematically, the formula for our first period, which contains 17,404 data, is:

0.025 = Pu,2 .5 = x/n +/- z{(1 /n)(x/n)[1 - (x/n)]}112 =0.022783 + 1.96*(1/17404* .022783*(1-.022783})0*5 Here the value of .022783 was found by iteration, and is a function of sample size only. Similarly, the upper limit is found from 0.975 = Pu,97. 5 =x/n +/- z{(1 /n)(x/n)[1 -(x/n)]}112 = 0.977217-1.96*(1/17404*.977217*(1-.977217)}0.s The process to determine the 95/95 limits is to first take the full seven years of data, 64,668 items, and arrange in cells 03:064670 in an Excel worksheet PPC. Column C is the designation of "OK" or "Unreliable" or "Unavailable." The top of our spreadsheet of data appears like this:

A B c D 1

2 FEEDFLOWCALIBRATIONST 3 1/1/2009 0:00:00 OK 100.1683 4 1/1/2009 1 :00:00 OK 100.1683 5 1/1/2009 2:00:00 OK 100.2104 6 1/1/2009 3:00:00 OK 100.1683 7 1/1/2009 4:00:00 OK 100.2104 8 1/1/2009 5:00:00 OK 100.2104 9 1/1/2009 6:00:00 OK 100.1683 10 1/1/2009 7:00:00 OK 100.1683 11 1/1/2009 8:00:00 OK 100.1683 12 1/1/2009 9:00:00 OK 100.1683 13 1/1/2009 10:00:00 OK 100.1683 3

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Our six periods of data are characterized as follows, with the solution of Equation E-4 in the right two columns.

Two year periods Period (1/1 to 1/1) Rows n Pu.2.5% Pu,97.5%

1 2009-2011 3 to 17523 17404 0.022783 0.977217 2 2010-2012 8763 to 26283 17237 0.022773 0.977227 3 2011-2013 17523 to 35067 17264 0.022775 0.977225 4 2012-2014 26283 to 43827 17465 0.022787 0.977213 5 2013-2015 35067 to 52587 17425 0.022784 0.977216 6 2014-2016 43827 to 61347 15726 0.022673 0.977327 Here the sample size n is calculated by typical formula =COUNTIF{PPC!C3:C17523,"0K"). The 6th period has bad data in May to July of 2015 causing the low value of n. The total number of rows per period is either 2*8760 = 17520, or 17544 for the periods 3 and 4 that include the leap year 2012.

Next, the good data is collected in cells k3:k17523 in spreadsheet ANALYSIS with equations such as

=IF(PPC!C3="0K",PPC!D3,""), replicated to capture all of the good data in the first two years. Those results are copied, pasted into cells L3:L17523, and then sorted from minimum to maximum. The 2.5%

limit is then the one in row 0.022783*n +2, and the 97.5% limit is in row 0.977217*n +2. The 95/95 peak to trough distance is then the 97.5% value minus the 2.5% value.

A similar section, covering the limiting Period 5, is as follows:

AE AF AG AH 2 Period 5 (unsort/sort) 3 99.62128 97.85396 2.5% 2.28%

4 99.62128 97.85396 438 399 5 99.62128 97.85396 98.14851 98.10643 6 99.57921 97.89604 7 99.62128 97.89604 97.50% 97.72%

8 99.57921 97.89604 16991 17030 9 99.62128 97.89604 99.91583 99.91583 10 99.62128 97.89604 1.767324 1.809396 Here the good data from 1/1/2013 to 1/1/2015 are in cells AE3:AE17523. They are copied into column AF, sorted, and fill up cells AF3:AF17427 (fewer than in column AE because of the blanks associated with bad data). The data in column AG is a check; the 2.5% row is row 438, calculated as

=ROUND{AG3*17425,0)+2. The 2.5% value is =AF438 and is 98.14851. Similarly, the 97.5% row is row 4

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16991, and that value is 99.91583. The difference is 1.767324, which would be our 95% range if the data set included all possible data.

The actual results are in column AH, where we look at the range from the 2.28% row to the 97.72% row, and find a total difference of 1.81%. This is our 95/95 value for greatest peak to trough difference of any of our examined periods.

Aside: note that there was no numeric difference between the 97.5% row 16991 and the 97.7% row 17030. That is not unusual. The data is reported to 5 digits after the decimal, but the actual values are limited to a set of values. Hence in this case, rows 16835 through 17353 were all 99 .91583.

Conclusion The final results are shown below.

Period Two year periods ( 1/1 to 1/1 ) 2.5% limit 97.5% limit 95/95 Range 1 2009-2011 98.61 100.13 1.51 2 2010-2012 98.53 99.87 1.35 3 2011-2013 98.49 99.96 1.47 4 2012-2014 98.19 99.92 1.73 5 2013-2015 98.11 99.92 1.81 6 2014-2016 98.02 99.79 1.77 The limiting 95/95 range is +/-1.81%. This value can be used as a non-random drift for all Columbia GS B-PPC points.

The quantified analysis and data are contained in the Excel file for the B-PPC eDNA Excel drift data for 95-95 analysis.xlsx. This file can be found in the EDMS folder of CGS AR 344042. The excel quantified analysis and data was developed for the 95/95 analysis from the PPC excel spreadsheet attached to the CGS transmittal document for B-PPC drift data as file B-PPC eDNA 1-1-09 to 5-16-16 per AR 344042-69 and also included in the EDMS folder of AR 344042 .

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v t e Letter Sequence Request
CAC:MF8060, Clarify Application of Setpoint Methodology for LSSS Functions (Open)
Initiation Request, Request Acceptance, Acceptance Supplement Administration Withholding Request Acceptance, Withholding Request Acceptance Questions 27 October 2016 1 November 2016 16 November 2016 22 November 2016 4 January 2017 4 January 2017 Answers 8 December 2016 12 January 2017 16 February 2017 Results Approval... Other: ML16207A016, ML16207A033
MONTHYEAR2016-08-18 [Table View]

GO2-16-096, License Amendment Request to Revise Operating License and Technical Specifications for Measurement Uncertainty Recapture Power Uprate

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