Text

Issuance of Amendment Measurement Uncertainty Recapture Power Uprate
	ML13364A131
Person / Time
Site:	Fermi
Issue date:	02/10/2014
From:	Thomas Wengert; Plant Licensing Branch III
To:	Plona J; DTE Electric Company
	Wengert T, 415-4037
References
	TAC MF0650
	Download: ML13364A131 (92)
	v • d • e

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 February 10, 2014 Mr. Joseph H. Plena Senior Vice President and Chief Nuclear Officer DTE Electric Company Fermi 2-210 NOC 6400 North Dixie Highway Newport, Ml 48166

SUBJECT:

FERMI 2- ISSUANCE OF AMENDMENT RE: MEASUREMENT UNCERTAINTY RECAPTURE POWER UPRATE (TAC NO. MF0650)

Dear Mr. Plena:

The U.S. Nuclear Regulatory Commission has issued the enclosed Amendment No. 196 to Facility Operating License No. NPF-43 for the Fermi 2 facility. The amendment is in response to your application dated February 7, 2013 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML13043A659), as supplemented by letters dated March 8, 2013 (ADAMS Accession No. ML13070A197), April 5, 2013 (ADAMS Accession No. ML13095A456), June 7, 2013 (ADAMS Accession No. ML13161A080), July 15, 2013 (ADAMS Accession No. ML13197A121), and September 27, 2013 (ADAMS Accession No. ML13273A464).

The amendment revises the Operating License and Technical Specifications to implement an increase of approximately 1.64 percent in rated thermal power from the current licensed thermal power of 3430 megawatts thermal (MWt) to 3486 MWt. The changes are based on increased feedwater flow measurement accuracy, which will be achieved by utilizing Cameron International (formerly Caldon) CheckPius' Leading Edge Flow Meter (LEFM) ultrasonic flow measurement instrumentation.

A copy of our safety evaluation is also enclosed. The Notice of Issuance will be included in the Commission's biweekly Federal Register notice.

Sincerely, Thomas J. Wengert, Senior Project Manager Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-341

Enclosures:

1. Amendment No. 196 to NPF-43

2. Safety Evaluation cc w/encls: Distribution via ListServ

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 DTE ELECTRIC COMPANY DOCKET NO. 50-341 FERMI2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 196 License No. NPF-43

1. The U.S. Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment by the DTE Electric Company (DTE, the licensee) dated February 7, 2013, as supplemented by letters dated March 8, 2013, April5, 2013, June 7, 2013, July 15, 2013, and September 27,2013, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; '

B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Facility Operating License No. NPF-43 is hereby amended to read as follows:

Enclosure 1

Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 196, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated into this license. DTE Electric Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3. This license amendment is effective as of its date of issuance and shall be implemented upon startup from the Sixteenth Refueling Outage. Coincident with the implementation of this amendment, DTE will fulfill the Regulatory Commitments identified in Enclosure 6 of its February 7, 2013, license amendment request (DTE letter NRC-13-0004).

FOR THE NUCLEAR REGULATORY COMMISSION Michele G. Evans, Director Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications and Facility Operating *ucense Date of Issuance: February 10, 2014

ATTACHMENT TO LICENSE AMENDMENT NO. 196 FACILITY OPERATING LICENSE NO. NPF-43 DOCKET NO. 50-341 Replace the following pages of the Facility Operating License and Appendix A Technical*

Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

REMOVE INSERT License Page 3 License Page 3 1.1-6 1.1-6 3.3-2 3.3-2 3.3-6 3.3-6 3.3-7 3.3-7 3.3-8 3.3-8 3.3-10 3.3-10 3.4-1 3.4-1

(4) DTE Electric Company, 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 such 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; (5) DTE Electric Company, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (6) DTE Electric Company, 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.

C. This 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 DTE Electric Company is authorized to operate the facility at reactor core power levels not in excess of 3486 megawatts thermal ( 100% power) in accordance with conditions specified herein and in Attachment 1 to this license. The items identified in Attachment 1 to this license shall be completed as specified.

Attachment 1 is hereby incorporated into this license.

(2) Technical Specifications and Environmental Protection Plah The Technical Specifications contained in Appendix A, as revised through Amendment No. 196 and the Environmental Protection Plan contained in Appendix B, are hereby incorporated into this license. DTE Electric Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

(3) Antitrust Conditions DTE Electric Company shall abide by the agreements and interpretations between it and the Department of Justice relating to Article I, Paragraph 3 of the Electric Power Pool Agreement between DTE Electric Company and Amendment No. 196

Definitions 1.1 1.1 Definitions (continued)

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

REACTOR PROTECTION The RPS RESPONSE TIME shall be that time interval SYSTEM (RPS) RESPONSE from when the monitored parameter exceeds its RPS TIME trip setpoint at 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 assuming that:

a. The reactor is xenon free;

b. The moderator temperature is 68°F; .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.

(continued)

FERMI - UNIT 2 1.1-6 Amendment No.IJ4 196

RPS Instrumentation-3.3.1.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. ---------NOTE- -------- B.1 Place channel in one 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Not applicable for trip system in trip.

Functions 2.a. 2.b, 2.c. 2.d. and 2.f.

.... ------- ................. -OR B.2 Place one trip system 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months One or more Functions in trip.

with one or more required channels inoperable in both trip systems.

c. One or more Functions C.1 Restore RPS trip 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months with RPS trip capability.

capability not maintained.

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

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

and referenced in Table 3.3.1.1-1.

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

(continued)

FERMI - UNIT. 2 3.3-2 Amendment No.134, 139 196

RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3. 3. 1. 1. 12 ------------------NOTE- -----------------

For Function 2.a. 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.

Perform CHANNEL FUNCTIONAL TEST. 184 days SR 3.3.1.1.13 Perform CHANNEL FUNCTIONAL TEST. 18 months SR 3.3.1.1.14 Perform CHANNEL CALIBRATION. 18 months SR 3.3.1.1.15 Perform LOGIC SYSTEM FUNCTIONAL TEST. 18 months SR 3.3.1.1.16 Verify Turbine Stop Valve-Closure and 18.months Turbine Control Valve Fast Closure.

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

(continued)

FERMI - UNIT 2 . 3. 3-6 Amendment No.+/-d4 196

RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3. 3. 1 . 1. 17 ------------------NOTES- ----------------

1. Neutron detectors are excluded.

2. For Function 5 "n" equals *4 channels for the purpose of determining the

STAGGERED TEST BASIS Frequency. .

Verify the RPS RESPONSE TIME is within 18 months on a limits. STAGGERED TEST BASIS SR 3 . 3 . 1 . 1. 18 ------------------NOTE- ------------------

Neutron detectors are excluded.

Perform CHANNEL CALIBRATION. 24 months SR 3.3.1.1.19 Perform LOGIC SYSTEM FUNCTIONAL TEST. 24 months SR 3.3.1.1.20 Verify OPRM is not bypassed when APRM 24 months Simulated Thermal Power is~ 27.5% and recirculation drive flow is< 60% of rated recirculation drive flow.

FERMI - UNIT 2 3.3-7 Amendment No.134, 139, 151 196

RPS Instrumentation 3.3.1.1 Table 3.3.1.1-1 (page 1 of 3)

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 :5 122/125 SR 3.3.1.1.4 divisions of SR 3.3.1.1.6 full scale SR 3.3.1.1.7 SR 3.3.1.1.11 SR 3.3.1.1.15 3 SR 3.3.1.1.1 :5 122/125 SR 3.3.1.1.5 divisions of SR 3.3.1.1.11 full seale SR 3. 3 .1.1.15

b. Inop 2 3 G SR 3.3.1.1.4 NA SR 3. 3 .1.1.15 5(a) 3 SR 3.3.1.1.5 NA SR 3.3.1.1.15

2. Average Power Range Monitors

a. Neutron Flux- Up sea 1e 2 3(C) G SR 3.3.1.1.2 :5 20% RTP (Setdown) SR 3.3.1.1.7 SR 3.3.1.1.8 SR 3. 3 .1.1.12 SR 3.3.1.1.18

b. Simulated Thermal 1 F SR 3.3.1.1.2 :5 0.62 (W-&W)

Power- Upsca 1e SR 3.3 .1.1.3 + 63.1% RTP SR 3.3.1.1.8 and :5 115. 5%

SR 3.3.1.1.12 RTP(b)

SR 3.3.1.1.18(d)(e)

(continued)

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

(b) &W = 8% when reset for single loop operation per LCO 3.4.1. "Recirculation Loops Operating."

Otherwise &W = 0%.

(c) Each APRM channel provides inputs to both trip systems.

(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 tol~rance around the Nominal Trip Setpoint (NTSP) at the completion of the surveillance: otherwise. the channel shall be declared inoperable. .Setpoirits more conservative than the NTSP are acceptable provided that the as-found and as-left tolerances apply to the actual setpoint implemented in the surveillance procedures (field setting) to confirm channel performance. The NTSP and the methodologies used to determine the as-found and as* left tolerances are specified in the Technical Requirements Manual.

FERMI - UNIT 2 3.3-8 Amendment No. +/-J4 196

RPS Instrumentation 3.3.1.1 Table 3.3.1.1*1 (page 3 of 3)

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. Scram Discharge Volume Water Leve 1 -High

a. Level 1,2 2 G SR 3.3.1.1.1 s 596 ft.

Transmitter SR 3.3.1.1. 9 0 inches SR .3.3.1.1.10 SR 3. 3 .1.1.14 SR 3.3.1.1.15 5<al 2 SR 3.3.1.1.1 s 596 ft.

SR 3.3.1.1.9 0 inches SR 3.3.1.1.10 SR 3.3.1.1.14.

SR 3. 3 .1.1.15

b. Float Switch 1,2 2 G SR 3.3.1.1.9 s 596 ft.

SR 3.3.1.1.14 0 inches SR 3.3.1.1.15 5(a) 2 SR 3.3.1.1.9 s 596 ft.

SR 3. 3 .1.1.14 0 inches SR 3. 3 .1.1.15

9. Turbine Stop Va 1ve- ~ 29.5% 4 E SR 3.3.1.1.9 s 7% closed Closure RTP SR 3. 3 .1.1.14 SR 3.3.1.1.15 SR 3.3.1.1.16.

SR 3.3.1.1.17

10. Turbine Control Valve ~ 29.5% 2 E SR 3.3.1.1. 9 NA Fast Closure RTP SR 3.3.1.1.15 SR 3. 3 .1.1.16 SR 3.3.1.1.17

11. Reactor Mode Switch- 1,2 2 G SR 3.3.1.1.13 NA Shutdown Position SR 3.3.1.1.15 5<a> 2 SR 3.3.1.1.13 NA SR 3.3.1.1.15

12. Manual Scram 1.2 2 G SR 3.3.1.1.5 NA SR 3.3.1.1.15 5(a) 2 SR 3.3.1.1.5 NA SR 3. 3 .1.1.15 (a) With any control rod withdrawn from a core cell containing one or more fuel assemblies.

FERMI - UNIT 2 3.3-10 Amendment No. +/-J4 196

Recirculation.Loops Operating 3.4.1 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.1 Recirculation Loops Operating LCD 3.4.1 Two recirculation loops with matched recirculation loop jet pump flows shall be in operation; OR One recirculation loop may be in operation provided the following limits are applied when the associated LCD is applicable:

1. LCD 3. 2.1. "AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)," single loop operation limits specified in the COLR; .

2. LCD 3. 2. 2, "MINIMUM CRITICAL POWER RATIO (MCPR)." single loop operation limits specified in the COLR;

3. LCO 3. 3 .1.1. "Reactor Protection System (RPS)

Instrumentation," Function 2.b (Average Power Range Monitors Simulated Thermal Pow~r-Upscale) Allowable Value of Table 3.3.1.1-1 is reset for single loop operation. when in MODE 1; and

4. THERMAL POWER is s 66.1% RTP.

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

Application of the required limitations for single loop operation may be delayed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after transition from two recirculation loop operations to single recirculation loop operation.

APPLICABILITY: MODES 1 and 2.

FERMI - UNIT 2 3.4.1 Amendment No. 134,139 196

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 196 TO FACILITY OPERATING LICENSE NO. NPF-43 DTE ELECTRIC COMPANY FERMI2 DOCKET NO. 50-341

1.0 INTRODUCTION

By letter to the U.S. Nuclear Regulatory Commission (NRC) dated February 7, 2013 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML13043A659), as supplemented by letters dated March 8, 2013 (ADAMS Accession No. ML13070A197), April5, 2013 (ADAMS Accession No. ML13095A456, June 7, 2013 (ADAMS Accession No. ML13161A080, July 15, 2013 (ADAMS Accession No. ML13197A121),

and September 27, 2013 (ADAMS Accession No. ML13273A464), DTE Electric Company (DTE, the licensee), formerly Detroit Edison Company, requested changes to the technical specifications (TSs) and facility operating .license for the Fermi Unit 2 Power Plant (Fermi 2).

The proposed changes would revise the Operating License and TSs to implement an increase of approximately 1.64 percent in rated thermal power (RTP) from the current licensed thermal power (CL TP) of 3430 megawatts thermal (MWt) to 3486 MWt. The changes are based on increased feedwater (FW) flow measurement accuracy, which will be achieved by utilizing Cameron International (formerly Caldon) CheckPius rM Leading Edge Flow Meter (LEFM) ultrasonic flow measurement instrumentation. The LEFM instrumentation was installed in Fermi 2 in 2010.

The supplements dated March 8, April 5, June 7, July 15, and September 27, 2013, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination as published in the Federal Register on June 11, 2013 (78 FR 35069).

2.0 BACKGROUND

2.1 Measurement Uncertainty Recapture (MUR) Power Uprates Nuclear power plants are licensed to operate at a specified maximum core thermal power, often called RTP. Appendix K, "[Emergency Core Cooling System] ECCS Evaluation Models," of Title 10 of the Code of Federal Regulations (1 0 CFR), Part 50, formerly required licensees to assume that the reactor has been operating continuously at a power level at least 1.02 times the licensed power level when performing loss-of-coolant accident (LOCA) and ECCS analyses.

This requirement was included to ensure that instrumentation uncertainties were adequately Enclosure 2

accounted for in the safety analyses. In practice, many of the design bases analyses assumed a 2 percent power uncertainty, consistent with 10 CFR Part 50, Appendix K.

A change to the Commission's regulations at 10 CFR Part 50, Appendix K, was published in the Federal Register on June 1, 2000 (65 FR 34913), which became effective July 31, 2000. This change allows licensees to use a power level less than 1.02 times the RTP for the LOCA and ECCS analyses, but not a power level less than the licensed power level, based on the use of state-of-the art FW flow measurement devices that provide a more accurate calculation of power. Licensees can use a lower uncertainty in the LOCA and ECCS analyses provided that the licensee has demonstrated that the proposed value adequately accounts for instrumentation uncertainties. As there continues to be substantial conservatism in other Appendix K requirements, sufficient margin to ECCS performance in the event of a LOCA is preserved.

However, this change to 10 CFR 50, Appendix K, did not authorize increases in licensed power levels for individual nuclear power plants. As the licensed power level for a plant is contained in its operating license, licensees seeking to raise the licensed power level must submit a license amendment request (LAR), which must be reviewed and approved by the NRC staff. Fermi 2 is currently licensed to operate at a maximum power level of 3430 MWt, which includes a 2 percent margin in the ECCS evaluation model to allow for uncertainties in RTP measurement.

The license amendment would reduce this uncertainty to 0.36 percent.

In order to provide guidance to licensees seeking an MUR power uprate on the basis of improved FW flow measurement, the NRC issued Regulatory Issue Summary (RIS) 2002-03, "Guidance on the Content of Measurement Uncertainty Recapture Power Uprate Applications,"

dated January 31, 2002 (ADAMS Accession No. ML013530183). RIS 2002-03 provides guidance to licensees on the scope and detail of the information that should be provided to the NRC staff for MUR power uprate LARs. While RIS 2002-03 does not constitute an NRC requirement, its use aids licensees in the preparation of their MUR power uprate LAR, while also providing guidance to the NRC staff for the conduct of its review. The licensee stated in its application dated February 7, 2013, that its LAR was submitted in accordance with the guidance of RIS 2002-03.

2.2 Implementation of an MUR Power Uprate at Fermi 2 In existing nuclear power plants, the neutron flux instrumentation continuously indicates the reactor core thermal power. This instrumentation must be periodically calibrated to accommodate the effects of fuel burnup, flux pattern changes, and instrumentation setpoint drift.

The reactor core thermal power generated by a nuclear power plant is determined by steam plant calorimetry, which is the process of performing a heat balance around the nuclear steam supply system (called a calorimetric). The accuracy of this calculation depends primarily upon the accuracy of. FW flow rate and FW net enthalpy measurements. As such, an accurate measurement of FW flow rate and temperature is necessary for an accurate calibration of the nuclear instrumentation. Of the two parameters; flow rate and temperature, the most important in terms of calibration sensitivity is the FW flow rate.

The instruments originally installed to measure FW flow rate in existing nuclear power plants were usually a venturi or a flow nozzle, each of which generates a differential pressure proportional to the FW velocity in the pipe. However, error in the determination of flow rate can be introduced due to venturi fouling and, to a lesser extent, flow nozzle fouling, the transmitter, and the analog-to-digital converter. As a result of the desire to reduce flow instrumentation uncertainty to enable operation of the plant at a higher power while remaining bounded by the

accident analyses, the industry assessed alternate flow rate measurement techniques and found that ultrasonic flow meters (UFMs) are a viable alternative. UFMs are based on computer-controlled electronic transducers that do not have differential pressure elements that are susceptible to fouling.

The licensee intends to use UFMs developed by the Cameron International Corporation (Cameron, formerly known as Caldon Ultrasonic Inc. (Caldon)), the leading edge flow meter (LEFM) CheckPius System, which provides a more accurate measurement of FW flow as compared to the accuracy of the venturi flow meter-based instrumentation originally installed at Fermi 2. The use of these UFMs to measure FW flow would allow the licensee to operate the plant with a reduced instrument uncertainty margin and an increased power level in comparison to its CLTP.

The Cameron LEFM CheckPius System was developed over a number of years. Cameron submitted a topical report in March of 1997, ER-80P, that describes the LEFM and includes calculations of power measurement uncertainty obtained using a Check system in a typical two-loop pressurized-water reactor or a two-FW-Iine boiling-water reactor. This topical report also provides guidance for determining plant-specific power calorimetric uncertainties. The NRC staff approved the use of this topical report in a safety evaluation (SE) dated March 8, 1999 (ADAMS Accession No. ML11353A017), which allowed a 1 percent power uprate. Following the publication of the changes to 10 CFR 50, Appendix K, which allowed for an uncertainty less than 2 percent, Cameron submitted topical report ER-160P (ADAMS Accession No. ML010510372), a supplement to ER-80P. The NRC staff approved ER-160P by letter dated January 19, 2001 (ADAMS Accession No. ML010260074), for use in a power uprate of up to 1.4 percent. Subsequently, in an SE dated December 20, 2001 (ADAMS Accession No. ML013540256), the NRC staff approved ER-157P, Rev. 5 (ADAMS Accession No. ML013440078), for use in a power uprate of up to 1. 7 percent using the CheckPius system.

The NRC staff also recently approved ER-157P, Rev. 8 and associated errata (ADAMS Accession Nos. ML081720323 and ML102950246). ER-157P, Rev. 8, corrects minor errors in Rev. 5, provides clarifying text, and incorporates revised analyses of coherent noise, non-fluid delays, and transducer replacement. It also adds two new appendices, Appendix C and Appendix D, which describe the assumptions and data that support the coherent noise and transducer replacement calculations, respectively.

As part of the implementation of this LAR, existing FW flow and temperature instrumentation will be retained and used for comparison monitoring of the LEFM system and as a backup FW flow measurement when needed.

Two Cameron LEFM CheckPius ultrasonic 8-path transit time flowmeters were installed at Fermi 2 in 2010 and commissioned iri 2011. As discussed above, the CheckPius design is described in Topical Reports ER-80P, ER-160P, and ER-157P, which have previously been approved by the NR.C staff for generic use. The LEFM CheckPius system will be used to develop a continuous calorimetric power calculation by providing FW mass flow and FW temperature input data to the plant computer system that is used for automated performance of the calorimetric power calculations.

The Check Plus system consists of one flow element {spool piece) installed in each of the two FW flow headers. The FW piping configurations are explicitly modeled as part of the CheckPius meter factor and accuracy assessment testing performed at Alden Research Laboratories (ARL). The installation location of each CheckPius conforms to the applicable requirements in Cameron's Installation and Commissioning Manual and Cameron topical reports ER.,.80P and

\

ER-157P. The bounding uncertainty analysis and the meter factor calculation and accuracy assessment are addressed in topical reports ER-781, Rev. 2, and ER-818, Rev. 0, respectively, which are included as proprietary enclosures to the LAR.

3.0 EVALUATION 3.1 Feedwater Flow Measurement Technique and Power Measurement Uncertainty 3.1.1 Regulatory Evaluation Early revisions of 10 CFR 50.46, and Appendix K to 10 CFR 50, required licensees to base their LOCA analysis on an assumed power level of at least 102 percent of the licensed thermal power level to account for power measurement uncertainty. The NRC later modified this requirement to permit licensees to justify a smaller margin for power measurement uncertainty.

Licensees may apply the reduced margin to operate the. plant at a level higher than the previously licensed power. The licensee proposed to use a Cameron LEFM CheckPius system to decrease the uncertainty in the measurement of FW flow, thereby decreasing the power level measurement uncertainty from 2.0 percent to 0.36 percent.

The licensee developed its license amendment request consistent with the guidelines in NRC RIS 2002-03.

3.1.2 Technical Evaluation 3.1.2.1 Licensee's Response to RIS 2002-03, Attachment 1,Section I In Attachment 1 to RIS 2002-03, the NRC staff issued "Guidance on the Content of Measurement Uncertainty Recapture Power Uprate [license amendment] Applications." This document provided guidance to licensees on one way to obtain NRC staff approval of their MUR LARs. In Section I of Attachment 1 to RIS 2002-03, the NRC staff provided guidance to licensees on how to address the issues of FW flow measurement technique and power measurement uncertainty in their MUR LARs. The following discusses the licensee's response to these guidelines in the LAR and the NRC staff's evaluation of these responses.Section I of to RIS 2002-03 contains eight items for the licensee to respond to and each of these is discussed in turn.

3.1.2.1.1 Items A. B, and C of Section I. Attachment 1 to RIS 2002-03 Item A requires the identification (by document title, number, and date) of the approved topical report on the FW flow measurement technique. Item B requires the licensee to reference the NRC's approval of the proposed FW flow measurement technique.

In response to Items A and B, the licensee identified, in Section 3.2.1 of Attachment 1 of the LAR, two proprietary Cameron Topical Reports: ER-80P, Revision 0, "Improving Thermal Power Accuracy and Plant Safety while Increasing Operating Power Level Using the LEFM Check System," March 1997, and ER-157P, Revision 8, "Supplement to Caldon Topical Report ER-80P: Basis for Power Uprates with an LEFM Check or an LEFM CheckPius System," May 2008. The licensee also referenced NRC safety evaluations "Staff Acceptance of Caldon Topical Report ER-80P: Improving Thermal Power Accuracy While Increasing Power Level Using the LEFM System," March 8, 1999 (ADAMS Legacy Library Accession No. 9903190065),

and "Final Safety Evaluation For Cameron Measurement Systems Engineering Report ER-157P, Revision 8," August 16, 2010 (ADAMS Accession No. ML102160663).

Item C requires a discussion of the plant-specific implementation of the guidelines in the topical report and the staff's letter/safety evaluation report (SER) approving the topical report for the FW flow measurement technique.

In response to Item C, the licensee provided a detailed description of the plant-specific implementation of FW flow measurement technique and the power increase gained as a result of implementing the technique throughout sections 3.1 and 3.2 of Attachment 1 of the LAR. In Section 3.1 of Attachment 1 of the LAR, the licensee stated that Fermi 2's application of the LEFM CheckPius System supports a 1.64 percent increase in rated thermal power (RTP) from current licensed thermal power (CL TP) of 3430 MWt to the proposed 3486 MWt. *The licensee also stated the LEFM CheckPius System was permanently installed in Fermi 2 during the fall 2010 refueling outage and was commissioned in June 2011, according to the requirements specified in Topical Reports ER-80P and ER-157P. The licensee provided a detailed description of the plant implementation in Section 3.2.2 of Attachment 1 of the LAR. Regarding the plant implementation, the licensee noted that in Loop B, the LEFM spool piece is installed 43 inches upstream of an elbow in the FW line. The licensee provided a justification for this exception in Enclosure 1 of the LAR, Section 3.2.4, "Disposition of NRC Criteria for Use of LEFM Topical Reports."

NRC Staff Conclusions on Items A, B. and C of Section I, Attachment 1 to RIS 2002-03 The NRC staff reviewed the licensee's response to items A, B, and C, and finds that the licensee has sufficiently addressed the plant-specific implementation of the Cameron LEFM CheckPius System using the proper guidelines from the applicable topical reports. The NRC staff also evaluated this information against the regulatory requirements of 1d CFR 50, Appendix K, and found it to be acceptable.

3.1.2.1.2 Item D of Section I. Attachment 1 to RIS 2002-03 The licensee's response to item D addresses the criteria established by the NRC staff in its approval of the FW flow measurement uncertainty technique used by the licensee in the LAR.

When the NRC staff approved ER-80P and ER-157P, Revision (Rev.) 8, in NRC staff SEs dated March 8, 1999 and August 16, 2010, respectively, it established nine criteria (four criteria from ER-80P and five criteria from ER-157P) that licensees were to address in order to implement these topical reports at their facilities. The licensee addressed these criteria in Section 3.2.4 of of its LAR. The NRC staff evaluated the licensee's apprqach to addressing each of these criteria, as discussed below.

Criterion 1 from ER-80P The licensee should discuss the maintenance and calibration procedures that will be implemented with the incorporation of the LEFM. These procedures should include processes and contingencies for an inoperable LEFM and the effect on thermal power measurement and plant operation.

Licensee Response In Enclosure 6 of the LAR, the licensee committed to revise plant maintenance and calibration

procedures to incorporate the vendor's maintenance and calibration requirements. Initial preventive maintenance scope and frequency will be based on vendor recommendations. The licensee further stated, in Section 3.2.4 of Attachment 1 of the LAR, incorporation of, and continued adherence to, these requirements will assure the LEFM system is properly maintained and calibrated.

LEFM and plant computer hardware and software configurations are maintained under the existing Fermi 2 configuration management system. The licensee discussed procedures addressing corrective actions, reporting deficiencies, and receiving and evaluating manufacturer's deficiency reports in Section 3.2.5 of Enclosure 1 of the LAR.

In Section 3.2.4 of Enclosure 1 of the LAR, the licensee described contingencies for LEFM inoperability. Technical Requirements Manual (TRM) actions will require channel checks every

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The LEFM will be considered non-operational for any single failure. The LEFM will also be considered non-operational for any self-diagnostic alarms. The TRM will permit remaining at uprate power for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months based on LEFM adjusted venturi-based flow signals while attempting to restore the LEFM to operable status. The contingency plans for plant operation with an inoperable LEFM are discussed later in this SE (Items G and H).

Based on its review of the licensee submittals, the NRC staff concludes the licensee adequately addressed Criterion 1 of Item D.

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

Licensee Response In Section 3.2.4 of Enclosure 1 of the LAR, the licensee stated:

The LEFM system was installed at Fermi 2 during the fall 2010 refueling outage and was commissioned in February 2011. The LEFM system is being used to supply the FW flow input to the plant process computer core thermal power calculation. Since the commissioning of the LEFM, the following maintenance issues have occurred:

In September 2011, the LEFM inputs to the plant computer system failed.

Feedwater flow inputs to the core thermal power calculation were transferred to the FW flow venturis. The LEFM [central processing unit (CPU)] was rebooted.

Post reboot data reports were analyzed by Cameron and they determined that the LEFM was functioning as designed and within the commissioning uncertainty bounds after the reboot.

In November 2011, during a plant down power, a "FW LEFM Meter B Status-Major Alert" alarm was received and the alarm cleared at approximately 23 percent of CL TP during the subsequent power increase. The cause of the alarm was determined to be a failed [resistance temperature detector (RTD)] in FW Meter B Plane 3. This failed component does not adversely affect the operation

of the LEFM and the RTD is currently scheduled to be replaced during the next refueling outage.

Also in November 2011, a small leak was discovered in the tubing for the FW Loop A pressure instrument. Feedwater flow inputs to the core thermal power calculqtion were transferred to the FW flow venturis, the pressure instrument was isolated, and the tubing leak was repaired.

In February 2012, a number of LEFM Meter B Plane 3 computer alarms were received. These alarms did not adversely affect the operation of the LEFM.

Troubleshooting was performed by Cameron and an adjustment was made to the "Kmax" factor for Meter B Plane 3. Post maintenance data reports were analyzed by Cameron and they determined that the LEFM was functioning as designed following the maintenance.

As mentioned above, final commissioning of the LEFM system at Fermi 2 was completed on February 24, 2011. The commissioning process verified bounding calibration test data, as described in Appendix F of [ER BOP, Rev. 0]. This step provided final confirmation that actual performance in the field meets the uncertainty bounds established for the instrumentation as described in [ER-7B1 (proprietary Enclosure 10 of the LAR)].

Based on its review of the licensee submittals, the NRC staff that concludes the licensee adequately addressed Criterion 2 of Item D.

Criterion 3 from ER-BOP The licensee should confirm the methodology used to calculate the uncertainty of the LEFM in comparison to the current FW instrumentation is based on 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 the venturi and the LEFM for comparison.

Licensee Response The licensee stated in Section 3.2.4 of Enclosure 1 of the LAR:

The LEFM system uncertainty calculation is based on the American Society of Mechanical Engineers (ASME) PTC 19.1-19B5 and the Instrumentation, Systems, and Automation Society (ISA) RP67.04.02-2000 methodologies, as described in Enclosure 10 of the LAR. This LEFM system uncertainty calculation methodology is based on a square-root-sum-of-squares (SRSS) calculation, as described in [Cameron Topical ReportER 157P, Revision B, "Supplement to Caldon Topical Report ER BOP: Basis for Power Uprates with an LEFM Check or an LEFM CheckPius System," May 200B (ADAMS Accession No. MLOB1720324)].

The Fermi 2 core thermal power uncertainty calculation for the LEFM FW flow instrumentation (Enclosure 13) was done in accordance with the Fermi 2 instrument setpoint methodology. The core thermal power uncertainty calculation for the existing F\f'! flow instrumentation was also done in accordance

with the Fermi 2 instrument setpoint methodology, and thus used consistent methodology.

Based on a review of the power uncertainty calculation (Enclosure 13 of the LAR), the NRC staff verified that the LEFM calculations are consistent with the Instrumentation, Systems, and Automation Society (ISA) RP67.04.02-2000 methodology. The licensee.used the NRC staff-approved GEH setpoint methodology for the previous venturi flow based calculation, which is also consistent with the ISA RP67.04.02-2000 methodology. The NRC staff concludes that the licensee adequately addressed Criterion 3 of Item D.

Criterion 4 from ER-80P For plant installations where the ultrasonic meter (including LEFM) was not installed with flow elements calibrated to a site-specific piping configuration (flow profiles and meter factors are not representative of the plant-specific installation), licensees should provide additional justification for its use. The justification should show the meter installation is either independent of the plant-specific flow profile for the stated accuracy, or the installation can be shown to be equivalent to known calibrations and plant configurations for the specific installation, including the propagation of flow profile effects at higher Reynolds numbers. Additionally, for previously installed calibrated elements, licensees should confirm the piping configuration remains bounding for the original LEFM installation and calibration assumptions.

Licensee Response The licensee stated in Section 3.2.4 of Enclosure 1 of the LAR:

Criterion 4 does not apply to Fermi 2. The calibration factors for the Fermi 2 spool pieces were established by tests of these spools at Alden Research Laboratory. These tests were performed on a full-scale model of the Fermi 2 hydraulic geometry. A discussion of the impact of the plant specific installation factors on the FW flow measurement uncertainty is provided in Cameron Report ER-781, Revision 2 (Enclosure 10 [of the LAR]) and Cameron Report ER-818, Revision 0 (Enclosure 11 (of the LAR]). The test configurations modeled the portion of piping upstream of the LEFM spool pieces. The tested configuration of the LEFM spool pieces can be compared to the plant installation drawings by comparing the drawings in ER-818, Figures 1 and 2, to the installation drawings*

in Enclosure 15 [of the LAR]. There is no significant difference between the Fermi 2 FW piping configuration and the model used at Alden Research Laboratory.

As was discussed above, the commissioning process for the Fermi 2 LEFM was completed by Cameron on February 24, 2011.

Based on the information presented above and the NRC staff review of the licensee's submitted calibration data in ER-781 and ER-818, the NRC staff concludes that the licensee adequately addressed Criterion 4 of Item D. The NRC staff notes the commissioning process is part of the bounding uncertainty verification and validation (V&V) activity.

In Section 3.2.4 of Enclosure 1 of the LAR, the licensee also addressed the criteria corresponding to the five items enumerated in the conclusion of the NRC staff SE of ER-157P, Revision 8. These criteria are discussed below:

Criterion 1 from ER-157P, Revision 8 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.

Licensee Response The licensee stated that the plant-specific justification for continued operation at the pre-failure level for a pre-determined 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 from ER-80P and in the licensee's response to Items G and H].

The NRC staff has reviewed the licensee's response and finds it a_cceptable.

Criterion 2 from ER-157P. Revision 8 A CheckPius 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 CheckPius at an increased uncertainty.

Licensee Response The licensee stated that Fermi 2 will not consider the CheckPius system with a single failure as a separate category. In these cases, the CheckPius LEFM system will be cqnsidered inoperable and the actions identified in the response to [Criterion 1 from ER-80P and in the licensee's response to Items G and H] will be implemented.

The NRC staff has reviewed the licensee's response and finds it acceptable.

Criterion 3 from ER-157P, Revision 8 An applicant with a comparable geometry can reference the findings in the NRC staff's SE for ER-157P Rev. 8 and Rev. 8 Errata (ADAMS Accession No. ML102160663) Section 3.2.1, to support a conclusion that downstream geometry does not have a significant influence on CheckPius calibration. However, CheckPius test results do not apply to a Check and downstream effects with use of a CheckPius with disabled components that make the CheckPius comparable to a Check must be addressed. An acceptable method is to conduct applicable Alden Research Laboratories (ARL) tests.

Licensee Response The licensee stated that the configuration of the LEFM spool pieces is described *in Section 3.2.2 of the LAR. The spool piece configuration for FW Loop B is comparable to that described in SeCtion 3.2.1 of the NRC staff's SE for ER-157P, Rev. 8. Based on tests at ARL and the NRC findings in Section 3.2.1 of theSE for ER-157P, Rev. 8, the downstream geometry does not have a significant influence on the Fermi 2 LEFM calibration. As is discussed in the response to Criterion 2 from ER-157P, Rev. 8, Fermi 2 will not consider the CheckPius system with a single failure as a separate category and the CheckPius LEFM system will be considered as inoperable.

The NRC staff has reviewed the licensee's response and finds it acceptable.

Criterion 4 from ER-157P. Revision 8 An applicant that requests an MUR with the upstream flow straightener configuration discussed in Section 3.2.2 of the NRC staff's SE for ER-157P, Rev. 8, should provide justification for claimed CheckPius uncertainty that extends the justification provided in Reference 17 of the NRC staff's SE. 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.

Licensee Response Fermi 2 does not have flow straighteners upstream of the LEFM spool piece installations. Thus, this criterion is not applicable to Fermi 2.

The NRC staff has reviewed the licensee's response and finds it acceptable .

.)

Criterion 5 from ER-157P. Revision 8 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 the NRC staff's SE for ER-157P, Rev.8.

Licensee Response The licensee stated that Fermi 2 conservatively assumes no moisture content in the Core Thermal Power Uncertainty Calculation. This is consistent with that described in Section 3.2.3 of the NRC staff's SE for ER-157P, Rev.8. Thus, this criterion is not applicable to Fermi 2.

The NRC staff has reviewed the licensee's response and finds it acceptable.

NRC Staff Conclusions on Item D of Section I. Attachment 1 to RIS 2002-03 Based on its review of the licensee's submittals as discussed above, the NRC staff finds the licensee has sufficiently addressed the plant-specific criteria in the SEs for Topical Reports ER-:-

80P and ER-157P, and therefore follows the guidance in Item D of Section I of Attachment 1 to RIS 2002-03 and meets the regulatory requirements of 10 CFR 50, Appendix K.

3.1.2.1.3 Item E of Section I. Attachment 1 to RIS 2002-03 Item E provides guidance to licensees concerning the submittal of a plant-specific total power measurement uncertainty calculation, explicitly identifying all parameters and their individual contributions to the power uncertainty.

To address Item E of RIS 2002-03, the licensee provided, as Enclosure 10 of the LAR, Cameron Engineering Report ER-781, Revision 2, "Bounding Uncertainty Analysis for Thermal Power Determination at Fermi Unit 2 Nuclear Generating Station Using the LEFM CheckPius System." In addition, the licensee listed each parameter's contribution and the values for the overall thermal power calorimetric uncertainty in Table 4.7-1 of Enclosure 13to the LAR, "Fermi 2 Calculation DC-6443, Volume I DCD 1, Revision A, 'Reactor Core Thermal Power Uncertainty with Feedwater Flow Measured by LEFM CheckPius C System."' The uncertainties

documented in this table are based on Cameron Engineering Report ER-781. The calculated overall thermal power calorimetric uncertainty with a fully operational plant computer and LEFM CheckPius C System is+/- 12.373 MWt which is+/- 0.361 percent of the current licensed thermal power or+/- 0.355 percent of the proposed uprate core thermal power.

The NRC staff reviewed these reports and determined the licensee properly identified all the parameters associated with the thermal power measurement uncertainty, provided individual measurement uncertainties, and calculated the overall thermal power uncertainty.

The licensee's fundamental approach used to determine the overall uncertainty is to statistically combine inputs. Channel statistical allowances are calculated for the instrument channels.

Dependent parameters are arithmetically combined to form statistically independent groups, which are then combined using the square root of the sum of the squares approach to determine the overall uncertainty. This methodology is consistent with the vendor's determination of the uncertainty of the Cameron LEFM CheckPius System, as described in the referenced topical reports, and is consistent with the guidelines in Regulatory Guide (RG) 1.1 05, Revision 3, "Setpoints for Safety-Related Instrumentation," issued December 1999 (ADAMS Accession No. ML993560062).

Therefore, the NRC staff concludes that the licensee has provided calculations of the total power measurement uncertainty at the plant, explicitly identifying all parameters and their individual contributions to the overall thermal power uncertainty. Therefore, the licensee has adequately addressed the guidance in Item E of Section I of Attachment 1 to RIS 2002-03 and has met the relevant regulatory requirements of 10 CFR 50, Appendix K.

3.1.2.1.4 Item F of Section I, Attachment 1 to RIS 2002-03:

Item F provides guidance to licensees in providing information to address the specified aspects of the calibration and maintenance procedures related to all instruments that affect the power

calorimetric.

In the Sections 3.2.4 and 3.2.5 of Enclosure 1 of the LAR, the licensee addressed each of the five aspects of the calibration and maintenance procedures listed in Item F of RIS 2002-03, as follows:

( 1) Maintaining Calibration The licensee stated that:

Implementation of the power uprate license amendment will include development of the necessary procedures and documents required for maintenance and calibration of the LEFM system. Plant maintenance and calibration procedures will be revised to incorporate Cameron's maintenance and calibration requirements prior to raising power above the Current Licensed Thermal Power (CL TP) of 3430 MWt. Initial preventive maintenance scope and frequency will be based on vendor recommendations (See Enclosure 6 [Summary of Regulatory Commitments], Item 3). The incorporation of, and continued adherence to, these requirements will assure the LEFM system is properly maintained and calibrated.

For instrumentation other than the LEFM system that c.ontributes to the power calorimetric 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.

(2) Controlling Hardware and Software Configuration The licensee stated that:

The LEFM system software and the plant process computer software configuration is maintained using existing Fermi 2 procedures, which include verification and validation of changes to software configuration. Configuration of the hardware associated with the LEFM system and the calorimetric process instrumentation is maintained in accordance with Fermi 2 configuration control procedures.

(3) Performing Corrective Actions The licensee stated, in part, that plant instrumentation affecting the power calorimetric, including the LEFM inputs, will be monitored by Fermi 2 personnel. Problems detected are documented per the Fermi 2 corrective action program, with necessary follow-up actions planned and implemented.

(4) Reporting Deficiencies to the Manufacturer The licensee stated that:

Problems with plant instrumentation identified by Fermi 2 personnel are also documented in the Fermi 2 corrective action program and necessary corrective actions are identified and. implemented. Deficiencies associated with the vendor's processes or equipment are reported to the vendor to support corrective action.

(5) Receiving and Addressing Manufacturer Deficiency Reports The licensee stated that:

Cameron has procedures to notify users of important LEFM deficiencies. Fermi 2 also has processes for addressing manufacturer's deficiency reports. Such deficiencies are documented in the Fermi 2 corrective action program.

The NRC staff's review concludes that the licensee addressed the calibration and maintenance aspects of the Cameron LEFM CheckPius System and all other instruments affecting the power calorimetric. Therefore, the NRC staff finds the licensee has met the guidance in Item F of Section I of Attachment 1 to Rl S 2002-03 and the relevant regulatory requirements of 10 CFR 50, Appendix K.

3.1.2.1.5 Items G and H of Section I, Attachment 1 to RIS 2002-03:

Items G and H provide guidance to licensees in proposing an allowed outage time (AOT) for the instrument and in proposing actions to reduce power if the AOT is exceeded.

The licensee proposed a 72-hour completion time for operation at any power level above the current licensed power of 3430 MWt with the LEFM inoperable, provided steady-state conditions persist throughout the 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . If the LEFM CheckPius System has not been declared

operable after the 72-hour completion time, the licensee will administratively control power at or below the pre-uprate power of 3430 MWt. The licensee's basis for the proposed 72-hour completion time is as follows:

Operations procedures and the Fermi 2 TRM will direct the use of the back-up calorimetric in the event of LEFM failure (Fail mode). This algorithm receives input from alternate plant instruments (FW venturis) for the FW flow rate calculation. The FW flow from the venturis will be calibrated to the last validated LEFM FW flow rate, so that the alternate calorimetric matches the primary LEFM based calorimetric.

A sudden de-fouling event during the 72-hour inoperability period 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 Fermi 2. Based on the drift value over the 18 month interval, the impact on thermal power measurement over the 72-hour period would be negligible.

LEFM repairs are expected to be completed within an 8-hour shift. A completion time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months provides plant personnel sufficient time to diagnose, plan* and package work orders, complete repairs, and verify normal system operation within original uncertainty bounds.

The 72-hour completion time begins when the annunciator alarm is received in the main control room. A control room alarm response procedure will be developed, providing guidance to the operators for initial alarm diagnosis. Methods to determine LEFM CheckPius System status and the cause of alarms are described in Cameron documentation. Cameron documentation will be used to develop specific procedures for operators and maintenance response actions.

A single path or plane malfunction (Maintenance mode) is not considered at Fermi 2. The LEFM will be considered inoperable following any single failure and the 72-hour completion time, as described above, will apply.

In the event that the plant process computer is inoperable, the LEFM would be considered non-operational and the proposed TRM actions will be applied, as described above. A procedure currently exists for reactor engineering personnel to manually calculate core thermal power. In addition, the licensee has stated that operators routinely monitor other indications of core thermal power, including Average Power Range Monitors (APRMs), steam flow, FW flow, turbine first stage pressure, and main generator output.

The licensee will operate Fermi 2 at the uprate power following a loss ofthe plant computer during the 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months completion time under the same TRM action statement for loss of the LEFM.

DTE provided a calculation in Enclosure 13 of the LAR for manual calculation of core thermal power (CTP). The NRC staff reviewed the calculation and determined the uncertainty associated with performing the CTP calculation manually is nearly identical to the uncertainty when CTP is calculated using the plant computer and supports remaining at the uprate power during the 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months completion time.

Based on the above discussion and the NRC staff review of the licensee's LAR and Cameron engineering reports, the NRC staff concludes that the licensee provided sufficient justifications for the proposed completion time and the proposed power reduction actions if the completion

time is exceeded. Therefore, the licensee has followed the guidance in Items G and H of Section I of Attachment 1 to RIS 2002-03 and has niet the regulatory requirements of 10 CFR 50, Appendix K.

3.1.2.2 General Acceptance Criteria for UFMs General acceptance criteria apply to all aspects of testing in a certified facility, transfer from the test facility, initial operation, and long-term in-plant operation. These criteria are:

Traceability to a recognized national standard. This requires no breaks in the chain of comparisons, all chain links must be addressed, and there can be no unverified assumptions.

Calibration.

Acceptable addressing of uncertainty beginning with an initial estimate of the bounding uncertainty and continuing through all aspects of initial calibration in a certified test facility, transfer to the plant, initial operation, and long-term operation.

For CheckPius, meeting these criteria includes documenting:

Design and characteristics information,

Calibration testing at a certified test facility,

Any potential changes associated with differences between testing and plant operation including certification that initial operation in the plant is consistent with pre-plant characteristics predictions, and

In-plant operation.

3.1.2.3 Initial Design and Characteristics To determine volumetric flow rate, the Caldon UFM transmits an acoustic pulse along a selected path and records the arrival of the pulse at the receiver. Another pulse is transmitted in the opposite direction and the time for that pulse is recorded. Since the speed of an acoustic pulse will increase in the direction of flow and will decrease when transmitted against the flow, the difference in the upstream and downstream transit times for the acoustic pulse provides information on flow velocity. Once the difference in travel times is determined, the average velocity of the fluid along the acoustic path can be determined. Therefore, the difference in transit time is proportional to the average velocity of the fluid along the acoustic path.

The CheckPius provides an array of 16 ultrasonic transducers installed in a spool piece to determine average velocity in 8 paths. The transducers are arranged in fixtures such that they form parallel and precisely defined acoustic paths. The chordal placement is intended to provide an accurate numerical integration of the axial flow velocity along the chordal paths.

Using Gaussian quadrature integration, the velocities measured along the acoustic paths are combined to determine the average volumetric flow rate through the flow meter cross section.

Note that this process assumes a continuous velocity profile in the flow area perpendicular to the spool piece axis. Although the velocity profile can be distorted, the distortion cannot be

such that the Gaussian quadrature process no longer provides an acceptable mathematical fit to the profile, such as may occur if the profile is distorted in a way that is not recognized by the CheckPius due to an upstream flow straightener.

To obtain the actual average flow velocity, a calibration factor is applied to the integrated average flow velocity indicated by the UFM. The calibration factor for the Caldon UFMs is determined through meter testing at ARL and is equal to the true area averaged flow velocity divided by the flow velocity determined from the average meter paths to correlate the meter readings to the average velocity and hence to the average meter volumetric flow. The mass flow rate is found by multiplying the spool flow area by the average flow velocity and density.

The mean fluid density is obtained using the measured pressure and the derived mean fluid temperature as an input to a table of thermodynamic properties of water. Typically, the difference between an uncalibrated CheckPius and ARL test results is less than 0.5 percent.

This close agreement means that obtaining a correction factor for a CheckPius is relatively insensitive to error for operation under test conditions. Further, as discussed in this safety evaluation, correction factor is not a strong function of the difference between test and plant conditions and the same conclusion applies.

Use of a spool piece and chordal paths improves the dimensional uncertainties including the time measurement of the ultrasonic signal and enables the placement of the chordal paths at precise locations generally not possible with an externally mounted UFM. This allows a chordal UFM to integrate along off-diameter paths to more efficiently sample the flow cross section. In addition, a spool piece has the benefit that it can be directly calibrated in a flow facility, improving measurement uncertainty compared to externally mounted UFMs that were historically installed in nuclear power plant FW lines.

The licensee acceptably addressed the above aspects of its proposed use of CheckPius UFMs, Flow straighteners are installed a significant distance downstream of the CheckPius installations and other potential distortions of the flow profile are either absent or acceptably addressed in ARL testing. Coverage of other aspects of the proposed use is addressed in other sections of this SE.

3.1.2.4 Test Facility Considerations Test facility considerations include test facility qualification, as well as test fidelity and range.

Test Facility Qualification Calibration testing at a qualified test facility and at a nuclear power plant involves ensuring traceability to a national standard, understanding facility uncertainty, and facility operation. In the LAR, the licensee used Cameron reports that reference the work of ARL to provide traceability to National Institute of Standards and Technology (NIST) standards. The testing at ARL (ADAMS Accession No. ML072710557) was audited by the NRC staff in 2006 (ADAMS Accession No. ML060400418) and the staff verified the traceability to NIST standards. The NRC staff's audit found th~t ARL's processes and operation were consistent with the claimed facility uncertainties. The staff also observed testing during a visit to ARL in 2009 (meeting notes at ADAMS Accession No. ML092680921 and ML092680922), and observed some improvements in test facility hardware. The NRC staff judged these changes would not change its previous conclusions regarding test operations and results. In ER-818, Rev. 0, Cameron restated that "all elements of the lab measurements ... are traceable to NIST standards."

Consequently, the staff finds that the documentation provides an acceptable basis for concluding that ARL meets the stated testing criteria.

Historically, all CheckPius installations have been calibrated at ARL. The NRC staff audit confirmed that ARL was providing acceptable test data for the configurations under test.

Consequently, the NRC staff finds that the qualification of ARL with respect to CheckPius is a~ceptable without further investigation or confirmation, provided test conditions remain consistent with the referenced conditions.

Test Fidelity and Test Range Test fidelity, such as test versus planned plant configuration, test variations to address configuration differences, and potential effects of operation on flow profile and calibration, should be addressed on a plant-specific basis. Licensee requests must provide a comparison of the test and plant piping configurations with an evaluation of the effect of any differences that could affect the UFM calibration. Further, sufficient variations in test configurations must be tested to establish that test-to-plant differences have been bracketed in the determination of UFM calibration and uncertainty. Historically, calibration testing has acceptably covered upstream effects by applying a variation of configurations to distort the flow profile. Further, if the spool piece may be rotated during plant installation from the nominal test rotation,' the effect

of rotation should be addressed during testing.

Plant piping configuration drawings must, at a minimum, include isometrics with dimensional information that describe piping, valves, FW flow meters, and any other components from the FW pumps to at least 10 pipe diameters downstream of the FW flow meter that is most distant from the FW pump. Preferable are scale three dimensional (30) drawings in place of isometrics that show this information. Test information must include 30 drawings of the test configuration including dimensions.

Topical report ER-818 provided the test configurations and the licensee's LAR provided the in-plant CheckPius locations. As discussed below, distances between the exit of the CheckPius spool pieces and the downstream flow straighteners and venturis are sufficient such that there will be no effect on the LEFM calibration. Test dimensions. and configurations upstream of the LEFM were acceptable when compared to the plant installations. In addition, tests with offset orifices provided flow distortions that are judged to significantly bound any flow behavior that differs between the tests and plant.

Weigh tank tests were run at different flow rates for each simulated FW loop. Tests included a variation of flow rates through the CheckPius and included an eccentric orifice upstream in the FW pipes containing the CheckPius. Most test results were included in the reported main FW

calculation.

3.1.2.5 In-Plant Installation and Operation of LEFMs In the LAR, the licensee addressed in-plant installation and operation of the CheckPius LEFMs.

3.1.2.5.1 Transfer from Test to Plant and In-Plant Installation Each licensee requesting an MUR power uprate must conduct an ih-depth evaluation of the UFM following installation at its plant that includes consideration of any differences between the test and in-plant results and must prepare a report that describes the results of the evaluation.

This should address such items as calibration traceability, potential loss of calibration, cross-checks with other plant parameters during operation to ensure consistency between thermal power calculation based upon the LEFM and other plant parameters, and final commissioning testing. The process should be described in written documentation and a final commissioning test report should be available for NRC inspection. In its LAR, the licensee stated that commissioning tests will be performed.

Historically, the Check and CheckPius UFMs are the only UFMs to have acceptably demonstrated UFM calibration traceability from the test facility to U.S. nuclear power plants.

This traceability is possible due to the ability to provide the flow distribution I velocity profiles as a function of radius and angular position in the spool piece, the small calibration correction necessary to fit test data to UFM indication, and the demonstrated insensitivity to changes in operation associated with transfer changes and plant changes. Although other means have been used to obtain flow rate, such as use of tracers in the FW, they have not attained the small uncertainty demonstrated by the CheckPius LEFM.

Experience to date is that an UFM must provide flow profile information and calibration traceability when extrapolating from test flow rates and temperature conditions to plant conditions. Transfer uncertainty is associated with any changes in mechanical. and operating ..

conditions in the plant due to any installations or other modifications. Changes in mechanical conditions include mechanical perturbations due to such things as transducer installation, mechanical misalignment, and fidelity between the test and plant. Changes in operating conditions involve consideration of potential effects such as noise due to pumps and valves, and changes in flow profile, including swirl, flow rate, and temperature.

As discussed above, the test facility configuration and test parameters are expected to provide a basis for providing fidelity between the test and plant. However, an exact correspondence is n9t possible. Potential differences must be addressed during implementation of the UFM and licensees are expected to have the ability to both identify differences and address them during operation.

The licensee addressed uncertainties introduced by the installation of the LEFM at Fermi 2 in ER~781. As stated above, the facility uncertainty is acceptable. Topical report ER-781 is referenced for transducer installation uncertainty. The content is essentially identical to Appendix D of ER-157P-A, Rev. 8, which was approved by the NRC staff on August 16, 2010.

Consequently, the NRC staff finds that the licensee's treatment of transducer installation uncertainty is acceptable. The licensee showed that LEFM commissioning will include verification of ultrasonic signal quality and evaluation of actual plant hydraulic flow profiles as compared to those documented during the ARL testing. The NRC staff has evaluated the licensee's approach to the commissioning test and finds it acceptable.

3.1.2.5.2 In-Plant Operation Many of the calibration aspects associated with transfer from a test facility to the plant apply during operation as valve positions change, different pumps are operated, and physical changes occur in the plant. The latter include such items as temperature changes, preheater alignment and characteristics changes, pipe erosion, pump wear, crud buildup and loss, and valve wear.

Further, potential UFM changes, such as transducer degradation or failure, may also occur and the UFM should be capable of responding to such behavior. Either the UFM must remain within calibration and traceability must continue to exist during such changes, or the UFM must clearly identify that calibration and traceability are no longer within acceptable parameters. Past

experience has shown that the CheckPius is capable of handling these operational aspects .

. Further, UFM operation should be cross-checked with other plant parameters that are related to FW flow rate. Should such checking identify abnormal behavior, it should be identified to the NRC, the validity of the final commissioning test report should be confirmed, and the final commissioning test report should be updated as necessary to reflect the new information.

Further, the UFM must be considered inoperable if its calibration is no longer established to be within acceptable limits.

Section 3 of Enclosure 1 of the LAR describes the training calibration, maintenance, procedures, corrective action program, and procedures to ensure compliance with the requirements of 10 CFR 50 Appendix B. The NRC staff has evaluated these aspects in the licensee's application and finds thatthe licensee's approach t_o in-plant operation is acceptable.

Operation with a Failed Component A brief description should be provided that covers system self-testing features, channel checks, control room alarms, and plant process computer functions. The following should be addressed to cover conditions when th~ CheckPius system becomes degraded or inoperable:

( 1) Operator response.

(2) Changes in FW flow input to the core thermal power calculation.

(3) Allowed Outage Time (AOT) Time when continued operation at full power is permitted and time when power must be reduced, including specification of the reduced power

level.

(4) Justification for the AOT with respect to such topics as calibration of FW venturis, venturi fouling or defouling, monitoring of other indications of core thermal power such as average power range monitors, steam flow rate, feed flow rate, turbine first stage pressure, and main generator output.

(5) Response if the plant computer system is not operable.

ER-157P, Rev. 8 states that "The redundancy inherent in the two measurement planes of an LEFM CheckPius also makes this system more resistant to component failures" when compared to the Check. "For any single component failure, continued operation at a power greater than that prior to the uprate can be justified with a CheckPius system ... since the system with the failure is no less than an LEFM Check." This is acceptable subject to two qualifications:

(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 quoted statement is that the CheckPius has 16 transducer housings interfacing with the flowing water whereas the LEFM Check has 8. Consequently, a CheckPius 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 as stated.

An acceptable quantification method is to establish the effect in an acceptable test configuration such as can be accomplished at ARL.

If the LEFM system or a portion of the system becomes non-operational, Fermi stated that control room operators will be alerted by a control room alarm. FW flow input to the core thermal power calculation would then be provided by the existing FW flow venturis.

The licensee continued by stating that, since the FW flow venturis will be corrected to the last validated data from the LEFM system, it is acceptable. to remain at the uprated RTP of 3486 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. As noted in the TRM changes provided, if the LEFM system is not repaired within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , power will be reduced and administratively controlled to remain less than or equal to the CLTP of 3430 MWt.

The 72-hour completion time for the LEFM system prior to reducing to the CLTP is stated to be acceptable since, during the 72-hour completion time, the existing FW flow venturi-based signals will be corrected to the last validated data from the LEFM system. Although the FW flow venturi measurements may drift slightly during this period due to fouling, fouling of the FW flow venturis results in a higher than actual indication of FW flow. This condition results in an overestimation of the calculated calorimetric power level, which is conservative, as the reactor will actually be operating below the calculated p~wer level.

A sudden de-foulir:lg event during the 72-hour inoperability period 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 Fermi 2.

Regarding potential drift in the measurement of FW aifferential pressure across the FW flow venturis, industry experience for similar BWRs shows that the instrument drift associated with FW flow measurements is insignificant over a 72-hour time period. Differences in the FW loop flow rates measured by the FW flow vent uris and the LEFM were compared for the time period since LEFM commissioning. Evaluation of the data indicates a maximum change in the difference between the FW flow venturi and the LEFM measured flow rates of approximately 0.3 percent over an 18 month operating cycle or less than 0.002 percent over a 72-hour period.

Thus, the effects of instrument drift and/or *fouling of the. FW floW venturis would be insignificant during the time period the LEFM is allowed to be out of service.

If the core power level is below the CLTP at the time the LEFM is declared non-operational or if the power level drops below .the current CLTP during the completion time, power may not be .

raised above 3430 MWt prior to the LEFM returning to operational status. In Section 3.0 of the Fermi 2 TRM, Technical Requirements Limiting Condition for Operation (TRLCO) 3.0.4 prohibits entering a condition specified in the applicability when a TRLCO is not met, except when either (a) the associated actions permit operation in that condition for an unlimited period of time, or (b) upon acceptable performance of a risk assessment and establishment of appropriate risk management actions. Exception (a) cannot be used for the LEFM, since the applicability for proposed LEFM TRLCO applies to core thermal power levels greater than 3430 MWt and the TRM actions only permit operation above 3430 MWt for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months with a nonoperational LEFM. Regarding exception (b), the proposed Fermi 2 TRM section includes a note stating that TRLCO 3.0.4.b does not apply to the LEFM. Thus, the application of TRLCO 3.0.4 to the proposed TRM section for the LEFM would prohibit raising power above 3430 MWt without the LEFM being operational.

In the event that the plant process computer is inoperable, the LEFM would be considered nonoperational and the proposed TRM actions would be applied. A procedure currently exists

for reactor engineering personnel to manually calculate core thermal power. In addition, operators routinely monitor other indications of core thermal power, including Average Power Range Monitors (APRMs), steam flow, FW flow, turbine first stage pressure, and main generator output.

The NRC staff agrees that under normal operating conditions the drift of a venturi over a 72-hour period would be minimal. The reference above as well as previous precedent of plants with similar operating experience and approved outage time provides reasonable assurance that the plant will operate safely for the 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months outage time and maintain the licensed power level.

Based on the considerations discussed above, the NRC staff finds that the licensee's planned operation with a failed CheckPius LEFM is acceptable.

Spool Piece Dimensional Effects on UFM Response Topical reports ER-80P and ER-157P-A, Rev. 8 address the effect of variation in such spool piece dimensions as as-built internal diameter and sonic path lengths, path angles, and path spacings. The NRC staff has reviewed the licensee's approach for addressing these effects and finds it acceptable.

Transducer Installation Sensitivity Transduce*rs may be removed after ARL testing to avoid damage during shipping the spool piece to the plant. Further, transducers may be replaced following failure or deterioration during operation. Replacement potentially introduces a change in position within the transducer housing that could affect the chordal acoustic path. Appendix D of ER-157P addresses replacement sensitivity by describing tests performed at the Caldon Ultrasonics flow loop and provides a comparison of test results to analyses of potential placement variations that shows that the test results are bounded by predicted behavior. One would expect an uncertainty associated with the test loop even if nothing was changed. This is not addressed in the ER-157P information. Rather, all of the test uncertainty is conservatively assumed to be due to transducer replacement. Further, the analyses predict a larger uncertainty than that obtained during testing, and.the analysis uncertainty is used for transducer replacement uncertainty.

The NRC staff has evaluated this approach and has judged it to be sufficiently conservative to cover the inability of the test loop to achieve flow rates comparable to those obtained in plant installations and to cover any analysis uncertainty associated with applications with pipe diameters that differ from the tests. Consequently, the NRC staff finds that transducer replacement has been acceptably addressed and the ER-157P, Revision 8, process for determining transducer replacement uncertainty is acceptable.

The Effects of Random and Coherent Noise of LEFM CheckPius Systems Appendix C of ER-157P provides a proprietary methodology for test- and plant-specific calculation of the contribution of noise to CheckPius uncertainty. Review of this methodology has established that licensees may use this methodology in their MUR requests.

The LAR and topical reports ER-781 and ER-818 show that critical performance parameters, including signal-to-noise ratio, are continually monitored for every individual meter path. Alarm setpoints are established to ensure corresponding assumptions in the uncertainty analysis

remain bounding. Signal noise will be minimized via strict adherence with Cameron design requirements. LEFM commissioning included verification of ultrasonic signal quality.

In ER-818, the licensee reported test signal-to-noise ratios for random and coherent noise that were within specifications and that the uncertainty attributable to the electronics and signal to noise ratio are included in the overall meter factor uncertainty.

The NRC staff has evaluated the test results and ER-781 and ER-818. The NRC finds that the licensee's approach for noise is sufficient to ensure that this topic is acceptably addressed.

Evaluation of the Effect of Downstream Piping Configurations on Calibration The turbulent flow regimes that exist when the plant is near full power result in limited upstream flow profile perturbation from downstream piping. Consequently, the effects of downstream equipment need not be considered for normal CheckPius operation provided changes in downstream piping, such as the entrance to an elbow, are located greater than two pipe diameters downstream of the chordal paths. However, if the CheckPius is operated with one or more transducers out of service, the acceptable separation distance is likely a function of transducer to elbow orientation. In such cases, if separation distance is less than five pipe diameters, it should be addressed. ~

As discussed in Section 3.1.2.4, above, separation from downstream components is needed so that CheckPius operation will not be affected. The in-plant separation from downstream piping components such as elbows and venturis is acceptable and will not affect CheckPius operation.

The NRC staff has reviewed the licensee's approach to evaluation of the effect of downstream piping configurations on calibration and finds it to be acceptable.

Evaluation of Upstream Flow. Straighteners on CheckPius Calibration Operation with an upstream flow straightener is known to affect CheckPius calibration to a greater extent than most other upstream ha*rdware. If a licensee proposes this configuration, it must*provide justification.

A previously undocumented effect of upstream tubular flow straighteners on CheckPius calibration was discovered during ARL testing while NRC staff members were at the site on August 24, 2009, that did not appear to apply to any previous CheckPius installations. As follow-up, additional tests were conducted with several flow straighteners and two different pipe/spool piece diameters to enhance the statistical data basis and to develop an understanding of the interaction between flow straighteners and the CheckPius. The results are provided in the proprietary topical report ER-790, Rev. 1, "An Evaluation of the Impact of 55 Tube Permutit Flow Cqnditioners on the Meter factor of an LEFM CheckPius," dated March 2010.

Cameron concluded that two additional meter factor uncertainty elements are necessary if a CheckPius is installed downstream of a tubular flow straightener and provided uncertainty values derived from the test results. The data also provide insights into the unique flow profile characteristics downstream of tubular flow straighteners and a qualitative understanding of why the flow profile perturbations may affect the CheckPius calibre3tion.

Cameron determined that the two uncertainty elements are uncorrelated and therefore

combined them as the root sum squared to provide a quantitative uncertainty. The Cameron approach is judged to be valid, but there is concern that the characteristics of existing tubular flow straighteners in power plants may not be adequately represented by samples tested in the laboratory. Any licensee that requests an MUR with the configuration discussed in this section of this SE, should provide justification for claimed CheckPius uncertainty that extends the justification provided in ER-790.

No flow straighteners are installed upstream of the LEFM locations in the FW lines at Fermi 2.

In addition, the flow straighteners are a significant distance downstream. Therefore, the NRC staff concludes that flow straightener effects are not a concern in this application.

3.1.2.6 Other Thermal Power Calculation Considerations Steam Moisture Content Some modern separators and dryers deliver steam with moisture content in the 0.05 percent range and these licensees often assume a zero moisture content that is conservative since the calculated power will be greater than actual power for such cases. No uncertainty is necessary if no moisture is assumed.

ER-160P discusses an analysis in which the uncertainty in thermal power due to measurement of all variables excluding moisture is assumed to be normally distributed with two standard deviations of 0.3357 percent, essentially the aggregate uncertainty of all contributors excluding moisture for the CheckPius system. The contribution of uncertainty due to moisture content was then calculated by multiplying a second, uniformly distributed random number times the uncertainty band assumed in Table A-1 of ER-160P and Monte Carlo calculations of total power uncertainty were obtained. The results are summarized in Figure 1 of ER-160P. The author "concluded that applicants 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." This was stated to be an acceptable approach in the documentation supporting publication of ER-157P-A, Rev .. 8 (ADAMS Accession No. ML102950245).

Deficiencies and Corrective Actions The licensee identified its process for addressing Cameron deficiency reports as well as reporting deficiencies to the manUfacturer. In each case, the licensee will use its corrective action program. In the case of receiving deficiency reports, the licensee will document and address applicable deficiencies in its corrective action program, as well.

Reactor Power Monitoring*

Licensees should identify guidance to ensure that reactor thermal power licensing requirements are not exceeded. Proposed guidance was addressed in the NRC's "Safety Evaluation Regarding Endorsement of [Nuclear Energy Institute] Guidance for Adhering to the License Thermal Power Limit," dated October 8, 2008 (ADAMS Accession No. ML082690105).

3.1.2. 7 Uncertainty The NRC staff reviewed Cameron's evaluation of the flow rate uncertainty associated with the test facility, measurement (including transducer installation), extrapolation from test

conditions to plant operating conditions, modeling, and data scatter, and finds it to be acceptable. The staff's evaluation of the power measurement uncertainty calculations are provided in Section 3.11 of this SE.

Test Facility Uncertainty The budgeted test facility uncertainty is consistent with past NRC staff evaluations and the value provided by ARL that was previously evaluated by the staff during the 2006 audit (ADAMS Accession No. ML060400418). Based on these considerations, the NRC staff finds this to be acceptable.

Measurement Uncertainty The licensee addresses uncertainty due to such contributors as thermal expansion; dimensions; temperature, pressure, and density determination; and transducer installation. The contribution of some of these contributors was discussed in the above report sections. Overall, measurement uncertainty is acceptably addressed.

Extrapolation Uncertainty Although calibration tests were performed, they were conducted at room temperature. This resulted in Reynolds numbers about a factor of ten less than would occur in the plant and an extrapolation is necessary to obtain the in-plant calibration factor. A positive aspect of the CheckPius is that the calibration factor is close to one and small errors in the extrapolation do not significantly affect extrapolation accuracy. Another aspect is that the Check and CheckPius characteristics permit an alternate extrapolation approach to the Reynolds number extrapolation. This involves the flatness ratio *(FR), which for the CheckPius is defined as the ratio of the average axial velocity at the outside chords (chords 1, 4, 5, and 8) to the average axial velocity at the inside chords (chords 2, 3, 6, and 7):

where FR is a function of Reynolds number, pipe wall roughness, and the piping system configuration.

The effect of the configuration is evaluated in laboratory tests. The effect of Reynolds number is deduced from a flow profile correlation. The advantage of this approach is that a plot of FR versus calibration factor is linear and the calibration factor is insensitive to variation in FR.

These results are consistent with analytic predictions and have been confirmed via ARL tests of many plant configurations. Further, minor changes in calibration factor observed in different hydraulics configurations are predictable and can be confirmed analytically. Therefore, if plant conditions result in a change in FR, the calibration factor may be adjusted to reflect the change in FR. This process is discussed further below.

Cameron also uses swirl rate.defined as:

v - v v - v. v - v. v - v ]

Swirl Rate= Average [ 5 ,~,-- , 3 2 6 2- Ys 2- Ys 2- YL 2- YL where ys and YL are normalized chord locations for outside/short and inside/long paths.

Cameron also uses swirl rate to characterize behavior obtained during ARL tests.

The licensee provided experimental data of Meter Factor (MF) as a function of other parameters for each of the CheckPius instruments in ER-818. MF variation over the range of test flow rates was typically shown to vary by less than 0.001.

Cameron includes an uncertainty term for extrapolation from laboratory conditions to plant conditions that is computed from empirical equations to account for change in Reynolds number and other effects such as a difference in pipe wall roughness. The calibration factor is shown to change in the fifth significant figure over a factor of ten change in Reynolds number between the test and plant conditions. With respect to extrapolation uncertainty, some of the uncertainty was likely already addressed by parametric testing over Reynolds numbers and FRs.

Modeling Uncertainty Cameron uses FR and swirl rate to characterize the velocity distribution and to validate the experimentally determined calibration factor when installed in a plant. Cameron has discussed application of calibration data obtained at ARL for 330 hydraulic configurations with 75

CheckPius UFMs with an average calibration factor of 1.002 with a standard deviation of

+/- 0.0039. .

Cameron discussed its experience in calibrating over 100 UFMs with close to 500 different test configurations since typically 4 or 5 configurations were tested for each UFM. An approach is discussed where one configuration subset was considered applicable to the applicant's installation and modeling sensitivity was computed using that information.

The licensee's method for determining modeling uncertainty is acceptable.

Data Scatter Uncertainty The precision with which the calibration factor is determined includes calibration data for each CheckPius and 95 percent confidence limits are calculated for test configurations that resemble the in-plant configuration.* The licensee's determination of data scatter uncertainty is acceptable.

3.1.2.8 Evaluation of Measurement Uncertainty Recapture Power Uprate Request Effect of Flatness Ratio (FR) change on Meter Factor (MF)

Flatness ratio, as discussed above, is defined as:

FR = v1 +V4+Vs+Va = Vs V2 +V3 +V6 +V7 VL where V1 , V4, Vs, and V8 are velocities measured along the outside chords (the short paths), V2 ,

V3 , V6 , and V7 are velocities measured along the inside chords (the long paths), Vs is the mean short part velocity, and VL is the mean long path velocity. The paths are illustrated by the

. horizontal lines in the following figure that correspond to the paths between the CheckPius transducers 1 :

1 Measurements are at an angle with respect to pipe length. Velocities are translated into this configuration for calculation purposes.

FR can be determined experimentally, such as by testing at ARL, where the CheckPius will provide the velocity data.

Once the V's are determined, the flow rate determined by the CheckPius can be calculated by multiplying the rectangular vertical widths (weighting factors) indicated in the following figure by the dash lines by the corresponding velocities times two:

SHORT PATH LONG PATH I

I

'~~-~-~----------- ~**~********-~-*~-LI I

I I I I I I I

I

...... - - __ I Once the CheckPius flow rate has been calculated, MF can be determined by comparing the CheckPius flow rate to the experimentally determined data.

FR and MF can also be calculated using an assumed symmetric velocity distribution that is a function of pipe radius, expressed as V(r), where r is the reduced radial position with the origin.

at the pipe centerline and 0 :s r :s 1. Since the CheckPius determines a mean velocity along the path, the calculation must be based on the same path, as illustrated by the "x" dimension in the following figure:

\

where the mean velocity is calculated by 1/X f0x V(x)dx where x= X at r =Rand V(x) is determined from the assumed V(r) where the relationship between x and r is obtained from the

geometry illustrated in the figure.

1 The calculations define MF as the flow rate calculated by 2rr f0 V(r) r dr divided by the calculated LEFM flow rate obtained by two times f V(x) dx over the short arid long path lengths multiplied by the corresponding weighting factors. The calculations result in a linear relationship between MF and FR with little variation in MF. They further allow extrapolation of MF to high Reynolds numbers in the plant that cannot be reached in the ARL tests by offsetting the calculated curve to pass through the data which, as shown in ER-818, provide a good fit to the offset curve.

Installation Considerations The transducers are located in the turbine building in a radiation field. Based on a radiation survey in the fall of 2011, at full power at the proposed installation locations, the licensee concluded that there would be negligible degradation in transducer performance.

3.

1.3 NRC Staff Conclusion

s Regarding Power Measurement Uncertainty The NRC staff reviewed the reactor systems and thermal-hydraulic aspects of the proposed LAR in support of implementation of an MUR power uprate. Based on the considerations discussed above, the NRC staff determined that the results of the licensee's analyses related to these areas continue to meet applicable acceptance criteria following implementation of the MUR.

The NRC staff has reviewed the licensee's response to RIS 2002-03, Attachment 1,Section I, and finds that the licensee has met the guidelines contained therein. The NRC staff finds that the licensee has adequately addressed the issues of FW flow measurement technique and power measurement uncertainty in its MUR LARs. The licensee has also adequately addressed general acceptance criteria for UFMs, adequately described the UFM design and characteristics, adequately addressed the test facility considerations, and adequately addressed issues with in-plant installation and operation of LEFMs.

3.2 Human Factors 3.2.1 Regulatory Evaluation The NRC human performance review* addresses programs, procedures, training, and plant design features related to operator performance during normal and accident conditions. The NRC's human performance evaluation was conducted to confirm that operator performance would not be adversely affected as a result of changes made to implement the proposed MUR power uprate. The scope of the review included the licensee-identified changes needed for the proposed MUR power uprate, including operator actions, human-system interfaces, procedures, and training.

3.2.2 Technical Evaluation The NRC staff has developed a standard set of questions for human performance reviews of MURs (Regulatory Information Summary (RIS) 2002-03, Attachment 1,Section VII, Items 1 through 4). The following evaluations ad9ress the licensee's response to those questions in its February 7 and April 5, 2013, submittals.

Operator Actions The licensee stated in its submittals that an evaluation was performed of the operator actions sensitive to power uprate and no substantive impacts were identified. The only* time-dependent operator actions identified by the licensee that are impacted by the MUR power uprate are those associated with post-fire plant shutdown from outside the main control room. The licensee determined that the increase in time required for. plant shutdown from outside the main control room was: (1) less than the time available as determined per verification and validation; and (2) was also less than the time available as stated in the Updated Final Safety Analysis Report (UFSAR).

The NRC staff has reviewed the licensee's statements in both the original submittal and its supplemental information relating to any impacts of the MUR power uprate on existing or new operator actions. The staff finds that the statements provided by the licensee are in conformance with Section Vll.1 of Attachment 1 to RIS 2002-03. The NRC staff concludes that the proposed MUR power uprate will not adversely impact the licensee-identified operator actions or their response times.

Emergency and Abnormal Operating Procedures The licensee stated in its submittal that the current emergency operating procedures (EOPs) and abnormal operating procedures (AOPs), were reviewed for potential changes due to the proposed MUR power uprate. The licensee concluded that no changes to operator actions in either the EOPs or AOPs were needed, and that the available times to perform operator actions would remain unchanged. The revisions to the EOPs and AOPs will be made to reflect the higher power level and minor setpoint changes, and will be made prior to MUR implementation.

The NRC staff has reviewed the licensee's evaluation of the effects of the MUR power up rate on the Fermi EOPs and AOPs, and concludes that the proposed MUR power uprate does not present any adverse impacts on the EOPs and AOPs. This conclusion is based upon: (1) the licensee revising the EOPs and AOPs to reflect the new power level and revised setpoints; and (2) the minor changes being made to the EOPs and AOPs will be reflected in the operator

training program prior to MUR implementation. The NRC staff finds that the statements and commitments provided by the licensee are in conformance with Sections VII.2.A, Vl1.3, and Vll.4 of Attachment 1 to RIS 2002-03.

Changes to Control Room Controls. Displays. and Alarms In its submittal, the licensee stated that changes to control room controls, displays (including the Safety Parameter Display System), and alarms related to the proposed MUR power uprate

would be identified and completed prior to MUR implementation. The licensee's review of plant systems indicated that only minor modifications that redefine the new 100 percent power are necessary. The licensee stated that the proposed changes to the control room displays, alarms, and controls will reflect the increased power level, but will not impact the operator's ability to perform operator actions or the available times needed to complete certain operator actions during accident scenarios. All changes to the control room, including modifications involving the Cameron/Caldon LEFM system installed in 2010, have been, or will be reflected in the operator training program prior to MUR implementation.

The NRC staff has reviewed the licensee's evaluation and proposed changes to the control roqm. The staff concludes that the proposed changes do not present any adverse effects to the operators' functions in the control room. The NRC staff finds that the statements provided by DTE are in conformance with Sections \/11.2.8 and Vll.3 of Attachment 1 to RIS 2002-03.

Control Room Plant Reference Simulator and Operator Training The licensee stated that the plant simulator will be modified to reflect the control room changes required for the MUR power uprate. DTE also stated that the changes made to the EOPs and AOPs will be included in the operator training program, and integrated with the control room modifications. The licensee has committed to making these changes prior to MUR implementation.

The NRC staff has reviewed DTE's proposed changes to the operator training and plant simulator related to the MUR power uprate. The staff concludes that the changes will not cause any adverse effects to the plant simulator or the operator training program. The licensee committed to make all modifications to the plant simulator and incorporating these changes into the operator training program prior to MUR power uprate implementation. The NRC staff finds that the statements provided by the licensee are in conformance with Sections VII.2.C, VII.2.D, and Vll.3 of Attachment 1 to RIS 2002-03.

Temporary Operation above Full Steady-State Licensed Power Levels In response to Section Vll.4 of Attachment 1 to RIS 2002-03, the licensee provided the following statement:

Plant operating procedures related to temporary operation above full steady-state licensed power levels (e.g., Fermi 2 procedure MOP03, "Policies and Practices") will be revised as part of the implementation of the MUR License Amendment, to reduce the magnitude of the allowed deviation from the licensed power level, as applicable.

The NRC staff finds that the statement provided by the licensee is acceptable and conforms to Section V11.4 of Attachment 1 to RIS 2002-03.

3.2.3 Conclusion The NRC staff has completed its human performance review of DTE's proposed changes and concludes that the licensee has adequately considered, or will consider, the impact of the proposed MUR power uprate on operator actions, EOPs and AOPs, control room components, the plant simulator, and operator training programs prior to MUR power uprate implementation.

The NRC staff also finds that the statements provided by DTE are in conformance with Section VII of Attachment 1 to RIS 2002-03.

3.3 Dose Consequences Analysis 3.3.1 Regulatory Evaluation A revision to 10 CFR Part 50, Appendix K, effective July 31,2000, allowed licensees to use a power uncertainty less than 2 percent in design basis LOCA analyses, based on the use of state-of-the art FW flow measurement devices that provide for a more accurate calculation of power. Originally, Appendix K of 10 CFR 50 did not require that the power measurement uncertainty be determined, but instead required a 2 percent margin. The revision to 10 CFR 50, Appendix K, allows licensees to justify a smaller margin for power measurement uncertainty based on power level instrumentation error. A power uprate of this type is commonly referred to as an MUR.

NRC Regulatory Issue Summary RIS 2002-03, "Guidance on the Content of Measurement Uncertainty Recapture Power Uprate Applications," recommends that to improve efficiency of the staff's review, licensees requesting an MUR uprate should identify existing design-basis analysis (DBA) analysis of record (AOR), which bound plant operation at the proposed uprated power level. For any existing DBA analyses of record that do not bound the proposed uprated power level, the licensee should provide a detailed discussion of the re-analysis. This SE section documents the NRC staff'~ review of the impact of the proposed changes on analyzed DBA radiological consequences.

On September 28, 2004, the NRC issued Amendment No. 160 to Facility Operating License, NPF-43, for Fermi 2 (ADAMS Accession No. ML042430179), approving the implementation of the alternative source term (AST) for the LOCA and the fuel handling accidents (FHA) in accordance with 10 CFR 50.67, and following the guidance provided in applicable sections of Regulatory Guide (RG) 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors." The licensee originally proposed a full implementation of the AST consistent with the guidance provided in RG 1.183, but subsequently reduced the amendment scope to a selective implementation involving only the LOCA and FHAs. Use of an AST at Fermi 2 for FHAs had been previously approved in two separate amendments (license Amendment Nos. 143 and 144). Therefore, the NRC staff conducted this evaluation to verify that the results of the licensee's LOCA and FHA DBA rad!ological dose consequence analyses continue to meet the dose acceptance criteria given in 10 CFR 50.67 and RG 1.183 at the proposed uprated power level.

Following the guidance provided in applicable sections of RG 1.195, "Methods and Assumptions for Evaluating Radiological Consequences of Design Basis Accidents at Light-Water Nuclear Power Reactors," the NRC staff also conducted an evaluation to verify that the results of the licensee's remaining DBA radiological dose consequence analyses continue to meet the dose acceptance criteria given in 10 CFR 100.11 and RG 1.195 for offsite doses and control room doses at the proposed uprated power level.

. The NRC staff utilized the regulatory guidance provided in applicable sections of RG 1.183, RG 1.195, NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants," (SRP) Section 6.4, for control room habitability, and Fermi 2 UFSAR Chapter 15, in performing this review.

3.3.2 Technical Evaluation The NRC staff reviewed the regulatory ~md technical analyses performed by the licensee in support of its proposed MUR power uprate license amendment, as they relate to the radiological consequences of DBA analyses. Information regarding these analyses was provided by the licensee in Enclosure 1 to the February 7, 2013, application. The findings of this SE are based on the descriptions and results of the licensee's analyses and other supporting information docketed by the licensee.

The NRC staff reviewed the impact of the proposed 1.64 percent MUR power uprate on DBA radiological consequence analyses, as documented in Chapter 15 of the UFSAR. The specific DBA analyses that were reviewed are as follows:

UFSAR Section 15.3.3 Recirculation Pump Seizure

UFSAR Section 15.4.9 Control Rod Drop Accident

UFSAR Section 15.6.2 Instrument Line Pipe Break

UFSAR Section 15.6.4 Steam System Piping Break Outside Containment

UFSAR Section 15.6.5 LOCA Inside Containment

UFSAR Section 15.6.6 Feedwater Line Break Outside Containment

UFSAR Section 15.7.4 Fuel Handling Accident In its application, the licensee indicated that postulated DBA events have been evaluated and analyzed to show that the NRC regu.lati9ns are met for 2 percent above the CLTP. DBA events have either been previously analyzed at 102 percent of CLTP or are not dependent on co~e thermal power. The main steam line break (MSLB) outside containment (as well as the Instrument Line Break) was evaluated using a 4 Ci/g dose equivalent 1-131 limit on reactor coolant activity. The limit on reactor coolant activity is unchanged for the Thermal Power Optimization (TPO) uprate condition. The evaluation/analysis was based on the methodology, assumptions, and analytical techniques described in the RGs, the SRP (where applicable}, and.

in previous safety evaluation reports (SERs). The NRC staff confirmed that the current licensing basis (CLB) dose consequence analyses remains bounding at the proposed MUR uprated power level of 3486 MWt with a margin that is within the assumed uncertainty associated with advanced flow measurement techniques, including use of the Cameron International CheckPius LEFM system credited by the licensee.

The NRC staff also confirmed that the licensee has accounted for the potential for an increase in measurement uncertainty should the LEFM system experience operational limitations. An out-of-specification condition will result in a self-diagnostic alarm condition, either for "Major Alert" status (i.e., increased flow measurement uncertainty), or "Fail" status and the licensee will reduce power to 3430 MWt (the pre-MUR approved RTP) in accordance with the Technical Requirements Manual (TRM) to ensure that the CLB dose consequence analyses remain bounding. In these cases, the LEFM will be considered non-operational and the proposed TRM actions will be applied. Additionally, if the interface between the LEFM system and the plant process computer has failed, the LEFM will be considered non-operational and the proposed TRM .actions .will be applied. Using the licensing basis documentation as contained in the

current Fermi 2 UFSAR, in addition to information in the February 7, 2013, application, the staff verified that the existing Fermi 2 UFSAR Chapter 15 radiological analyses and release assumptions bound the conditions for the proposed 1.64 percent power uprate to 3486 MWt, considering the higher accuracy of the proposed FW flow measurement instrumentation.

3.3.3 Conclusion As described above, the NRC staff reviewed the assumptions, inputs, and methods used by the licensee to assess the radiological consequences of the postulated DBA dose consequence analyses at the proposed uprated power level. The staff finds that operating the Fermi 2 at the proposed uprated power level will continue to meet the applicable dose limits following implementation of the proposed 1.64 percent MUR power uprate. The NRC staff further finds reasonable assurance that Fen:ni 2, as modified by this approved license amendment, will continue to provide sufficient safety margins, with adequate defense-in-depth, to address unanticipated events and to compensate for uncertainties in accident progression, analysis assumptions, and input parameters. Therefore, the NRC staff concludes that the proposed license amendment is acceptable with respect to the radiological dose consequences of the DBAs ..

3.4 Fire Protection 3.4.1 Regulatory Evaluation The purpose of the fire protection program is to provide assurance, through a defense-in-depth design, that a fire will not prevent the performance of necessary plant safe-shutdown functions, nor will it significantly increase the risk of radioactive releases to the environment. The NRC staff's review focused on the effects of the increased decay heat due to the MUR power uprate on the plant's safe-shutdown anaiysis to ensure that structures, systems, and components (SSCs) required for the safe-shutdown of the plant are protected from the effects of the fire and wil.l continue to be able to achieve and maintain safe-shutdown following a fire. The NRC's criteria for the fire protection program are based on (1) 10 CFR 50.48, "Fire protection," insofar as it requires the development of a fire protection program to ensure, among other things, the capability to safely shutdown the plant; (2) General Design Criterion (GDC) 3 of Appendix A to 10 CFR Part 50, insofar as it requires that [a] SSCs important to safety be designed and located to minimize the probability and effect of fires, [b) noncombustible and heat resistant materials be used, and [c] fire detection and suppression systems be provided and designed to minimize the adverse effects of fires on SSCs important to safety; and (3) Criterion 5 of Appendix A to 10 CFR Part 50, insofar as it requires that SSCs important to safety not be shared among nuclear power units unless it can be shown that sharing will not significantly impair their ability to perform their safety functions.

GDC 3 requires that SSCs important to safety shall be designed and located to minimize, consistent with other safety requirements, the probability an~ effect of fires and explosions.

Noncombustible and heat resistant materials shall be used wherever practical throughout the unit, particularly in locations such as the containment and control room. Fire detection and fighting systems of appropriate capacity and capability shall be provided and designed to minimize the adverse effects of fires on structures, systems, and components important to safety. Firefighting systems shall be designed to assure that their rupture or inadvertent operation does not significantly impair the safety capability of these structures, systems, and components ..

GDC 5 requires that SSCs important tq safety shall not be shared among nuclear power units unless it can be shown that such sharing will not significantly impair their ability to perform their safety functions, including, in the event of an accident in one unit, an orderly shutdown and cooldown of the remaining units.

A revision to 10 CFR Part 50, Appendix K, effective July 31, 2000, allowed licensees to use a power uncertainty of less than 2 percent in design basis LOCA accident analyses, based on the use of state-of-the-art FW flow measurement devices that provide for a more accurate calculation of reactor power. Appendix K to 10 CFR Part 50 did not originally require that the reactor power measurement uncertainty be determined, but instead required a 2 percent margin. The revision allows licensees to justify a smaller margin for power measurement uncertainty based on power level instrumentation error. This type of change is also commonly referred to as an MUR power uprate.

RIS 2002-03, "Guidance on the Content of Measurement Uncertainty Recapture Power Uprate Applications," dated January 31, 2002, Attachment 1, Sections II and Ill (ADAMS Accession No. ML013530183) recommends that, to improve the efficiency of the NRC staff's review, licensees requesting an MUR power uprate should identify current accident and transient analyses of record which bound plant operation at the proposed uprated power level. For any DBA, for which the existing analyses of record do not bound the proposed uprated power level, the licensee should provide a detailed discussion of the re-analysis.

3.4.2 Technical Evaluation The licensee developed the LAR consistent with the guidelines in RIS 2002-03. In the LAR, the licensee re-evaluated the applicable SSCs and safety analyses at the proposed MUR core power level of 3486 MWt against the previously analyzed core power level of 3430 MWt.

The NRC staff reviewed Enclosure 7 to NRC-13-004, "General Electric Company (GE)-Hitachi Safety Analysis Report for Fermi 2, Thermal Power Optimization, NEDC-33578P, Revision 0, January 2013, Section 6.7, "Fire Protection." The staff also reviewed the licensee's commitment to 10 CFR 50.48, "Fire protection" (i.e., approved fire protection program). The review covered the impact of the proposed MUR power uprate on the results of the safe-shutdown fire analysis as noted in RIS 2002-03, Attachment 1, Sections II and Ill. The review focused on the effects of the MUR power uprate on the post-fire safe-shutdown capability and increase in decay heat generation following plant trips.

The NRC staff reviewed Section 6.7 of Enclosure 7 of the February 7, 2013, LAR, and identified areas in which additional information was necessary to complete the review of the proposed MUR power uprate LAR. In its June 7, 2013, supplement, the licensee responded to a request for additional information (RAI) regarding its fire protection program.

In its June 7, 2013, response to RAI question 1, which requested that the licensee summarize any changes to the combustible loading and the impact of these changes on the plant's compliance with the fire protection program licensing basis (1 0 CFR 50.48 or applicable portions of 10 CFR 50, Appendix R) as a result of the MUR power uprate, the licensee stated, in part, that the combustible loading calculations have been revised to reflect the addition of cables in trays, and the installation of a new instrument rack as part of this modification. The licensee further stated that the results of the calculations were not impacted and the fire rating of the areas was not changed. Increases in combustible loading as a result of LEFM installation were

satisfactorily evaluated and the combustible loading calculations were appropriately updated, as required by the modification process. *

Further, the licensee stated that the additional physical plant modifications associated with the MUR power uprate consisted of the replacement of two FVV heater relief valves and a number of setpoint changes that resulted in no changes in combustible loading.

The NRC staff reviewed the licensee's response and found it satisfactory and, thus, RAI question 1 is considered resolved based on the response. The staff notes that, based on the licensee response, the changes to the combustible loading, physical plant modifications and configuration changes identified are not impacted by the MUR power uprate.

In its June 7, 2013 supplement, the licensee responded to RAI question 2, which requested the licensee to verify that procedures necessary for systems required to achieve and maintain safe-shutdown will not change and remain adequate for the MUR power uprate. The licensee responded, in part, that operator actions associated with the fire safe shutdown procedures that are sensitive to power uprate, and the effects of power uprate on the time available for the operator actions, have been identified and evaluated. The time available for operator actions remains unchanged.

In addition, the licensee stated that no new operator actions have been identified, and no additional personnel or equipment are necessary to perform actions in the fire safe-shutdown procedures within their designated times. Thus, the existing fire safe-shutdown evaluation, procedures and resources necessary to achieve and maintain safe-shutdown are unaffected by the MUR power uprate.

The NRC staff reviewed the licensee's response and concluded that it satisfactorily addressed RAI question 2. RAI question 2 is considered resolved based 6n the following. For the MUR conditions, the licensee reviewed and ev~luated its safe-shutdown procedures and the effects of the MUR power uprate on the time available for the operator manual actions and concluded that the time available for operator actions remains unchanged. In addition, the licensee indicated that no new operator manual actions have been identified as a result of the MUR power uprate.

The NRC staff notes that this SE does not approve any new or existing operator manual actions concerning the Fermi 2 fire safe-shutdown analysis.

In its June 7, 2013 supplement, the licensee responded to RAI question 3, which requested the licensee to discuss if Fermi 2 credits aspects of its fire protection system for other than fire protection activities, e.g., utilizing the fire water pumps and water supply as backup cooling or inventory for non-primary reactor systems. The staff requested that, if Fermi 2 credits its fire protection system in this way, the LAR identify the specific situations and discuss to what extent, if any, the MUR power uprate affects these "non-fire-protection" aspects of the plant fire protection system. If Fermi 2 does not take such credit, the staff requested that the licensee verify this as well, and discuss how any non-fire suppression use of fire protection water (if applicable) will impact the ability to meet the fire protection system design demands.

In its June 7, 2013, response to RAI question 3, the licensee stated, in part, that there are situations beyond the design basis in which Fermi 2 could use the fire protection system as an alternate water source when other sources of water are unavailable (i.e., Extreme Damage Mitigation Guidelines (EDMGs) used for implementation of the Fermi 2 8.5.b mitigation

strategies). In* addition to firefighting, the Fermi 2 EDMGs provide guidance for the use of the fire protection system as a potential source of water, if necessary and available, to perform the

following functions: (1) Spent Fuel Pool (SFP) Spray; (2) SFP Makeup; (3) Reactor Pressure Vessel Makeup; (4) Containment Flooding; (5) Condensate Storage Tank Makeup; and (6)

Hotwell Makeup. These beyond-design-basis scenarios are unaffected by the MUR power uprate, since the scenarios are non-power dependent. Therefore, the MUR power uprate h~s no impact on the use of the fire protection system in these situations.

The NRC staff reviewed the licensee's response and concluded that it satisfactorily addressed RAI question 3 based on the following. The licensee identified several provisions to use other features of the fire protection system for non-fire protection functions beyond-design-basis events. The licensee indicated tha't the fire water is utilized to backup the water supply makeup for the SFP, SFP spray, reactor pressure vessel, containment flooding, makeup for condensate storage tank, and hotwell. The licensee indicated that these scenarios are non-power dependent and concluded that all these beyond-design-basis events crediting the fire protection system are unaffected by the MUR power uprate. Therefore, the staff finds the response to RAI question 3 acceptable because the licensee's analysis concluded that all of the above functions of non-fire suppression uses of fire protection water are beyond design-basis and would not be affected by the MUR power uprate.

Based on the licensee's fire-related safe-shutdown assessment and responses to RAI questions, the NRC staff concludes that the licensee has adequately accounted for the effects of the MUR power uprate on the ability of the required fire protection systems to achieve and maintain safe-shutdown conditions. The staff finds this aspect of the capability of the associated SSCs to perform their design-basis functions after implementation of the MUR acceptable with respect to fire protection.

3.4.3 Conclusion Based on the NRC staff's review, the staff has concluded that the proposed MUR power uprate will not have a significant impact on the fire protection program or post-fire safe shutdown capability and, therefore, finds the LAR acceptable with respect to these analyses.

3.5 Materials and Chemical Engineering 3.5.1 Piping Evaluation (Erosion/Corrosion: Flow-accelerated Corrosion (FAC))

3.5.1.1 Regulatory Evaluation FAC is a corrosion mechanism occurring in carbon steel components exposed to single-phase or two-phase water flow. Components made from stainless steel are immune to FAC, and FAC is significantly reduced in components containing even small amounts of chromium or molybdenum. The rates of material loss due to FAC depend on the system flow velocity, component geometry, fluid temperature, steam quality, oxygen content, and pH. During plant operation, it is not normally possible to maintain all of these parameters in a regime that minimizes FAC; therefore, loss of material by FAC can occur. The NRC staff reviewed the effects of the proposed MUR power up rate on FAC and the adequacy of the licensee's FAC program to predict the rate of material loss so that repair or replacement of damaged components could be made before reaching a critical thickness. The licensee stated that the FAC program is based on Generic Letter 89-08, Erosion/Corrosion-Induced Pipe Wall Thinning, dated May 2, 1989 (ADAMS Accession No. ML031200731), the guidelines in the Electric Power Research Institute's (EPRI's) Report NSAC-202L, Recommendation for an Effective Flow-

accelerated Corrosion Program, and the Institute of Nuclear Power Operations' (INPO's) report

IN PO EPG-06, Engineering Program Guide - Flow Accelerated Corrosion (FAC). The NRC's acceptance criteria are based on the structural evaluation of the minimum acceptable wall thickness for the components undergoing degradation by FAC.

3.5.1.2 Technical Evaluation The licensee stated that the impact of the proposed MUR power uprate on FAC is minor. The licensee's FAC monitoring program, which includes a predictive method to calculate wall thinnin"g of components susceptible to FAC, was evaluated at MUR power uprate conditions by the licensee and the licensee stated that the predicted wall thinning of balance of plant piping will have a minimal effect on components susceptible to FAC. Additionally, the licensee stated that the FAC program will be updated to include the effects of the MUR power uprate and changes to piping inspections will be made to the FAC program to ensure that adequate margin continues to exist for the systems affected by the up rate.

The NRC staff reviewed the licensee's submittal and the predicted overall change in plant parameters resulting from the MUR pow)3r uprate and finds that an increase of RTP by 1.64 percent will have only a minor effect on the systems susceptible to FAC. Additionally, the NRC staff found that the current FAC program incorporates adequate conservatism to ensure that components susceptible to FAC are managed appropriately prior to exceeding minimum wall thickness. The NRC staff found that the updated FAC program, with the incorporated system changes resulting from the MUR power uprate, will provide reasonable assurance that components susceptible to FAC will be managed appropriately post MUR power uprate implementation.

3.5.1.3 Conclusion The NRC staff has reviewed the licensee's evaluation of the proposed MUR power uprate on the FAC analysis and concludes that the licensee has adequately addressed the impact of changes in plant operating conditions on the FAC analysis. Additionally, the NRC staff concludes that the licensee has demonstrated that the updated analyses will predict, with reasonable assurance, the loss of material by FAC, and will ensure timely repair or replacement of degraded components following implementation of the proposed MUR power uprate.

Therefore, the NRC staff finds the proposed MUR power uprate acceptable with respect to FAC.

3.5.2 Reactor Water Cleanup System .

3.5.2.1 Regulatory Evaluation The reactor water cleanup (RWCU) system provides a means for maintaining reactor water quality by filtration and ion exchange and a path for removal of reactor coolant when necessary.

Portions of the RWCU comprise the reactor coolant pressure boundary (RCPB). Applicable acceptance criteria for the RWCU are based on (1) 10 CFR 50 Appendix A General Design Criterion 14 (GDC-14), Reactor Coolant Pressure Boundary, as it requires that the RCPB be designed, fabricated, erected, and tested to have an extremely low probability of rapidly propagating fracture; (2) GDC-60, Control of Releases of Radioactive Materials to the

Environment, as it requires that the plant design include means to control the release of radioactive effluents; and (3) GDC-61, Fuel Storage and Handling and Radioactivity Control, as it requires systems that contain radioactivity to be designed with appropriate confinement.

Specific review criteria are contained in SRP Section 5.4.8, Reactor Water Cleanup System (BWR).

3.5.2.2 Technical Evaluation The licensee stated that the generic evaluation of the RWCU system is provided in NEDC-32938P, Generic Guidelines for General Electric Boiling Water Reactor Thermal Power Optimization, Sections 5.6.6 and J.2.3.4. The licensee concluded that there is no significant effect on operating temperature and pressure conditions in the high pressure portion of the system and there will be no effect on reactor water chemistry due to MUR power uprate conditions.

The NRC staff's review of the RWCU included component design parameters for flow; temperature, pressure, heat removal capability, impurity removal capability; and the instrumentation and process controls for proper system operation and isolation. The review consisted of evaluating the adequacy of the plant's TSs in these areas under proposed MUR power up rate conditions. In addition, the staff reviewed NEDC-32938P (ADAMS Accession No. ML023170605) as well as the NRC safety evaluation of NEDC-32938P (ADAMS Accession No. ML031050138). The staff finds that the requested MUR power uprate is applicable to NEDC-32938P and that the associated NRC safety evaluation pertaining to the NEDC-32938P is applicable to the requested MUR power uprate. Section 5.6.6 of the NRC safety evaluation states, in part, that the flow through the RWCU system is not significantly affected by reactor power and recirculation flow conditions, therefore the increase of rated power due to a thermal power optimization uprate will not affect system capability. Based on the information reviewed, the staff finds that the effects of the proposed MUR power uprate on the RWCU system are acceptable and the staff has reasonable assurance that the RWCU system will continue to perform its designed function.

3.5.2.3 Conclusion The NRC staff has reviewed the licensee's evaluation of the effects of the proposed MUR power uprate on the RWCU system and concludes that the licensee has adequately addressed changes to the reactor coolant and its effects on the RWCU system. The staff further concludes that the licensee has demonstrated that the analyses of record for the RWCU system will continue to be acceptable and meet the requirements of GDC-14, -60, and -61. Therefore, the staff finds the proposed MUR power uprate acceptable with respect to the RWCU system.

3.5.3 Containment Coatings 3.5.3.1 Regulatorv Evaluation Protective coating systems (paints) provide a means for protecting the surfaces of facilities and equipment from corrosion and contamination from radionuclides and also provide wear protection during plant operation and maintenance activities. The NRC staff's review covered protective coating systems used inside the containment for their suitability for and stability under design basis LOCA accident conditions, considering radiation and chemical effects. The NRC's acceptance criteria for protective coating systems are based on ( 1) 10 CFR Part 50, Appendix B, which states quality assurance requirements for the design, fabrication, and construction of safety-related structures, systems, and components (SSCs) and (2) Regulatory Guide (RG) 1.54, Revision 2, Service Levell, II, and Ill Protective Coatings Applied to Nuclear

Power Plants. Specific review criteria are contained in SRP Section 6.1.2, Protective Coating Systems (Paints)- Organic Materials Review Responsibilities.

3.5.3.2 Technical Evaluation The licensee stated that current containment coatings will continue to bound the DBA temperature, pressure, and radiation levels at MUR power uprate conditions. The coatings are qualified to temperature, pressure, and radiation levels of 340°F, 70 psi, and 1x1 0 9 Rads, respectively. The licensee indicated that the expected post DBA environment conditions in containment, post MUR power uprate implementation, with respect to temperature, pressure, and radiation levels are 340°F, 56 psi, and 4.47x1 08 Rads, respectively. As such, the Service Level I coating qualifications continue to bound the DBA temperature, pressure, and radiological profiles under the proposed power uprate conditions.

The NRC staff reviewed the licensee's amendment request' and the Fermi 2 UFSAR and finds that, based on the qualification of the current Service Level I coatings continuing to bound the predicted conditions in containment post DBA following MUR power up rate implementation, the NRC staff has reasonable assurance that the coatings will not be adversely impacted by the MUR power uprate. Therefore, the staff finds the MUR power uprate request acceptable with regards to protective coatings.

3.5.3.3 Conclusion The NRC staff has reviewed the licensee's evaluation of the effects of the proposed MUR power uprate on protective coating systems and concludes that the licensee has appropriately addressed the impact of changes in conditions following a DBA and their effects on the protective coatings. The staff further concludes that the licensee has demonstrated that the protective coatings will continue to be acceptable following implementation of the proposed MUR power uprate. Specifically, the protective coatings will continue to meet the requirements of 10 CFR Part 50, Appendix B, and RG 1.54. Therefore, the NRC staff finds the proposed MUR power uprate acceptable with respect to protective coatings systems 3.6 Mechanical and Civil Engineering 3.6.1 Regulatory Evaluation The NRC staff's review of the LARin the areas of mechanical and civil engineering focused on verifying that the licensee has provided reasonable assurance that the structural and pressure boundary integrity of SSCs at Fermi 2 will continue to be adequately maintained following the implementation of the MUR power uprate under normal and abnormal loading conditions, as stipulated by the facility's design-basis requirements. Reasonable assurance is provided by demonstrating compliance with the NRC regulations listed below:

The NRC staff's assessment of the proposed MUR power uprate in the area of mechanical and civil engineering considered the following regulations: 10 CFR 50.55a and GDCs 1, 2, 4, 10, 14, and 15, which are located in 10 CFR 50, Appendix A. The acceptance criteria are based on continued conformance with the requirements of the following regulations: (1) 10 CFR 50.55a, and GDC 1 as they relate to structures and components being designed, fabricated, erected, constructed, tested, and inspected to quality standards commensurate with the importance of the safety function to be performed; (2) GDC 2 as it relates to structures and components important to safety being designed to withstand the effects of earthquakes combined with the.

effects of normal or accident conditions; (3) GDC 4 as it relates to structures and components important to safety being designed to accommodate the effects of, and to be compatible with, the environmental conditions of normal and accident conditions and these structures and components being appropriately protected against dynamic effects, including the effects of missiles, pipe whipping, and discharging fluids; (4) GDC 10 as it relates to the design of reactor internals with appropriate margin to assure that specified acceptable fuel design limits are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences; (5) GDC 14 as it relates to the reactor coolant pressure boundary being designed, fabricated, erected, and tested to have an extremely low probability of abnormal leakage, of rapidly propagating failure, and of gross rupture; and (6) GDC 15 as it relates to the reactor coolant system (RCS) being designed with sufficient margin to ensure that

the design conditions are not exceeded.

The design and licensing bases for the facility establish the principal means with which the facility demonstrates compliance with applicable NRC regulations. As such, the NRC staff's review primarily focused on verifying that the design and licensing basis requirements related to the structural and pressure boundary integrity of SSCs affected by the proposed MUR power uprate, will continue to be satisfied at MUR conditions. This, in turn, provides reasonable assurance that continued compliance with the applicable regulations will be maintained.

Section 3.1 of the Fermi 2 UFSAR describes how the facility complies with the GDC.

The primary guidance used by DTE and other licensees for LARs involving MUR power uprates is outlined in RIS 2002-03, "Guidance on the Content of Measurement Uncertainty Recapture Power Uprate Applications," which provides licensees with a guideline for organizing LAR submittals for MUR power uprates. *section IV of RIS 2002-03, "Mechanical/Structural/Material Component Integrity and Design," provides information to licensees on the scope and detail of the information* that should be submitted to the NRC staff regarding the impact an MUR power uprate has on the structural and pressure boundary integrity of SSCs affected by the implementation of an MUR power uprate ..

Section IV.1 .A of RIS 2002-03 notes that certain SSCs must be evaluated to determine whether these SSCs are able to support the implementation of the proposed MUR power uprate. The evaluations discussed in Section IV of RIS 2002-03 focus on determining what impact the MUR power uprate would have on the analysis of record (AOR) for a particular sse in order to determine whether the AOR needs to be revised as a result of the power uprate. Furthermore,Section IV.1.B of RIS 2002-03 indicates that the AOR for those SSCs that are affected by implementation of an MUR power uprate, the licensee should address the following, as they relate to the impact of the uprate on the AOR: stresses, cumulative usage factors (CUFs) (i.e.,

fatigue), flow induced vibration (FIV), and changes in temperature, pressure and flow rates resulting from the power uprate. As discussed in the RIS, if the AOR for a particular SSC was performed using conditions that bound those that will be present at the MUR power level, no further.evaluation is required.

In reviewing the Fermi 2 MUR power uprate LAR, the NRC staff also considered the guidance provided by proprietary report NEDC-32938P-A, "Generic Guidelines and Evaluations for General Electric Boiling Water Reactor Thermal Power Optimization," dated November, 2002, also referred to as the "TLTR." This NRC-approved topical report outlines the format and content which should be included in LARs for MUR power uprates at boiling water reaCtor (BWR) commercial power facilities. The NRC staff's SER accompanying the TL TR was issued on April 1, 2003, and documents the NRC staff's review and approval of the methods described

in the TLTR for licensee's proposing to implement an MUR power uprate (ADAMS Accession No. ML031050138).

3.6.2 Technical Evaluation The NRC staff's review in the areas of mechanical and civil engineering covered the structural and pressure boundary integrity of the piping, components and supports which make up the nuclear steam supply system and the balance-of-plant (BOP) systems. The mechanical and civil engineering revieyv scope also included an evaluation of other new or existing SSCs which are affected by the implementation of the proposed MUR power uprate. Specifically, this review focused on the impact of the proposed MUR power uprate on the structural integrity of the Fermi 2 pressure-retaining components and their supports and the reactor vessel internals (RVIs).

The NRC staff's review also considered the impact of the proposed MUR power uprate on postulated high-energy line break (HELB) locations and corresponding dynamic effects resulting from the postulated HELBs, including pipe whipping and jet impingement. The staff's review focused on verifying that the licensee has provided reasonable assurance of the structural and pressure boundary integrity of the aforementioned piping systems, components, component' internals, and their supports under normal and transient loadings, including those due to postulated accidents and natural phenomena, such as earthquakes.

The technical bases for the proposed MUR power uprate at Fermi 2 are documented in to the licensee's February 7, 2013, submittal. Enclosure 7, "Safety Analysis Report for Fermi Generating Station Unit 2 Thermal Power Optimization," also referred to as the Thermal Power Optimization Safety Analysis Report (TSAR). The TSAR is organized in accordance with the TL TR, described previously, to provide the necessary justification for the .

proposed MUR power uprate. Enclosure 9 to the licensee's LAR provides a non-proprietary version of the TSAR (ADAMS Accession No: ML13043A655). In accordance with Appendix K requirements denoted in the SER Section 2.0, the licensee noted in Section 1.2.1 of the TSAR that the current ECCS analyses of record AOR are based on a core power level of 102 percent of the CLTP of 3,430 MWt. As such, the licensee has previously performed these analyses assuming a power level of 3,498.6 MWt and the implementation of the proposed MUR power*

uprate would revise the licensed thermal power to a level lower than the licensee has already analyzed.

3.6.2.1 Power Uprate Evaluation Parameters and Design Bases Table 1-2 of the TSAR shows the pertinent temperatures, pressures, and flow rates for the current and proposed (uprated) conditions. At full power, the dome temperature remains at a constant 550.0 degrees Fahrenheit (°F) from the current to uprated conditions. Similarly, the dome pressure remains at a constant 1045.0 pounds per square inch absolute (psi a) from the current to uprated conditions. At the revised power level, the steam flow increases from 14.86 million pounds per hour (Mib/hr) to 15.14 Mlb/hr. Similarly, the FW flow rate increases from 14.83 Mlb/hr to 15.11 Mlb/hr. The FW temperature increases from 424.5 °F to 426.5 °F at the revised power level. With respect to the full power core flow range, the minimum core flow increases from 81.0 Mlb/hr to 83.0 Mlb/hr, while the maximum core flow remains at 105.0 Mlb/hr. The proposed uprate does not change heatup or cooldown rates, or the number of cycles assumed in the design analyses. In addition, the limiting analyses for design transients are still bounding.

The information related to the *structural qualification of SSCs at Fermi 2 is contained in Chapter 3 of the Fermi 2 UFSAR. The information contained in the UFSAR documents the design

criteria applicable to the Fermi 2 SSCs, including loads, load combinations, and acceptance criteria stipulated by the applicable codes of record for these SSCs. Additional information regarding the design specifications, functional description, design loads, and design code requirements for the reactor internals is located in Section 4.5 of the Fermi 2 UFSAR. The NRC staff considered the effects of the proposed MUR power uprate on the aforementioned design basis information in its review. The NRC staff noted that most of the design basis loads for*

SSCs are unaffected by the MUR power uprate due to the following: (1) the changes in temperatures, pressures, and flow rates associated with the MUR power uprates are generally small, and (2)the design bases for SSCs typically consider loads that bound those that will be experienced by SSCs at the revised power level.

3.6.2.2 Pressure-Retaining Components and Component Supports In accordance with Section IV.1.A of RIS 2002-03, the licensee evaluated the Fermi 2 pressure-retaining components and supports, including the reactor vessel, RCPB piping, and BOP piping, to determine whether these SSCs would be affected by the proposed power uprate. These evaluations were performed to demonstrate that the design basis acceptance criteria specific to these SSCs would remain satisfied following implementation of the proposed MUR power uprate.

In reviewing the licensee's evaluation of pressure-retaining components and their supports, the NRC staff focused its review on components and supports that would be affected by implementing of the MUR power Liprate. Affected components and supports refer to those for which their AOR is not bounded at MUR conditions. Pressure-retaining components and their supports generally remain unaffected by the implementation of an MUR power uprate based on the fact that they have been analyzed at conditions that are more limiting than those that will be present at MUR conditions (i.e., bounded).

Reactor Vessel The licensee's assessment of the structural integrity of the reactor vessel to support the implementation of the proposed power uprate is documented in Section 3.2.2 of the TSAR. Due to the generally insignificant increases in pressure, temperature, and flow rates accompanying an MUR power uprate, the structural integrity of most reactor vessel components remain unaffected by this type of power uprate. However, as indicated in the TL TR, certain components, such as the FW nozzles, main steam nozzles, and the FW sparger, may require re-analysis to demonstrate continued compliance with the design code requirements for these SSCs. To this end, the licensee presented the results of the reconciliations performed for the stress and fatigue analyses of the limiting reactor vessel components associated with the proposed power uprate (i.e., the N6 FW nozzle, N11 recirculation outlet nozzle, and N10 recirculation inlet nozzle) in Table 3-4 of the TSAR.

The NRC staff issued two RAis to the licensee regarding the stress increases associated with these components at the proposed power level. In its July 15, 2013, response, the licensee indicated that its use of The American Society of Mechanical Engineers (ASME) Code Case 1441 to disposition instances where the 3Sm primary-plus-secondary stress limit was exceeded, is consistent with the Fermi 2 current licensing basis. Additionally, the licensee also indicated that the primary-plus-secondary stress results, included in Table 3-4 of the TSAR for the recirculation inlet and outlet nozzles, are based on a thermal power level of 4,031 MWt, which is above the proposed power level. The NRC staff also issued one RAI to the licensee regarding its evaluation of the FW sparger to support the proposed power uprate. In its July 15, 2013,

'"4.1 -

response, the licensee indicated that the controlling loads on the FW sparger are thermal, seismic, and annulus pressurization loads, of which only the thermal loads are affected by the proposed power uprate. To this end, the licensee indicated that the increased FW temperature actually reduces the thermal stresses imposed on the FW sparger. As such, the loads for the normal, upset, emergency, and faulted loading conditions associated with the FW sparger remain bounded by those considered in the current AOR for the FW sparger, and this component is unaffected by the proposed power uprate.

RCPB Piping Systems The licensee's assessment of the Fermi 2 RCPB piping to support the proposed power uprate is documented in Section 3.5.1 of the TSAR. The scope of this assessment includes the MS and FW piping systems, inside containment, and piping connected to these systems. Additionally, the licensee indicated that nozzles, supports, penetrations, valves, pumps, heat exchangers, and anchors associated with the RCPB were evaluated for acceptability at the proposed power level. These evaluations were performed using the methods described in Appendix K of the TL TR to determine the displacements, stresses, CUFs, and interface. loads in the aforementioned piping, components, and supports at the revised power level.

This method has been used for previous power uprates, including extended power uprates (EPUs), and esse*ntially scales the displacements, stresses, CUFs, and interface loads in affected piping systems in accordance with the increased temperature, pressure, or flow rate a system may encounter at the revised power level.

For the MS system, the licensee indicated that the increased MS flow rate associated with the proposed power uprate does result in increased loading on the MS piping and MS piping system supports, including snubbers, hangers, struts, and whip restraints. However, the licensee determined that these increased loads do not result in any design code allowable values being exceeded at the revised power level. As such, the structural integrity of the MS piping system inside containment was found to be acceptable at the proposed power level. Similarly, the portions of the FW system inside containment were evaluated for the effects of increased temperature and flow rate associated with the increased power level. The licensee indicated that while the loads induced on the FW piping system are affected by the aforementioned parameter changes associated with the proposed power uprate, all of the design code allowable values remain satisfied at the revised power level. Additionally, the licensee notes thatFW piping supports inside containment are able to withstand the increased FW flow rate and temperature, as indicated by the continued satisfaction of the design basis acceptance criteria for these supports at the proposed power level.

Balance of Plant (BOP) Piping Systems The licensee's assessment of BOP piping is documented in Section 3.5.2 of the TSAR.. In this section, the licensee identified 24 piping systems, or portions of piping systems evaluated in support of the proposed MUR power uprate; these systems include the portions of the FW and MS piping outside containment. The licensee indicated that a portion of the FW system outside containment was the only BOP piping which required re-analysis to support the proposed MUR power uprate; all other systems or portions of systems were identified as unaffected or have insignificant stress increases as a result of changes in pressure, temperature, or flow rate accompanying the power uprate.

In its July 15, 2013, response to an NRC staff RAI regarding the re-analysis of the FW piping

denoted above, the licensee indicated that this portion of piping was re-analyzed due to the fact that an LEFM was installed in the FW piping system outside containment. As such, the piping models associated with this portion of the FW system were updated to reflect the addition of the LEFM and the revised operating conditions associated with the MUR power uprate. There-analysis of the updated model showed that the maximum stresses in the piping remain within the allowable values prescribed by the applicable codel5 of record associated witti this portion of the FW piping. The licensee's evaluation of BOP piping systems also included a review of increased pipe support loads for affected BOP piping systems, including MS and FW piping supports. Based on its evaluations, which included the effects of increased fluid transient loadings in the MS and FW systems, the licensee concluded that all supports remain acceptable for operation at the revised power level. *

NRC Staff Evaluation

The NRC staff considered the licensee's assessments of the pressure-retaining components.

and component supports acceptable based on the following rationale. Consistent with the NRC-approved TL TR, the staff acknowledges that many components associated with the reactor vessel will be unaffected by an MUR power uprate due to the fact that the operating parameter changes (i.e., temperature, pressure, and flow rate) are generally small. With respect to the results of the licensee's evaluation of those reactor vessel components affected by the proposed power uprate, the staff noted that where stresses associated with the FW nozzles and recirculation nozzles exceed the applicable primary-plus-secondary stress intensity limits, the licensee performed additional analyses in accordance with the provisions ofthe* ASME Code, including ASME Code Case 1441, to justify the stresses being above the prescribed limits. The NRC staff considered this acceptable, given that these analyses were consistent with the provisions of the ASME Code and, as such, consistent with the design basis requirements associated with the reactor vessel. Considering that the licensee was able to demonstrate

.continued compliance with the applicable reactor vessel design code requirements, the staff concluded that there is reasonable assurance that the reactor vessel will maintain its structural integrity at the proposed power level.

The NRC staff's review of the RCPB piping systems considered the methods used in evaluating these systems for acceptability at the proposed power level and the results of these evaluations.

The staff noted that the licensee utilized the methods of Appendix K in the TL TR to perform these evaluations, and this method has been reviewed and approved by the NRC for generic use through the TL TR in a number of other power MUR uprates. The NRC staff noted that this method produces conservative resl!lts, given that the scaling factors generally overestimate the additional loads which will be imposed on the piping systems at the higher power level. The staff also noted that the licensee was able to demonstrate that all design code allowable values for the .RCPB piping systems will remain satisfied at the revised power level. As such, the NRC staff concluded that there is reasonable assurance that the structural integrity of the RCPB piping systems will be maintained at the revised power level.

In its review of the licensee's assessment of BOP piping systems, the NRC staff noted that only one system will be appreciably affected by the proposed power up rate (i.e., a portion of the FW system outside containment). The staff noted that this is consistent with the TL TR and the staff's SER documenting the acceptance of the TLTR, in that both documents note that it is not expected that BOP piping systems are likely to be significantly affected by an MUR power uprate. For the affected FW piping, the NRC staff noted that the licensee's re-analysis performed to support the LEFM installation found that the design code allowable values for the FW piping will remain satisfied following the implementation of the MUR power uprate. Given

that the design basis requirements for all. BOP piping systems, including supports, will remain satisfied following the proposed power uprate, the*staff concluded that there is reasonable assurance that the Fermi 2 BOP piping systems will maintain their structural integrity at the revised power level.

3.6.2.3 Reactor Vessel Internals In accordance with Section IV.1.A.ii of RIS 2002-03, the licensee evaluated the effects of the proposed MUR power uprate on the Fermi RVIs. As previously stated,Section IV.1.B of RIS 2002-03 indicates that for those SSCs, including RVIs, whose AOR are affected by implementation of an MUR power uprate, the licensee should address the following, as they relate to the impact of the uprate on the AOR: stresses, CUFs (i.e., fatigue), FIV, and changes in temperature, pressure, and flow rates resulting from the power uprate. The licensee summarized its evaluation of the effects of the proposed power uprate on the structural integrity of the RVIs in Section 3.3.2 of the TSAR. Structural evaluations were performed by the licensee to determine any effects on the RVIs, including both core support structures and non-core support structures, due to the conditions that would be present following the implementation of .

the proposed MUR power uprate. Evaluations were also performed to determine the susceptibility of the RVIs to damage due to FIV at the increased power level. These evaluations are documented iri Sections 3.3.3 and 3.4 of the TSAR.

With respect to the structural evaluations performed for the RVIs, Section 3.3.2 of the TSAR indicated that the evaluations performed for the RVIs considered loads due to deadweight, reactor internal pressure difference (RIPD), seismic, annulus pressurization/jet reaction (AP/JR) loads, acoustic loads resulting from a recirculation line break (RLB), hydraulic flow, and thermal loads. The evaluations performed for the RVIs focused on potential increases in any of these.

loads, as a result of the proposed power uprate, and the subsequent effects on the structural integrity of the RVIs. The licensee indicated that the design basis structural integrity evaluations of the Fermi 2 RVIs were performed assuming a power level of 120 percent of the original licensed thermal power (OLTP). As such, the loads which will be imposed on the RVIs at the revised power level are bounded by those in the AOR for the normal, upset, emergency, and faulted loading conditions. Table 3-5 of the TSAR provides a comparison between the governing stresses at the proposed power level and those currently in the AOR for a number of RVI locations. As this table demonstrates, the MUR power uprate does not affect the structural integrity of the RVIs.

Flow-Induced Vibration Section 3.4.2 of Enclosure 1 to the LAR submittal discusses the potential adverse flow effects resulting from the proposed MUR power uprate. A majority of the discussion in this portion of the LAR submittal focuses on the potential degradation of the steam dryer due to increased steam flow rates resulting from the increased power level. Section 3.4 of the TSAR discusses the FIV evaluations performed for other RVIswhich were performed by considering startup testing data from Browns Ferry Nuclear Plant Unit 1 (designated prototype plant). The licensee indicated that the FW sparger, shroud, and shroud head and separator are expected to see increases in FIV at the revised power level. However, when the stresses resulting from the increased FIV for these components were compared to the allowable stress intensity, the licensee concluded that the increased stresses will remain below the established limits. As such, damage resulting from high-cycle fatigue on these RVIs is not expected at the revised power level.

The licensee indicated in the LAR submittal that an independent analysis of the Fermi 2 steam dryer had concluded that loads on the steam dryer will increase slightly at the proposed power level, but the structural integrity of the steam dryer will be maintained. Section 3.3.3 of the TSAR also discusses the evaluations performed for the Fermi 2 steam separator and steam dryer for the purposes of MUR power uprate implementation. In response to an NRC staff RAI regarding the independent analysis performed on the steam dryer, the licensee provided a detailed description of the testing and analyses which were used to determine the increased stresses on the dryer at the revised power level. Specifically, the licensee indicated that the natural frequencies for the safety relief valve (SRV) standpipes, the MS line drain pots, and a capped standpipe branch were determined analytically and verified through scale model testing.

Scale model testing reliability was demonstrated by comparison to previous power ascension data. Based on these evaluations, the licensee determined that the onset of double vortex mode resonances could be expected for most of the openings, identified above, prior to reaching the proposed power level.

The licensee confirmed that the onset of single vortex mode resonances was not expected for any of the* MS line openings. Further, the licensee noted that double vortex mode resonances from SRV standpipes and MS line drain pots currently exist at CLTP. As a result of the increased MS line flow, a 3.4 percent increase in stress on the dryer is expected as a result of MS velocity ratio square increases. Additionally, based on a comparable steam dryer evaluation performed for a similar plant which has implemented an EPU, the licensee indicated that dryer stresses resulting from acoustic resonances will increase by 169 psi. By comparing these increases to the high cycle fatigue acceptance criterion of 13,600 psi, the licensee concluded that the dryer would maintain its structural integrity at the revised power level.

NRC Staff Evaluation

The NRC staff considers the licensee's assessment of the Fermi 2 reactor internals acceptable.

This acceptance is based on the fact that the loads considered in the design bases for the RVIs were developed assuming a power level of 120 percent of OLTP, a level that is above the requested power level of 3486 MWt. As such, the loads that will be imposed on the RVIs at the revised power level are. less than those for which the RVIs have already been evaluated and found acceptable. Therefore, the staff concluded that there is no impact on the structural integrity of the RVIs as it relates to the design basis loads imposed on these SSCs. The staff also considers the licensee's evaluation of the effects of FIV on the RVIs acceptable. For those components evaluated in Section 3.4 of the TSAR, the NRC staff noted that the licensee's approach is consistent with the TL TR and the staff's SER documenting the acceptance of the methods in the TL TR. To this end, the staff notes that the licensee was able to quantitatively demonstrate that the expected stress increases in the RVIs affected by FIV will continue to satisfy the applicable high-cycle fatigue stress limits. As such, the NRC staff concluded that this portion of the licensee's evaluation is acceptable.

The NRC staff's review of the Fermi 2 steam dryer evaluations performed to support the proposed power uprate focused on verifying that the increased steam flow associated with the uprate will not result in a loss of structural integrity of the steam dryer. The staff noted that the approach used by the licensee to evaluate the steam dryer is consistent with the TLTR and its associated SER documenting the staff's acceptance of the TL TR. The NRC staff acknowledged in its SER for the TL TR, that although load increases on the steam dryer associated with the implementation of an MUR power uprate are expected to be small, a plant-specific confirmation of this small increase should be performed. To this end, the staff considers the licensee's approach, which considered stress increases resulting from velocity-squared effects and

acoustic resonances, acceptable and consistent with previous steam dryer evaluations. Based on the fact that the licensee was able to demonstrate that the stress increases expected for the steam dryer will be a small percentage of the acceptable stress limit, the NRC staff concluded that there is reasonable assurance that the steam dryer will maintain its structural integrity under the conditions at the proposed power level.

3.6.2.4 Postulated Pipe Ruptures and Associated Dynamic Effects The licensee evaluated the effects of the proposed MUR power uprate on systems classified as high energy to determine whether any changes to the HELB AOR will result from the implementation of the power up rate. This assessment is summarized in Section 10.1 of the TSAR. As indicated in the summary of the licensee's assessment, the current AOR for HELBs was reviewed to determine whether the MUR power uprate would have any impact on the current HELB AOR. The licensee concluded that because the piping configurations associated with high energy systems are not changed as a result of the proposed power uprate, there are no new postulated rupture locations. Section 10.1.2. 7 of the TSAR summarizes the licensee's evaluations of the effects of the proposed power up rate on the dynamic effects resulting from postulated pipe ruptures (i.e., pipe whipping and jet impingement). The licensee. confirmed that these dynamic effects remain unaffected by the proposed power uprate, given that there is no change in the nominal dome pressure. Consequently, the existing features used to protect SSCs against dynamic effe<;;ts (i.e., pipe whip restraints, .jet impingement shields, and supports) remain adequate. *

NRC Staff Evaluation

The NRC staff reviewed the licensee's evaluations related to determinations of pipe rupture locations and their corresponding dynamic effects and considers the licensee's assessments performed for these areas acceptaq[e. This acceptance is based on the information presented above, which demonstrates that the analyses of record related to postulated pipe ruptures and dynamic effects resulting from postulated pipe ruptures will remain bounding under the proposed MUR power level. The staff considers this 'conclusion reasonable, given the small magnitudes in temperature, pressure, and flow rate increases that accompany MUR implementation. Correspondingly, as previously discussed, these small changes generally have no impact on pressure-retaining components such as piping. Further, the licensee's evaluations are consistent with those discussed in the TL TR and the NRC staff's associated SER documenting the acceptance of the TL TR, which both acknowledge that the changes to postulated pipe ruptures and associated consequences resulting from an MUR power uprate are insignificant.

. 3.6.3 Conclusion The NRC staff has reviewed the licensee's assessment of the impact of the proposed MUR power uprate on the structural and pressure boundary integrity of pressure-retaining components and supports and RVIs. Additionally, the NRC staff reviewed the licensee's assessment of the effects on the Fermi 2 HELB AOR, including associated dynamic effects.

Based on the review delineated above, the NRC staff finds the MUR power uprate acceptable with respect to the structural integrity of the aforementioned SSCs affected by the power uprate.

This acceptance is based on the licensee's demonstration that the intent of the aforementioned regulatory requirements, related to the mechanical and civil engineering purview, will continue to be satisfied following implementation of the power uprate.

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Specifically, the licensee demonstrated that: (1-) the structural and pressure boundary integrity pressure retaining components and supports, including piping and pipe supports, at Fermi 2 are either not affected by the proposed MUR power uprate or will continue to satisfy the design code requirements applicable to affected SSCs; (2) the RVIs at Fermi 2 remain unaffected, when considering the impact ofMUR implementation on the design basis loads associated with

the RVIs, and demonstrate acceptable FIV characteristics at the revised power level; and (3) the Fermi 2 AOR related to the postulation of HELB locations, including dynamic effects associated with these postulated pipe ruptures, remains unaffected by the proposed MUR power uprate.

Based on these considerations, the NRC staff concluded that there is reasonable assurance that the structural integrity of SSCs at Fermi 2 will be adequately maintained following implementation of the MUR power uprate such that the MUR power uprate will not preclude the ability of these SSCs to perform their intended functions.

3.7 Safety Related Valves. Pumps. and Snubbers Safety Related Valves The NRC staff reviewed the licensee's safety-related valve analysis for Fermi Unit 2. The NRC's acceptance criteriafor review are based on 10 CFR 50.55a, "Codes and Standards."

Additional information is also provided by the plant-specific evaluations of Generic Letter (GL) 89-10, "Safety-Related Motor-Operated Valve Testing and Surveillance," GL 95-07, "Pressure Locking and Thermal Binding of Safety-Related Power-Operated Gate Valves," and GL 96-05, "Periodic Verification of Design-Basis Capability of Safety-Related Power-Operated Valves."

In its February 7, 2013, submittal, the licensee reviewed the impact of the proposed MUR power uprate conditions on the existing design basis analyses for the safety-related valves. In Sections 1.3, 3.1, 3.8, 4 ..1, and 5.3.4 of Enclosure 7, "Safety Analysis Report for Fermi Generating Station Unit 2," the licensE?e reviewed the revised design and operating conditions resulting from the MUR power uprate against previous licensing evaluations. In Section 1.3, the licensee indicates that only small changes in reactor coolant flow, operating pressure; and temperature resulted from MUR power uprate. Therefore, no changes to the functional requirements of the existing safety-related valves are identified as a result of the MUR power up rate.

In Section 3.1, the licensee reviewed the MUR power uprate impact on Nuclear System Safety Relief Valves (NSSRVs), and concluded that there was no increase in nominal operating pressure and no changes in the NSSRV setpoints were required for the MUR power uprate conditions. In Section 3.8, the licensee reviewed the MUR power uprate impact on Main Steam Isolation Valves (MSIVs) and concluded that all requirements for the MSIVs remain unchanged for the MUR power uprate conditions~ In Section 5.3.4, which concerns Safety Relief Valves (SRVs), the licensee indicates that because there is no increase in reactor operating dome pressure, the SRVs are not changed. In Section 4.1, the licensee also evaluated the MUR power uprate impact on the requirements of GL 89-10, GL-95-07, andGL 96-05. The evaluation shows that no required changes are identified, and all GL 89-10 motor-operated

.valves remain capable of performing their design basis functions. Since there are insignificant changes in operating conditions and no changes to the design basis requirements, the inservice testing (1ST) program for safety-related valves will not be affected by the MUR power uprate.

The review concluded that the MUR power uprate does not impact the design and operation of the safety-related valves since the operating ranges of pressure, temperature, and flow are bounded by previous evaluations. Therefore, the NRC staff finds that the performance of

existing safety-related valves and the current 1ST program are acceptable with r~spect to the MUR power uprate.

Safety-Related Pumps The NRC staff reviewed the licensee's safety-related pumps analysis. The NRC's acceptance criteria for reviewing the safety-related pumps analysis are based on 10 CFR 50.55a, "Codes and Standards."

In its February 7, 2013, submittal, the licensee reviewed the impact of the proposed MUR power uprate conditions on the existing design basis analyses for the safety-related pumps. In Section 1, the licensee identified small changes in thermal-hydraulic parameters for the thermal power optimization (TPO) uprate. Table 2 of Attachment 7 showed that the operating temperature and pressure ranges and flow conditions for safety-related pumps due to the MUR power uprate are bounded by the original design parameters. Also, the original design transients for the safety-related pump~ bound the transients associated with the MUR power uprate. The licensee stated that the MUR power uprate accident analysis required flows are not changing, and therefore the 1ST program for safety-related pumps will not be affected.

Therefore, the NRC staff finds the performance of existing safety-related pumps with existing 1ST programs to be acceptable with respect to the MUR power uprate.

Snubbers The NRC staff reviewed the licensee's pipe support (including snubbers) analysis for Fermi 2."

The NRC's acceptance criteria for review are based on 10 CFR 50.55a, "Codes and Standards."

In its February 7, 2013, submittal, the licensee reviewed the impact of the proposed MUR power uprate conditions on the existing design basis analyses for the pipe supports, which includes snubbers. In Sections 3.5.1, and 3.5.2 of Attachment 7, "Safety Analysis Report for Fermi Generating Station Unit 2," the licensee reviewed the revised design and operating conditions resulting from the MUR power uprate against previous licensing evaluations. In Sections 3.5.1 and 3.5.2, the licensee states that the review of the increase in flow associated with the TPO uprate indicates that piping load changes do not results in any load limit exceeded. No changes*

to the functional requirements of the existing snubbers and snubber inservice examination and testing are required as a result of the MUR power uprate. Therefore, the NRC staff finds the existing snubber program related inservice examination and testing to be acceptable with respect to the MUR power uprate.

3.8 Accident Analyses In the LAR, the licensee generally concluded that existing analyses were bounding of uprated plant operation with reduced uncertainty. The analyses were shown to be bounding in one of three different ways:

For analyses that assume steady-state plant operation with a core power of 3499 MWt,

there is a 2 percent margin for power measurement uncertainty at the CL TP, 3430 MWt.

These analyses are bounding also of plant operation at the MUR RTP of 3486 MWt. with an operating margin of 0.371 percent, which is greater than the stated 0.36-percent calorimetric power measurement uncertainty.

For analyses that assume steady-state plant opera~ion with a core power of 3430 MW1, the licensee evaluated the accident or transient, and reanalyzed as necessary.

Zero-power transients were not reanalyzed.

RIS 2002-03, "Guidance on the Content of Measurement Uncertainty Recapture Power Uprate Applications," states the following, in part:

When licensees submit measurement uncertainty recapture power uprate applications, the [NRC) staff intends to use the following general approach for their review:

In areas (e.g., accident/transient analyses, components, systems) for which the existing analyses of record do bound plant operation at the proposed uprated power level, the [NRC] staff will not conduct a detailed review.

The NRC staff used such an approach in its review of the LAR. The staff did not review the licensee's analyses that were performed at 102 percent of the CLTP. For those analyses, the NRC staff found that existing AORs bound plant operation. Thus, the NRC staff finds that those AORs are acceptable without a detailed review.

Anticipated Operational Occurrences (AOOs) Analysis The AOOs were evaluated generically for TPO uprates plants under TL TR, NED0-32938 Revision 1, "Licensing Topical Report Generic Guidelines and Evaluations for General Electric Boiling Water Reactor Thermal Power Optimization", Appendix E (ADAMS Accession No. ML023170607). Appendix E provides generic evaluation of transient events for TPO uprated plants, and includes documentation of the bases and assumptions for each evaluation. The plant-specific transient analysis for all bounding events is provided for each fuel cycle, including the first cycle of TPO operation. Appendix E gives the pertinent generic bases, methods, and assumptions that show that for incremental change in transient performance for TPO uprate, up to a maximum of 1.5 percent, does not significantly change the required fuel operating limits.

The Operating Limit Minimum Critical Power Ratio (OLMCPR) changes for 1.64 percent uprate may appear to be slightly larger than other evaluations for transient minimum critical power ratio (MCPR) events, but these changes are to be expected within the normal cycle-to-cycle variation.

Limiting events will be reanalyzed for TPO uprate for normal reload preparation for the first fuel cycle. This analysis observes all events that establish the core thermal operating limits and events that show bounding conformance to other transient protection criteria. The OLMCPR changes for 1.64 per~ent uprate may appear to be slightly larger than other evaluations for transient MCPR events, but these changes are to be expected within the normal cycle-to-cycle variation as previously stated. Overpressure events and a loss of FW transient are currently performed assuming 2 percent overpower. Thus they are applicable and bounding for the TPO up rate.

The recirculation flow increase events and the flow-dependent operating limits due to TPO uprate experiences no significant change due to the maximum operating boundary at partial core flow not being increased. The upper limit on the core flow runout is not being changed; thus the maximum power increase due to increased flow is still bounded by the current analysis and is acceptable.

Tl)e main steam line isolation valve (MSIV) closure for high neutron flux SCRAM (reactor trip) will be analyzed for the first TPO reload, which is typical with current reload analysis practice, because it has beeri determined to be the worst overpressure evaluation event. This new .

analysis will be performed from an initial power level consistent with the uncertainty for FW instrumentation capability defined by the utility for the new MUR uncertainty. The overpressure case analysis of the first TPO reload is acceptable due to the current analysis included assuming initial 2 percent overpower, and the existing case bounds th.e TPO uprate.

Loss of FW flow transient analysis is not required for a TPO uprate. Previous safety analyses for this specific event that show coolant is maintained above the top of active fuel include conservatisms like the 102 percent of CLTP. This analysis bounds reactor performance at TPO operating conditions due to the new analysis conditions are the same as current analysis.

The evaluations and conclusions presented in TL TR NED0-32983 Revision 1 Appendix E are applicable for the Fermi 2 TPO uprate being an NRC-approved method and Fermi 2 falls under the envelope for the approval. It is acceptable for the plant to perform standard reload analyses for the first fuel cycle that implements the TPO uprate.

Design Basis Accidents (DBAs) Analysis The NRC staff's review of DBAs consider radiological consequences that are basically proportional to the quantity of radioactivity released to the environment. For TPO uprate, the radiological releases are expected to increase in proportion to the core inventory increases, which will be in proportion to the power increase. Postulated DBA events are analyzed to ensure that NRC regulations are met during operation at 102 percent above the CLTP. DBA events are either not dependent on core thermal power or have already been previously analyzed at 102 percent of CLTP. These evaluations and analyses are based on the m.ethodology, assumptions, and analytical techniques that are described in applicable RGs, the SRP, and previous SERs. Looking at the DBA LOCA, the small increase in the post-accident radiation levels has no significant effect on the plant or on-site Emergen<::y Response facilities.

The staff's review of the areas requiring post-accident occupancy concluded that needed access to these areas for accident mitigation is not significantly affected by the TPO uprate.

Since the operating pressures and temperatures change only slightly for the TPO uprate, these changes remain within acceptance limits for High Energy Line Break (HELB) mass and energy releases. These original HELB analyses remain adequate and bounding.

Anticipated Transient Without SCRAM (ATWS) Analysis The TLTR in Sections 5.3.5 and Appendix L of NED0-32983, Revision 1 present a generic evaluation for the sensitivity of an ATWS to a change in power typical for a TPO uprate. The methodology used to analyze and evaluate the Fermi 2 ATWS event, NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate [CPPUs]," was approved by the NRC as an acceptable methodology for evaluating the effects of CPPUs. NEDC-33004P-A is also commonly referred. to as the "CLTR." The NRC-approved code ODYN is used for TPO RTP A TWS analysis that ensures the following ATWS acceptance criteria are met: Maintain reactor vessel integrity (i.e., peak vessel bottom head pressure less than the ASME Service Level C limit of 1500 psig), maintain containment integrity (i.e., maximum containment pressure and temperature less than the limiting pressure (56 psig) and temperature (198 °F) of the containment structure), and maintain coolable core geometry. The ATWS analysis results meet the ATWS acceptance criteria. Therefore, the NRC staff concludes that the Fermi 2 ATWS event analysis at TPO is acceptable.

Conclusion for Accident Analyses The NRC staff reviewed the reactor systems and thermal-hydraulic aspects of the LARin support of implementation of an MUR power uprate. Based on the considerations discussed above, the NRC staff determined that the results of the licensee's analyses related to these areas continue to meet applicable acceptance criteria following implementation of the MUR.

Most of the current analyses of record are based on operation at CLTP of 3430 MWt and 102 percent of CLTP, which includes 2.0 percent measurement uncertainty. The proposed amendment is based on the use of a Cameron LEFM CheckPius system that would decrease the uncertainty in the FW flow, thereby decreasing the power level measurement uncertainty from 2.0 percent to 0.36 percent. In these cases, the proposed MUR rated thermal power of 3486 MWt is bounded by the current analyses of record. Therefore, the NRC staff finds the proposed license amendment acceptable with respect to the accident analyses.

3.9 Reactor Pressure Vessel (RPV) and Internals Integrity The NRC staff's review in the area of RPV integrity focuses on the impact of the proposed MUR on the licensee's reactor coolant system (RCS) pressure-temperature (PIT) limits, RPV upper shelf energy (USE) assessment, and RPV surveillance capsule withdrawal schedule, considering the increase in RPV neutron fluence associated with MUR conditions. The NRC staff's review was conducted in accordance with the guidance contained in RIS 2002-03 to verify that the results of the licensee's analyses of these parameters continue to meet the requirements of 10 CFR 50.60, "Acceptance Criteria for Fracture Prevention Measures for Lightwater Nuclear Power Reactors for Normal Operation," 10 CFR Part 50, Appendix G, "Fracture Toughness Requirements," and 10 CFR Part 50, Appendix H, "Reactor Vessel*

Material Surveillance Program Requirements," following implementation of the proposed MUR.

The NRC staff's review in the area of reactor internals integrity focuses on the impact of the proposed MUR on the on the integrity of the reactor internals and core support structure components, taking into consideration the ability of these components -to maintain their safety functions under MUR conditions.

3.9.1 P/T Limits and Upper Shelf Energy Regulatory Evaluation Appendix G of 10 CFR Part 50 provides fracture toughness requirements for ferritic materials (low alloy steel or carbon steel) in the reactor coolant pressure boundary (RCPB), including USE requirements for the RPV beltline region and requirements for calculating PIT limits for the RCS. USE values are used for assessing the safety margins of the RPV beltline materials against ductile tearing. 10 CFR Part 50, Appendix G requires that the RPV beltline materials maintain USE values of no less than 50 ft-lbs throughout the operating life of the RPV, unless it is demonstrated that lower values of USE will provide margins of safety against fracture equivalent to those required by the ASME Code,Section XI, Appendix G. USE values for the RPV beltline materials must account for the effects of neutron radiation, as required by 10 CFR Part 50, Appendix G. The NRC staff's review of the licensee's USE assessment addressed the impact of the proposed MUR on the neutron fluence and USE values for the RPV beltline materials through the end of the current licensed operating period (end-of-license, or EOL),

which corresponds to 32 effective full power years (EFPY) of facility operation.

. PIT limits are established to ensure the structural integrity of the ferritic components of the RCPB against non-ductile fracture during any condition df normal operation, including anticipated operational occurrences and hydrostatic tests. 10 CFR Part 50, Appendix G requires that the PIT limits be at least as conservative as those that would be generated using the methods of the ASME* Code,Section XI, Appendix G. 10 CFR Part 50, Appendix G also requires that applicable surveillance data from RPV material surveillance programs be incorporated into the calculations of plant-specific PIT limits and that the P/T limits for operating reactors be generated using a method that accounts for the effects of neutron radiation on the material properties of the RPV beltline materials. The NRC staff's review of the licensee's PIT limits covered. the licensee's PIT limits methodology and the current technical specification PIT limit curves for 24 EFPY and 32 EFPY, considering neutron fluence and neutron embrittlement effects on the RPV beltline materials under MUR conditions.

The NRC staff's regulatory guidance related to the evaluation of neutron embrittlement for USE evaluations and PIT limit curves is found in RG 1.99, Rev. 2, "Radiation Embrittlerrient of Reactor Vessel Materials." RG 1.99, Rev. 2 specifies methods for determining the projected decrease in USE and adjustments to the reference nil-ductility temperature (RT NoT) for the RPV beltline materials as a result of neutron embrittlement, taking into consideration credible surveillance data as appropriate. The projected decrease in USE is used to determine the USE at EOL. The adjusted RTNoT (ART) due to neutron embrittlement is*the critical material property for calculating PIT limits for the RPV beltline region.

Technical Evaluation The licensee's evaluation of the impact of the proposed MUR on the USE and PIT limits is addressed in Section 3.2 of the MUR safety analysis report (SAR), which was provided as an enclosure to the* February 7, 2013, LAR. The licensee stated in the SAR that the Fermi 2 RPV was evaluated at a power level corresponding to 115 percent of CLTP, which bounds the MUR condition of 101.64 percent of CLTP. The 'staff confirmed that the RPV was evaluated against.

the requirements of 10 CFR Part 50, Appendix G to demonstrate that these regulatory criteria would continue to be satisfied at 115 percent of CLTP through EOl. (32 EFPY).

The 32 EFPY ART values corresponding to 115 percent of CLTP, along with the critical input parameters- specifically, the material properties and neutron fluence values, are provided in Table 3-1 of the SAR. The staff reviewed the 32 EFPY ART values and determined that they were accurately determined in accordance with the procedures in RG 1.99, Rev. 2. For the limiting RPV beltline shell material, the staff verified that the licensee incorporated surveillance data from the representative Boiling Water Reactor (BWR) Vessel and Internals Project (BWRVIP) Integrated Surveillance Program (ISP) material to establish the limiting ART value, in accordance with the RG 1.99, Rev. 2, Position 2.1 guidelines. The staff confirmed that all other material property inputs for the ART calculations correspond to those established in previous docketed and staff-approved submittals for the TS PIT limit curves. Therefore, the NRC staff finds the licensee's 32 EFPY ART values acceptable for MUR conditions.

The current PIT limits for Fermi 2 are established in the plant's TSs for 24 EFPY and 32 EFPY.

These PIT limits were developed based on the methods and calculations documented in the GE technical report, NED0-33133, "Pressure-Temperature Curves for DTE Energy Fermi Unit 2,"

Rev. 0, February 2005 (ADAMS Accession No. ML050870587), which was provided in an enclosure to the March 17, 2005, LAR to implement the curves in the Fermi 2 TS. By letter dated January 25, 2006, the NRC staff issued Amendment No. 168 (ADAMS Accession No.

ML053120186) to the Fermi 2 operating license, approving the implementation of these curves in the Fermi 2- TS for 24 EFPY and 32 EFPY.

The NRC staff noted that the current TS PIT limits for Fermi 2 are based on the analysis of the limiting RPV beltline shell material. Excluding the minimum temperature criteria (which were correctly established based on the RT NDT for the RPV flange region per Table 1 of 10 CFR Part 50, Appendix G), the PIT limits were generated using the 24 EFPY and 32 EFPY ART values for the limiting beltline shell material at the one-quarter RPV wall thickness (1/4T) and three-quarters RPV wall thickness (3/4T) locations, as specified in the ASME Code,Section XI, Appendix G. Additionally, as documented in NED0-33133, the development of these curves also took into consideration the non-beltline regions of the RPV, the upper RPV region, and the RPV lower head. For the non-beltline regions of the RPV, the licensee generated PIT limits for the bounding structural discontinuities based on the limiting stress concentration effects associated with these components. The PIT limits for the non-beltline components were found to be less restrictive than those for the limiting beltline shell material.

Section 3.2 of the SAR states that the RPV water level instrument nozzles are bounded by the current TS PIT limit curves only up to 21 EFPY and that this issue affects the operation of Fermi 2, regardless of the MUR implementation. Section 3.2 of the SAR also indicates that resolution of this issue is being pursued in a separate LAR. The NRC staff noted that the analyses documented in NED0-33133 did not address the RPV beltline water level instrument nozzles for the generation of the 24 EFPY and 32 EFPY TS P/T limit curves. 10 CFR Part 50, Appendix G requires that PIT limits be developed by considering all beltline and non-beltline ferritic RCPB components, in particular RPV nozzles, penetrations, and other discontinuities that exhibit higher stresses than the RPV beltline shell region, and which could result in more restrictive PIT limits, even if the RT NDT for these components is not as high as that of the limiting .RPV beltline shell material. Therefore, the PIT limit curves must bound the water level instrument nozzles in order for the curves to remain in compliance with 10 CFR Part 50, Appendix G beyond 21 EFPY.

By letter dated December 21, 2012 (ADAMS Accession No. ML13004A134), the licensee

submitted an LAR to implement a PIT limits report (PTLR) in accordance with criteria established in GL 96-03, "Relocation of the Pressure-Temperature Limit Curves and Low Temperature Overpressure Protection System Limits." This LAR addresses the necessary PIT limit curve revision for ensuring that the curves bound the water level instrument nozzles, as referenced in Section 3.2 of the MUR SAR. The proposed PTLR contains new PIT limit curves for 24 EFPY and 32 EFPY. The NRC staff's review of the new P/T limit curves confirmed that these curves are bounding for all ferritic RCPB components, including the water level instrument *

. nozzles. Additionally, the new PIT curves continue to remain bounding for the non-beltline regions of the RPV since there is no significant neutron embrittlement affecting the component-specific limits for these non-beltline components. For the bounding RPV components, which include the limiting beltline shell material and the water level instrument nozzles, the proposed 32 EFPY PIT limit curves were generated using 32 EFPY ART values that correspond with those listed in Table 3-1 of the SAR. Therefore, since the curves are calculated based on neutron fluence projections and corresponding ART values for 115 percent of CLTP, which bounds plant operation for the proposed MUR (1 01.64 percent of CLTP), the NRC staff finds that the PIT limit curves included in the proposed PTLR would be acceptable for MUR conditions. The staff's detailed review and findings regarding the acceptability of the proposed PTLR and PIT limit curves, relative to criteria of GL 96-03 and the requirements of 10 CFR Part 50, Appendix G, are documented in the staff's SE associated with Amendment No. 195, dated February 4, 2014 (ADAMS Accession No. ML133468067).

Regarding the USE evaluation, Section 3.2 to the SAR states that the projected USE remains greater than 50 ft-lbs, thereby demonstrating compliance with the USE requirements of 10 CFR Part 50, Appendix G. Table 3-2 of the SAR provides the projected EOL (32 EFPY) USE values, supporting calculations, and material property inputs for the RPV beltline materials. The staff independently confirmed that all USE values are projected to remain greater than 50 ft-lbs, as required by 10 CFR Part 50, Appendix G. As with the licensee's 32 EFPY ART projections, the USE projections provided in SAR Table 3~2 are based on projected neutron fluence inputs corresponding to 115 percent of CLTP, which conservatively bounds plant operation under the proposed MUR conditions of 101.64 percent of CLTP. Therefore, the 32 EFPY USE projections are acceptable for the proposed MUR.

The NRC staff has reviewed the licensee's proposed LAR to increase the rated core thermal power for Fermi 2 by 1.64 percent and has evaluated the impact that the MUR conditions will have on the integrity of the RPV, relative to the USE and PIT limit requirements of 10 CFR Part 50, Appendix G. The NRC staff has determined that the changes associated with MUR conditions will not impact the safety functions or safety margins required for the continued safe operation of the RPV, and Fermi 2 will continue to remain in compliance with the requirements of 10 CFR Part 50, Appendix G through EOL (32 EFPY) for MUR conditions.

3.9.2 RPV Material Surveillance Program Regulatory Evaluation The RPV material surveillance program provides a means for monitoring the fracture toughness of the RPV beltline materials to support analyses for ensuring the structural integrity of the ferritic components of the RPV beltline region. 10 CFR Part 50, Appendix H, identifies the requirements for the design and implementation of the RPV material surveillance program.

I Technical Evaluation The licensee discussed the impact of the proposed MUR on the RPV material surveillance program in Section 3.2 of the SAR. The licensee stated that the surveillance program at Fermi 2 originally consisted of three plant-specific surveillance capsules, two of which are still in the RPV. The licensee also stated that Fermi 2 participates in the BWRVIP ISP and will continue to comply with the requirements of that program.

The NRC staff verified that there are sufficient representative surveillance capsules available for Fermi 2 under the ISP to meet the requirements of 10 CFR Part 50, Appendix H for 32 EFPY and beyond. The staff also confirmed that the proposed MUR will not affect the validity of the current ISP surveillance capsule withdrawal schedule for Fermi 2, relative to the requirements of 10 CFR Part 50, Appendix H. Therefore, Fermi 2, through its participation in the BWRVIP ISP, will remain in compliance with the requirements of 10 CFR Part 50, Appendix H, following implementation of the proposed MUR.

The NRC staff has reviewed the licensee's proposed LAR to increase the rated core thermal power for Fermi 2 by 1.64 percent and has evaluated the impact that the MUR conditions will have on the RPV material surveillance program, relative to the requirements of 10 CFR Part 50, Appendix H. The NRC staff has determined that Fermi 2, through its participation in the BWRVIP ISP, will continue to remain in compliance with the requirements of 10 CFR Part 50, Appendix H through EOL (32 EFPY) for MUR conditions.

3.9.3 RPV Internals and Core Support Materials Regulatory Evaluation The RPV internals and core support structures include structures, systems, and components (SSCs) that perform safety functions or whose failure could affect safety functions performed by other SSCs. These safety functions include reactivity monitoring and control, core cooling, and fission product confinement (within both the fuel cladding and the reactor coolant pressure boundary). The NRC's acceptance criteria for RPV internals and core support materials are based on GDC-1 and 10 CFR 50.55a for material specifications, controls on welding, and inspection of RV internals and core supports.*

Technical Evaluation The licensee discussed the impact of the proposed MUR on the structural integrity of the RPV internal and core support structure components in Section 3.3 of the SAR. The staff's evaluation of the licensee's analysis of the impact of the MUR on the loading of the RPV internals and core support structures is provided in Section 3.6.2.3 of this SE, wherein the NRC staff determined that the relatively minor chang.es in loads on the RPV internals associated with the MUR will not result in any of these components exceeding its design basis acceptance criteria for applied stress.

In addition to the applied loading and stress analysis results discussed in Section 3.3 of the SAR, Fermi 2 conducts inspections of the RPV internals and core support structures in accordance with the BWRVIP guidelines. The licensee's use of the BWRVIP guidelines for the internals was formally established for the third 10-year interval in service inspection program at Fermi 2 based on the NRC staff's authorization of an ~lternative, pursuant to 10 CFR 50.55a(a)(3)(i), to implement the BWRVIP guidelines in lieu of the ASME Code,Section XI examination requirements for inservice inspection of !he RPV internals and core support structures. The staff's basis for authorizing this alternative is documented in a February 17, 2012, SE (ADAMS Accession No. ML120370286). The NRC staff's approval of the licensee's use of the BWRVIP guidelines in lieu of the ASME Code,Section XI examination requirements is based on its finding that the BWRVIP guidelines provide more comprehensive inser\tice examination coverage requirements for the RPV internals and core support structures than those specified in the ASME Code,Section XI, and accordingly, the licensee's alternative was found to provide an acceptable level of quality and safety, per 10 CFR 50.55a(a)(3)(i). All RPV internal and core support structure components that are within the scope of evaluation for MUR conditions are inspected in accordance with the guidelines provided in the applicable BWRVIP document for that component. The BWRVIP documents covering inspection and evaluation of RPV internal components are referenced iri the staff's February 17, 2012, safety evaluation.

The NRC staff considers that the combination of relatively small changes made in implementing the MUR, in concert with the licensee's performance of inspections in accordance with the staff-approved BWRVIP guidelines, provide assurance for the continuing safety-related function of the RPV internal components.

The NRC staff has reviewed the licensee's proposed LAR to increase the rated core thermal power for Fermi 2 by 1.64 percent and has evaluated the impact that the MUR conditions will have on the structural integrity assessments for the RPV internals and core support structures.

The staff has determined that the changes associated with MUR conditions will not impact the

safety functions or safety margins required for the RPV internal and core support structure components.

3.10 Electrical Systems 3.1 0.1 Regulatory Evaluation The regulatory requirements which the NRC staff applied in its review of the application include:

GDC 17, "Electric power systems," of 10 CFR 50, Appendix A requires that an onsite power system and an offsite electrical power system be provided with sufficient capacity and capability to permit functioning of structures, systems, and components important to safety.

10 CFR 50.63, "Loss of all alternating current (AC) power," requires, in part, that all nuclear plants have the capability to withstand a loss of all AC power (station blackout) for an established period of time, and to recover therefrom.

10 CFR 50.49, "Environmental Qualification (EQ) of Electric Equipment Important to Safety for Nuclear Power Plants," requires, in part, that licensees establish programs to qualify electric equipment important to safety, located in harsh environment.

3.1 0.2 Technical Evaluation The licensee developed the license amendment request (LAR) consistent with the guidelines in .

NRC RIS 2002-03, "Guidance on the Content of Measurement Uncertainty Recapture (MUR)

Uprate Applications."

In the LAR, Enclosure 7, the licensee provided a report by GE-Hitachi (NEDC-33578P, Revision 0), which summarizes the results of all significant safety evaluations performed that justify increasing the licensed thermal power at Fermi 2. This document addresses a Thermal Power Optimization (TPO) power uprate of 1.64 percent of the Current Licensed Thermal Power (CLTP), consistent with the MUR uprate for Fermi 2, an increase in licensed thermal power from 3,430 MWt' to 3,486 MWt, and an expected increase in the electrical power of approxil!lately 22 megawatts-electric (MWe). [Note: TPO power uprate is synonymous with MUR power uprate for the purpose of this license amendment - both representing an increase in MWt of 1.64 percent].

The NRC staff reviewed the licensee's evaluation of the impact of the MUR power uprate on following electrical systems/components:

AC Distribution System

Power Block Equipment (Main Generator, Power Transformers, Isolated-phase bus duct)

Direct Current (DC) system

Emergency Diesel Generators (EDGs)

Switchyard

Grid Stability

Station Blackout (SBO)

- EQ Program

AC Distribution System The onsite AC Distribution System is the source of power for the non-safety.:related buses and for the safety-related emergency buses. It consists of the 4.16 kV, 480 V, and 120 V systems (not including the EDGs). In the LAR, Enclosure 7, Section 6.1.2, the licensee stated that the electrical loads that will be affected by the TPO power uprate are the condenser pump, heater feed pumps, and heater drain pumps. The brake horsepower (BHP) of each of these pumps will slightly increase (0.47 percent to 1.03 percent) or decrease (0.17 percent). The licensee stated that the increase in these loads will not exceed their motor nameplate ratings and the electrical supply and distribution components will have sufficient capacity. The NRC staff requested the licensee to provide the nameplate ratings and the BHP loads for these pumps. In response to the request for additional information (RAI), the licensee in the letter dated July 15, 2013, provided nameplate ratings and loadings at TPO conditions as follows: condenser pump 1500 horsepower (HP) nameplate, loading 1186 BHP; heater feed pump 3000 HP nameplate, loading 2625 BHP; and heater drain pump 1750 HP nameplate, loading 1645 BHP. The loadings remain within the pump motor nameplate ratings.

The NRC staff reviewed the LAR and the licensee's RAI response and concludes that the onsite AC distribution system will experience only minor load changes, and will continue to have adequate capacity to support the plant loading for TPO/MUR power uprate conditions.

Power Block Equipment As a result of the power uprate, the rated thermal power will increase to 3486 MWt from the previously analyzed core power level of 3430 MWt.

In the LAR, Enclosure 7, Tables 6-1 and 6-2, the licensee stated that the main generator is rated at 1350 MegaVolt Amperes (MVA), 1,215 MWe at a 0.90 power factor. At TPO uprate conditions, the maximum generator load is 1,342 MVA that is bounded by the generator design rating of 1350 MVA. In the response to NRC staff's RAI, the licensee in a letter dated July 15, 2013, provided the generator nominal MWe loading at CL TP conditions as 1184.59 MWe, and maximum loading at TPO conditions as 1208.29 MWe (an increase of 23.7 MWe). The NRC staff verified that the increased electrical output (1208.9 MWe) at TPO conditions remains bounded by the design rating ( 1215 MWe) of the generator, and that the main generator will be operating within the existing generator capability curve for TPO uprate. Therefore, the NRC staff finds that the generator is capable of operation at TPO uprate conditions.

In response to the NRC staff's RAI on protective relay settings, the licensee stated that the TPO power uprate has no impact on protective relaying design. The protective settings are based upon equipment nameplate ratings which remain unchanged for TPO conditions. The NRC staff reviewed the licensee's response, in regards to protective relay settings, and finds it acceptable.

In the LAR, Enclosure 7, Section 6.1.1, the licensee stated that the isolated phase main bus connected to the main generator output, is rated at 37,000 amperes (A) and that the transformer bus subsections, connected to the main generator step-up transformers (#2A and #28), are rated at 18,500A. In response to the NRC staff's RAI, the licensee provided further data for the main bus and transformer bus subsections. At CLTP conditions the maximum electrical load through the main bus subsection is 35,1 08A, while at TPO uprate conditions it is 35,804A. At CL TP conditions the maximum electrical load through the transformers 28 and 2A bus subsections is 18,000A and 17,108A respectively while at TPO uprate conditions it is 18,357A and 17,447A respectively. The maximum loading for both the main bus and transformer bus

subsections are lower than their ratings of 37,000A and 18,500A respectively. Therefore, the NRC staff finds that the isolated phase bus is capable of operation at TPO up rate conditions.

According to the updated final safety analysis report (UFSAR) Figure 8.3-1, the main generator feeds electric power through a 22 kilovolt (kV) isolated phase bus to two parallel step-up transformers (#2A and #2B), stepping the. generator voltage of 22 kV up to the transmission voltage of 345 kV. Main generator step-up transformers 2A and 2B are rated at 710 MVA and 874 MVA respectively. In response to the NRC staff's RAI, the licensee stated that both transformers ratings bound their respective loadings at TPO conditions, 654.18 MVA for transformer 2A and 688.32 MVA for transformer 2B. Based on the above, the NRC staff finds that the main generator step-up transformers are capable of operation at TPO uprate conditio~s.

Plant auxiliary power is provided through the station service transformers #64 and #65. The loadings of station service transformers are provided in LAR, Enclosure 6, Tables 6-4a and 6-4b. During normal operating conditions the station service transformers power the 4.16 kV switchgear, 480 V load centers, and motor control centers. Transformer #64 is rated at 13.2/4.16 kV, 15/20 MVA. It supplies Division I power requirements and it receives its power via transformer #1, rated at 120/13.2 kV, 24/32 MVA from120 kV switchyard. Its loading at TPO uprate condition is 15.10 MVA and is bounded by its maximum rating(20MVA). Transformer

65 is rated at 345/4.16i4.16 kV, 28/37.3 MVA. It supplies Division II power requirements and it receives its power from the 345 kV switchyard. Its loading at TPO uprate condition is 35.60 MVA and is bounded by its maximum rating (37.3 MVA). Therefore, the NRC staff finds that the station service transformers are capable of operation at TPO uprate conditions.

Based on the review of information provided in the LAR and the supplemental information, the NRC staff finds that the power block equipment will remain adequate to support the plant loading for TPO/MUR power uprate conditions.

DC System According to the UFSAR, Section 8.3.2, the DC power system consists of two independent Class 1E battery systems, one system per division. Each system supplies DC power at 260 V DC and 130 V DC. There is also a 260/130 V DC BOP [balance of plant] system serving BOP loads. Further, each high-voltage switchyard has its own independent source of DC power for circuit breaker control. There are. two batteries and chargers in the 345 kV switchyard and one battery and charger in the 120 kV switchyard.

In the LAR, Enclosure 7, Section 6.2, the licensee stated that the DC loading requirements in

the UFSAR and station load calculations were reviewed and no reactor power dependent loads were identified for the DC system. The plant operation at the TPO RTP [rated thermal power]

level will not increase any DC loads or control logic. The NRC staff finds that the licensee's assessment that the DC system will not be impacted by TPO uprate conditions is reasonable and acceptable.

According to UFSAR, Section 8.3.1.1.8, the standby emergency AC power source for Fermi 2 consists of four diesel generators, two per division. Each EDG is of sufficient capacity to carry the essential load of its respective bus.

In the LAR, Enclosure 7, Section 6.1.2, the licensee stated that there are no load changes to the ECCS, and emergency operation at the TPO uprate level is achieved by utilizing existing equipment operating at or below the nameplate rating. Thus, the amount of power required to perform safety-related functions does not increase, and the EDG system has adequate capacity and capability to power the safety-related loads at TPO uprate conditions. Also, the duty cycle and duration for design basis EDG loads are based on analytical power levels of at least 102 percent of the CLTP. The NRC staff finds the licensee's assessment that the EDG system will not be adversely impacted by TPO uprate conditions is reasonable and acceptable.

Switch yard According to the UFSAR, Section 8.2, and Figure 8.3-1, the Fermi switchyard system is composed of two independent and physically separated switchyards. They serve two 345 kV lines and three 120 kV line. The 345 kV switchyard is arranged in a nominal double breaker-double bus configuration and is connected to the two main generator transformers. The 120 kV switchyard is arranged with two buses tied together by a normally closed circuit breaker.

Auxiliary power for Fermi .2 comes from both the 345 kV and 120 kV systems. Each system supplies loads of one of the two redundant divisions. The primary function of the switchyard and distribution system is to connect the station electr,ical system to the transmission grid.

In the LAR, Enclosure 7, Section 6.1.1, the licensee stated that the main generator transformers and the associated switchyard components are rated for maximum generator output and hence adequate for. the TPO uprate output. In response to the NRC staff's RAI, the licensee in a letter dated July 15, 2013, provided loading/rating analysis of circuit breakers, disconnect switches, current transformers, and overhead conductors for both 345 kV switchyard and 120 kV switchyard. The NRC staff reviewed the response arid verified that the ratings of switchyard components remain adequate for the TPO/MUR power up rate conditions.

Grid Stability In the LAR, Enclosure 1, Section 3.4.5, the licensee stated that a grid adequacy study was performed to evaluate Fermi 2 MUR power uprate condition, which was based on the methodology for periodic (annual) grid studies specified in the Fermi 2. Nuclear Plant Operating Agreement (NPOA). The operation of the grid and Fermi 2 offsite power system is governed by a NPOA between DTE Energy, the International Transmission Company and Midwest Independent Transmission Operator. The study was performed at th~ main generator's nameplate rating (1215 MWe) output, which bounds the TPO uprate. The study inCluded steady-state, transient and fault analysis of the transmission/grid system. The results showed that the Fermi 2 and the rest of the transmission system will remain stable for all conditions studied. The grid is capable of supplying the required off site power if the plant trips off line, which will allow safe shutdown during a Design Basis Accident condition.

Based on the above grid stability results, the NRC staff finds that the Fermi 2 TPO/MUR power uprate would not adversely impact the stability of the transmission/grid system.

According to UFSAR, Section 8.4.2.1, Fermi's SBO coping duration is four hours. This is based on the licensee's evaluation of the offsite power design characteristics, emergency AC power system configuration, and EDG reliability, in accordance with the evaluation procedure outlined in NUMARC 87-00 and Regulatory Guide 1.155.

In the LAR, Enclosure 7, Section 9.3.2, the licensee made a reference to the NRC-approved licensing topical report (TLTR) NEDC-32938P-A, "Generic Guidelines and Evaluations for General Electric Boiling Water Reactor Thermal Power Optimization." The licensee stated the TLTR Appendix L provides a generic evaluation of the SBO event for the plant response and coping *capability analyses for typical power up rate projects. This evaluation is for confirmation of continued compliance to 10 CFR 50.63. The generic evaluation is based on the following main considerations: the adequacy of the condensate/reactor coolant inventory, the capacity of the Class 1E batteries, the SBO compressed Nitrogen requirements, the ability to maintain containment integrity, and the effect of loss of ventilation on rooms that contain equipment essential for plant response to a SBO event. According to Section L.5.1 of the TL TR, the plants that have analyzed a SBO event at nominal CLTP level can assess the effect of TPO uprate from existing sensitivity analysis provided in Table L-3 of TLTR. If the TPO plant shows that it currently has sufficient margin, no plant-specific SBO analysis need be performed. The licensee stated that Fermi 2 currently has margin of 37,438 gallons to the available condensate storage inventory volume and 29°F to the containment peak temperature limit. These margins are well in excess of the margins required per Table L-3 of TL TR [NEDC-32938P-A].

The NRC staff agrees with the licensee's assessment that, based on the sufficient margins, no further Fermi 2 plant-specific SBO analysis is required. Based on this information, the NRC staff finds that the TPO uprate will have minimal impact on Fermi's SBO coping duration.

Therefore, the staff finds that Fermi 2 will continue to meet the requirements of 10 CFR 50.63 under TPO/MUR power uprate conditions.

Environmental Qualification (EQ) Program In the LAR, Enclosure 7, Section 10.3, the licensee stated that the environmental conditions for safety-related electrical equipment were reviewed to ensure that the existing qualification for the normal. and accidental conditions for the area where the devices are located remain adequate.*

Inside the containment at the TPO uprate accident conditions, the temperature and pressure values remain bounded by the environmental qualification (EQ) analyses performed at ~102 percent CL TP. Outside the containment, the HELB pressure and temperature profiles bound the TPO uprate conditions. There is adequate margin in the qualification envelopes to accommodate the small changes to TPO conditions. In response to the staff's RAI, the licensee provided the following details in a letter dated July 15, 2013:

Inside containment, the containment parameters for the CLTP evaluation of temperature and pressure conditions for post-LOCA response assumed a 3,499 MWt which bounds the TPO uprate value of 3,486 MWt. Outside containment, a bounding system pressure of 1060 pounds per square inch absolute (psi a) is used in the CLTP calculations which evaluate the temperature and pressure impacts associated with the Main Steam line break (MSLB) and other line breaks for HPCI [high pressure coolant injection], RCIC

[reactor core isolation cooling], and RWCU [reactor water cleanup].

The licensee stated that for normal non-accident conditions, the heating, ventilation, and air conditioning systems were found to be capable of maintaining environmental conditions within their normal ranges of fluctuations.

Regarding radiation levels inside and outside containment, the licensee stated that a review of EO program radiation dose bases for all EQ zones, was performed at a core thermal power of 3,952 MWt, which bounds the TPO uprate value of 3,486 MWt.

~ 60-Based on above information, the NRC staff finds that the current EQ parameters will remain bounding for the TPO power uprate conditions. Therefore, the NRC staff finds that Fermi 2 will continue to meet the requirements of 10 CFR 50.49 under TPO/MUR power uprate conditions.

3.1 0.3 Electrical Systems Conclusion Based on the technical evaluation provided above, the NRC staff finds that Fermi 2 will continue to meet the requirements of GDC 17, 10 CFR 50.63, and 10 CFR 50.49. Therefore, the NRC r staff finds the TPO/MUR power uprate acceptable.

3.11 Instrumentation and Controls 3.11.1 Introduction This MUR power uprate is based on the use of the FW flow measurement techniques of a Cameron (formerly Caldon) Leading Edge.Fiow Meter (LEFM) CheckPius' System. The LAR references the following topical reports: ER-80P and ER-157P, Rev. 8, and their respective SEs dated March 8, 1999, and August 16, 2010.

These two topical reports, which are generically applicable to nuclear power plants, document the ability of the Cameron LEFM CheckPius Systems to increase the accuracy of flow measurement. ER-80P describes the LEFM technology, includes calculations of power measurement uncertainty using a Cameron LEFM Check System in a typical two-loop pressurized-water reactor (PWR) or two-FW-Iine boiling-water reactor (BWR), and provides guidelines and equations for determining the plant-specific power calorimetric uncertainties.

ER-157P, Rev. 8, and supplements describe the Cameron LEFM CheckPius System and list the results of a typical PWR or BWR thermal power measurement uncertainty calculation using the Cameron LEFM CheckPius System. Together, these two topical reports, along with theSEs approving them, provide the generic safety basis for an MUR power uprate.

The plant-specific bases for the proposed MUR uprate at Fermi 2 are described in more detail in proprietary Enclosures 10 and 11 to the LAR.

3.11.2 Regulatory Evaluation Nuclear power plants are licensed to operate at a specified core thermal power. The regulation at 10 CFR 50, Appendix K, requires loss-of-coolant accident (LOCA) and emergency core cooling system (ECCS) analyses to assume "that the reactor has been operating continuously at a power level at least 1.02 times the licensed thermal power level to allow for instrumentation error ... " Alternatively, *Appendix K allows such analyses to assume a value lower th~n the 102 percent, but not less than the CLTP, "provided the proposed alternative value has been demonstrated to account for uncertainties due to power level instrumentation error." This allowance gives licensees the option of justifying a power uprate with reduced margin between the CLTP and the power level assumed in the ECCS analysis by using more accurate instrumentation to calculate the reactor thermal power.

As the maximum power level of a nuclear plant is a licensed limit, the NRC must review and approve a proposal to raise the licensed power level under the license amendment process.

The LAR should include a justification for the reduced power measurement uncertainty to support the proposed power uprate.

ER-80P and ER-157P, Rev. 8, describe the Cameron LEFM CheckPius System for the

. measurement of FW flow and provide a generic basis for the proposed power uprate. In its review, the NRC staff also considered the guidance contained in RIS 2002-_03, "Guidance on the Content of Measurement Uncertainty Recapture Power Uprate Applications," dated January 31,

_2002.

The NRC regulatory requirements related to the content of the TSs are set forth in 10 CFR Section 50.36, "Technical specifications," which requires that the TSs include limiting safety system settings (LSSSs). This regulation requires, in part, that "Where a limiting safety system setting is specified for a variable on which a safety limit has been placed, the setting must be chosen so that automatic protective action will correct the abnormal situation before a safety limit is exceeded." Accordingly, the limits for instrument channels that initiate protective functions must be included in the TSs.

In accordance with General Design Criterion (GDC) 20, "_Protection System Functions," of Appendix A to 10 CFR Part 50, "General Design Criteria for Nuclear Power Plants," the protection system shall be designed (1) to initiate automatically the operation of appropriate systems including the reactivity control systems, to assure that specified acceptable fuel design limits are not exceeded as a result of anticipated operational occurrences and (2) to sense accident conditions and to initiate the operation of-systems and components important to safety.

Adherence to acceptable fuel design limits is called for in GDC 10, "Reactor Design," these limits are specified in each plant's Core Operating Limits Report (COLR) and maintained as a part of the Administrative TSs.

The guidance contained in RG 1.1 05, Revision 3, "Setpoints for Safety-Related Instrumentation" (ADAMS Accession No. ML993560062), describes a method acceptable to the NRC staff for assuring that setpoints for safety-related instrumentation are initially within and remain within the limits set by the plant's TSs. The method described in RG 1.105 for combining instrument uncertainties can be used for combining the uncertainties associated with calorimetric calculation. This allows licensees to justify a power uprate with reduced margin between the CL TP and the power level assumed in the ECCS analysis by using more accurate instrumentation to calculate the reactor thermal power.

RIS 2006-17, "NRC Staff Position on the Requirements of 10 CFR 50.36, 'Technical Specifications,' Regarding Limiting Safety System Settings During Periodic Testing and Calibration of Instrument Channels" (ADAMS Accession No. ML051810077), discusses issues that could occur during testing of LSSS and which, therefore, may have an adverse effect on equipment operability. RIS 2006-17 also presents an approach, found acceptable to the NRC staff, for addressing issues for use in licensing applications that require prior NRC staff approval. .

3.11.3 Technical Evaluation Neutron flux instrumentation is calibrated to the core thermal power, which is determined by an automatic or manual calculation of the energy balance on the plant nuclear steam supply system. The accuracy of this calculation depends primarily on the accuracy of FW flow and FW net enthalpy measurements. FW flow is the most significant contributor to the core thermal power uncertainty. A more accurate measurement of this parameter will result in a more accurate determination of core thermal power.

FW flow rate is typically measured using a venturi. This device generates a differential pressure

proportional to the FW velocity in the pipe. Because of the need to improve flow instrumentation measurement uncertainty, tlie industry evaluated other flow measurement techniques and found the Cameron LEFM Check and CheckPius UFMs to be a viable alternative.

3.11.3.1 LEFM Technology and Measurement Both the Cameron LEFM Check and LEFM CheckPius systems use transit time methodology to measure fluid velocity. The basis of the transit time methodology for measuring fluid velocity and temperature is that ultrasonic pulses transmitted through a fluid stream travel faster in the direction of the fluid flow than through the opposite flow. The difference in the upstream and downstream traversing times of the ultrasonic pulse is proportional to the fluid velocity in the pipe. The temperature is determined using a correlation between the mean propagation velocity of the ultrasound pulses in the fluid and the fluid pressure.

Both systems use multiple diagonal acoustic paths instead of a single diagonal path, allowing velocities measured along each path to be numerically integrated over the pipe cross-section to determine the average fluid velocity in the pipe. This fluid velocity is multiplied by a velocity profile correction factor, the pipe cross-section area, and the fluid density to determine the FW mass flow rate in the piping. The mean fluid density may be obtained using the measured pressure and the derived mean fluid temperature as an input to a table of thermodynamic properties of water. The velocity profile correction factor is derived from calibration testing of the LEFM in a plant-specific piping model at a calibration laboratory.

The Cameron LEFM CheckPius System, proposed for use at Fermi 2, uses 16 transducers, eight each in two orthogonal planes of the spool piece. In the Cameron LEFM CheckPius System, when the fluid velocity measured by an acoustic path in one plane is averaged with the fluid velocity measured by its companion path in the second plane, the transverse components of the two velocities are canceled and the result reflects only the axial velocity of the fluid. This makes the numerical integration of four pairs of averaged axial velocities and the computation of volumetric flow inherently more accurate than a result obtained using four acoustic paths in a single plane. Also, since there are twice as many acoustic paths and there are two independent clocks to measure the transit times, errors associated with uncertainties in path length and transit time measurements are reduced.

The Fermi 2 Integrated Plant Computer System (IPCS) receives LEFM CheckPius System mass flow and temperature measurements and directly substitutes the LEFM indications for the existing venturi-based flow and the RTD temperature indications in the plant computer. Using these values, the plant computer would then calculate FW enthalpy and thermal power as it does now. As an alternative, the calorimetric power can be manually calculated, using LEFM indications and following a prescribed procedure.

The NRC staff's review in the area of instrumentation and controls covers the proposed plant-specific implementation of the FW flow measurement technique and the power increase gained as a result of implementing this technique, in accordance with the guidelines provided in Section I of Attachment 1 to RIS 2002-03 (evaluated by the NRC staff is Section 3.1 of this SE).

The NRC staff conducted its review to confirm that the licensee's implementation of the proposed FW flow measurement device is consistent with NRC staff-approved Topical Reports ER-80P and ER-157P, Rev. 8, and that the licensee adequately addressed the additional requirements listed in the NRC staff's SE for these topical reports.

The NRC staff also reviewed the power measurement uncertainty calculations to ensure that:

(1) the conservatively proposed uncertainty value correctly accounts for all uncertainties associated with power level instrumentation errors, and (2) the uncertainty calculations meet the relevant requirements of 10 CFR 50, Appendix K, as described in Section 3.11.2 of this SE.

3.11.3.2 Changes to Technical Specifications (TSs), Protection System Settings, and Emergency System SettingsSection VIII of Attachment 1 to RIS 2002-03 provides guidance to licensees concerning changes to the plant's TSs, protection system settings, and/or emergency system settings required to support the power uprate.

Items A through C Items A, B, and C in Section VIII of Attachment 1 to RIS 2002-03 guide licensees in providing a description of the change, identification of analyses affected by and/or supporting the change, and the justification for the change for any analyses that support and/or are affected by the change.

The NRC staff evaluation of the identified instrumentation for the new power level is based on the analytical limits documented by the licensee in its application.

In its LAR, the licensee proposed to make the following changes to TSs:

Required Action E.1 in TS 3.3.1.1, "Reactor Protection System (RPS)

Instrumentation," associated with the applicable modes or other specified conditions for the Turbine Stop Valve- Closure and Turbine Control Valve- Fast Closure trip functions, is revised to change the value from <30% RTP to <29.5%

RTP.

Surveillance Requirement (SR) 3. 3.1. { 16 in TS 3. 3.1.1, "Reactor Protection System (RPS) Instrumentation," associated with the bypass of the Turbine Stop Valve- Closure and Turbine Control Valve- Fast Closure trip functions is revised to change the value from ~30% RTP to ~29.5% RTP.

Surveillance Requirement (SR) 3.3.1.1.20 in TS 3.3.1.1, "Reactor Protection System (RPS) Instrumentation," associated with bypass of the Average Power Range Monitors Oscillation Power Range Monitor (OPRM) Upscale trip function is revised to change the value from ~28% RTP to ~27.5% RTP.

TS Table 3.3.1.1-1, "Reactor Protection System Instrumentation," Function 2.b, Average Power Range Monitors Simulated Thermal Power- Upscale trip function Allowable Values (AVs) are revised as follows:

Allowable Value:

Current: ~ 0.63 0/V-l::.W) + 64.3% RTP and :5 115.5% RTP Proposed: :5 0.62 (W-l::.W) + 63.1% RTP and :s; 115.5% RTP TS Table 3.3.1.1-1, "Reactor Protection System Instrumentation," Function 9, Turbine Stop Valve - Closure, and Function 10, Turbine Control Valve- Fast Closure, applicable

modes or other specified conditions are revised to change the value from ~30% RTP to

~29.5% RTP.

TS 3.4.1, "Recirculation Loops Operating," Limiting Condition for Operation (LCO) 3.4.1 associated with single loop operation is revised to change the thermal power limits from

5 67.2% RTP to :5 66.1% RTP.

The NRC staff reviewed Table 3.3.1.1-1, "Reactor Protection System Instrumentation," Function 2.b, Average Power Range Monitors Simulated Thermal Power- Upscale trip function AVs calculations in Section 9 of Enclosure 14 of the LAR. The NRC staff determined that the licensee followed the NRC staff-approved GEH setpoint methodology, General Electric Report NEDC-31336P, "General Electric Instrument Setpoint Methodology," (proprietary) which was approved by the NRC staff on November 6, 1995. Therefore, the NRC staff concludes that the change is acceptable.

Other setpoint changes that this LAR proposes are intended to maintain the affected TSs at the same absolute value. The calculations related to turbine stop and control valve closure, recirculation loops and oscillation power range monitor (OPRM) upscale are changes to the percent of full power, all of which are aimed at maintaining the TS values equivalent in terms of absolute reactor power. Therefore, the NRC staff finds the proposed changes keep the values at equivalent values relative to the absolute values at the CLTP.

Additionally, the NRC staff reviewed the licensee's calculations for each of the proposed TS changes and found that the licensee maintains positive margin and operates with proper bounds of as-found tolerance and as-left tolerance. Thus, the licensee provides adequate assurance that the control and monitoring of these setpoints are established and maintained in a manner consistent with plant safety function requirements. Therefore, the NRC staff concludes that the licensee's proposed TS changes comply with 10 CFR 50.36 requirements and are acceptable.

3.11.3.2.1 Adoption of TSTF-493 The licensee also proposed to implement the Technical Specifications Task Force (TSTF) traveler TSTF-493, "Clarify Application of Setpoint Methodology for LSSS Functions," Revision 4 (ADAMS Accession No. ML101160026), recommendations for MUR power uprate instrumentation setpoints that changed and that meet the 10 CFR 50.36 requirements for limiting safety system settings (LSSSs).

The proposed change will resolve operability determination issues described in NRC RIS 2006-17, "NRC Staff Position on the Requirements of 10 CFR 50.36, 'Technical Specifications,'

Regarding Limiting Safety System Settings During Periodic Testing and Calibration of Instrument Channels" (ADAMS Accession No. ML051810077), which is associated with potentially non-conservative TSs Allowable Value (AV) LSSSs, calculated using some methods in the industry standard ISA-S67.04-1994 Part 2, "Methodologies for the Determination of Setpoints for Nuclear Safety-Related Instrumentation." The concern is that when these allowable values are used to assess instrument channel performance during testing, non-conservative decisions about the equipment operability may result. In addition, the proposed change will resolve operability determination issues related to solely relying on Allowable Values associated with TS LSSSs to ensure that TS requirements, not plant procedures, will be used for assessing instrument channel operability.

The licensee proposed adding the following two notes toTS Table 3.3.1.1-1, "Reactor Protection System Instrumentation," Function 2.b, Average Power Range Monitors Simulated Thermal Power- Upscale trip function:

Note 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.

Note e: The instrument channel setpoint shall be reset to a value that is within the as-left tolerance around the Nominal Trip Setpoint (NTSP) at the completion of the surveillance; otherwise, the channel shall be declared inoperable. Setpoints more conservative than the NTSP are acceptable provided that the as-found and as-left tolerances apply to the actual setpoint implemented in the surveillance procedures (field setting) to, confirm channel performance. The NTSP and the methodologies used to determine the as-found and as-left tolerances are specified in the Technical Requirements Manual.

I 3.11.3.2.1.1 Regulatory Evaluation The Commission's regulatory requirements related to the content of the TS are contained in 10 CFR 50.36. The regulation requires, in part, that the TS include limiting safety systems settings.

Section 50.36(c)(1)(ii)(A) states, in part:

Limiting safety system settings for nuclear reactors are settings for automatic protective devices related to those variables having significant safety functions.

Where a limiting safety system setting is specified for a variable on which a safety limit has been placed, the setting must be so chosen that automatic protective action will correct the abnormal situation before a safety limit is exceeded. If, during operation, it is determined that the automatic safety system does not function as required, the licensee shall take appropriate action, which may include shutting down the reactor.

3.11.3.2.1.2 Technical Evaluation The APRM Simulated Thermal Power- Upscale trip setpoint documented in the Fermi 2 Technical Requirements Manual Table TR 3.3.1.1-1 is the LSSS. It corresponds to the number calculated as the "NTSP" in the Fermi 2 GE methodology calculation, which TSTF-493 would recognize as the calculated limiting trip setpoint (LTSP). In the Fermi 2 setpoint methodology the NTSP is the limiting setting for the channel trip setpoint considering all credible instrument errors associated with the instrument channel. Therefore, the NTSP is the least conservative value (with an as-left tolerance) to which the channel must be reset at the conclusion of periodic testing to ensure that the analytical limit will not be exceeded during an anticipated operational occurrence or accident before the next periodic surveillance or calibration. It is impossible to set a physical instrument channel to an. exact value, so a calibration tolerance is established around the NTSP. Therefore, the NTSP adjustment is considered successful if the as-left instrument setting is within the setting tolerance (i.e., a range of values around the LTSP.

In this case of the APRM Simulated Thermal Power- Upscale trip setpoint, the Allowable Values are in TSs to satisfy the 10 CFR 50.36 requirements that LSSSs be in TSs. Additionally, to ensure proper use of the NTSP, the methodology for calculating the as-left and as-found

tolerances is incorporated, by reference, in TSs surveillance Footnote (e) as the Technical Requirements Manual (TRM). The licensee-proposed Table 3.3.1.1-1 Note (e) correctly identifies the "NTSP" as the LSSS documented in the TRM.

The NRC staff reviewed these two notes and found they are consistent with the wording of the two notes in Option A of TSTF-493, Revision 4. Therefore, the NRC staff finds the addition of Notes d and e to be acceptable for meeting the requirements for LSSS under 10 CFR 50.36 for the APRM Simulated Thermal Power- Upscale trip function.

Based on the above discussion and the NRC staff's review of the licensee's LAR, the NRC staff concludes that the licensee provided sufficient justifications for the proposed TS changes. The .

NRC staff concludes that the licensee has followed the guidance in Items A through C in Section VIII of Attachment 1 to RIS 2002-03, and therefore has met the relevant regulatory requirements of 10 CFR Part 50, Appendix K.

3.11.4 Instrumentation and Control Systems Conclusion The NRC staff reviewed the licensee's proposed plant-specific implementation of the Cameron LEFM OheckPius FW flow measurement device and the power uncertainty calculations. Based on its review of the licensee's LAR, uncertainty calculations, and referenced topical reports, the NRC staff finds that the licensee's proposed amendment is consistent with the approved Cameron Topical Report ER-80P and its supplement, Topical Report ER-157P, as well as the guidance of RIS 2002-03. The NRC staff also finds that the licensee adequately accounted for all instrumentation uncertainties in the reactor thermal power measurement uncertainty calculations. Therefore, the licensee's proposed amendment meets the relevant requirements of Appendix K to 10 CFR 50.

The NRC staff further concludes that the proposed TS changes meet the requirements of 10 CFR 50.36, the guidance of RG 1.1 05, Revision 3, and TSTF-493, Revision 4. On the basis of these considerations, the NRC staff finds the instrumentation and controls aspects of the proposed thermal power uprate acceptable.

3.12 Plant Systems 3.12.1 Regulatory Evaluation The NRC staff's review in the area of plant systems covers the impact of the proposed MUR power uprate on balance of plant piping, safety-related cooling water systems, ultimate heat sink, radioactive waste systems, and spent fuel pool (SFP) storage and cooling. The licensee evaluated the effect of the MUR on the plant systems. This evaluation is reflected in Enclosures 7 and 9 of the licensee's application dated February 7, 2013.

3.12.2 Technical Evaluation Balance of Plant Piping The principal construction codes for the Balance of Plant (BOP) systems are listed in the UFSAR Chapter 3, "Design of Structures, Components, Equipment, and Systems." Principal codes include ASME Boiler arid Pressure Vessel Code Sections Ill and VIII and American National Standards Institute B31.1 Standard Code for Pressure, Power Piping. Section 3.5.2 of the Fermi Safety Analysis Report (NEDC-33578P) lists 24 BOP systems evaluated for TPO

power uprate. Seventeen of these systems have no change in operating conditions from CLTP to TPO and are therefore acceptable for TPO. Eight systems have temperature increases of less than 2°F due to TPO.

The licensee has judged the corresponding piping stresses to be insignificant and therefore acceptable for TPO. One system, Offgas, has temperature increases of less than 1 percent and flow rate increase less than 2 percent due to TPO. The licensee has judged the piping stresses for Offgas piping to be insignificant and remain acceptable for TPO. One section of Feedwater (FW), the piping from No. 6 FW heater to containment, has temperature increases greater than 2°F and/or flow rates greater than 1 percent due to TPO.

The piping was analyzed by the licensee and found to be acceptable for TPO conditions. This piping was analyzed according to the process described in the TPO Licensing Topical Report, Appendix K, "Methods and Assumptions for Piping Evaluation of TPO Uprate." The process as described in Appendix K determines the percent increases in code equations for pressure, temperature and flow. The locations with the highest calculated stress to allowable ratio are determined. The calculated stress is then increased according to the appropriate equation percent and then compared to Code allowable stress. Cumulative fatigue usage factors are evaluated in a similar manner. As stated above, the licensee found the FW piping from the No.

6 FW heater to containment acceptable for TPO conditions. The licensee stated that associated pipe support loads will experience a small increase in the thermal loads(< 1 percent). This represents an insignificant load increase and is acceptable for TPO.

The licensee also evaluated the Turbine Stop Valve closure transient and the FW pump trip and found that no changes are required for TPO. Operation at TPO power results in changes to parameters affecting Flow Accelerated Corrosion (FAG) in systems associated with the turbine cycle. The plant FAG program monitors the effects of FAG and appropriate changes to piping inspections per the FAG program will project the need for maintenance/replacement prior to exceeding minimum wall thickness requirements. Thus the FAG program provides assurance that TPO will not adversely affect piping systems susceptible to FAG.

The licensee reviewed its response to Generic Letter (GL) 96-06, "Assurance of Equipment Operability and Containment Integrity during Design-Basis Accident Conditions," for the TPO uprate. The containment design temperatures and pressures in the current GL 96-06 evaluation*

are not exceeded under post-accident conditions for the TPO uprate. Therefore, the licensee's response to GL 96-06 remains valid under TPO uprate conditions.

Based on the considerations discussed above, the NRC staff concurs with the licensee's assessment that the BOP piping and supports are acceptable for TPO.

Safety-Related Cooling Water Systems The safety-related Emergency Service Water (ESW) system consists of the Emergency Equipment Service Water (EESW) system, Residual Heat Removal Service Water (RHRSW) system, and the Emergency Diesel Generator Service Water (EDGSW) system. Th*e ESW provides cooling to essential equipment during design basis accidents. The required performances of the EDGs do not change for TPO, therefore, the TPO heat loads remain within the rating of the EDGs. The performance of the RHRSW and EESW systems do not change for TPO because the LOCA analysis and containment response analysis were based on 102 percent of CLTP which bounds TPO power. The Emergency Equipment Cooling Water (EECW) system loads are .unaffected by TPO uprate.

Therefore, the NRC staff concludes that the safety-related cooling water systems are satisfactory for TPO.

Ultimate Heat Sink The Safety Analysis Report (NEDC-33578P) in the LAR states:

The Ultimate Heat Sink (UHS) for Fermi 2 is the Residual Heat Removal (RHR) complex, which consists of redundant enclosed reservoirs and mechanical draft cooling towers. The ESW systems, consisting of RHRSW, EDGSW, and EESW systems, provide water from the UHS for equipment cooling throughout the plant.

As a result of operation at the TPO Rated Thermal Power level, the actual post-LOCA heat load increases slightly, primarily due to higher reactor decay heat. However, the ability of the UHS to perform required safety functions is demonstrated with previous analyses based on 102 percent of CLTP. Therefore, all safety aspects of the UHS are within previous evaluations and the requirements are unchanged for TPO power uprate conditions. The current TS for UHS limits are adequate due to conservatism in the current design.

The NRC staff concludes that the UHS system is satisfactory for TPO based on the licensee's statement that the analysis of the adequacy of the UHS is based on DBA conditions at 102 percent of CLTP.

Radioactive Waste Systems The liquid radwaste system collects, monitors, processes, stores and returns processed radwaste for reuse, discharge or shipment. The major sources of liquid and solid radwaste are from the Condensate Filter Demineralizers (CFD). TPO uprate results in approximately a 2 percent increase in condensate flow causing a possible reduction in average time between back washes of CFD resin. The activated corrosion products in the waste stream will ir~crease in proportion to TPO uprate caused by increase in power and flow through the CFD. The Reactor Water Cleanup (RWCU) system and Fuel Pool Cooling Cleanup (FPCC) system will also have small increases in activity. None of thes.e factors affect plant safety and have minimal effect on the radwaste system. The requirements of 10 CFR 20 and 10 CFR 50 Appendix I will continue to be met.

The gaseous radwaste management systems include the offgas system and the various building ventilation systems, which collect control and process gaseous radwaste. The TPO increase in power will cause more radiolytic decomposition of water into hydrogen and oxygen, causing a higher heat load on offgas components. However, the increase is well within the design of the offgas system thus not causing any adverse effect on the offgas system. The non-condensable radioactive gases from the main condenser will increase from TPO uprate, but will have no effect on the release limit which is administratively controlled and not a function of core power. Therefore the TPO uprate does not affect the offgas system design or operation. Based on the above considerations, the NRC staff concludes that the Radioactive Waste Systems are satisfactory for TPO.

Spent Fuel Storage and Fuel Pool Cooling System The licensee has calculated that spent fuel pool heat loads for TPO remain within the capability of the FPCC system by controlling the timing of the discharge (fuel offload) to the SFP. The

FPCC system heat exchangers assisted by the RHR assist mode are sufficient to remove decay

heat during normal refueling. For a full core offload at TPO uprate, the RHR assist mode is sufficient to maintain the SFP water temperature below the design limit. Based on the above considerations, the NRC staff concludes that the Spent Fuel Pool Storage and Fuel Pool Cooling System is satisfactory for TPO.

3.12.3 Conclusion The NRC staff has reviewed the licensee's analyses of the impact of the proposed MUR power uprate on the BOP piping, safety-related cooling water systems, ultimate heat sink, radioactive waste systems, and SFP storage and cooling. The NRC staff has determined that the results of the licensee's analyses related to these areas will continue to meet the applicable acceptance criteria following implementation of the MUR power uprate. Therefore, the NRC staff finds the proposed MUR power uprate to be acceptable with respect to the plant systems review.

3.13 Containment and Heating, Ventilating and Air Conditioning (HVAC) 3.13.1 Regulatory Evaluation 3.13.1.1 Containment 10 CFR 50 Appendix A General Design Criterion (GDC) 4, Environmental and dynamic effects design basis, addresses the environmental qualification of SSCs important to safety. The NRC staff reviewed the licensee's prediction of conditions in containment during postulated accidents.

No regulation specifically addresses the determination of the mass and energy release into the containment following a postulated design basis accident. However, 10 CFR Part 50 Appendix A, GDC 16 and 50 address the reqt,Jirements for the containment pressure resulting from the discharge of mass and energy into the containment as a result of a postulated design basis LOCA.

GDC 16, .Containment design, specifies that the reactor containment and associated systems shall be provided to establish an essentially leak tight barrier against the uncontrolled release of radioactivity to the environment and to assure that the containment design conditions important to safety are not exceeded for as long as postulated accident conditions require.

GDC 38, Containment heat removal, specifies that a system to remove heat from the reactor containment be provided and that this system shall reduce rapidly, consistent with the functioning of other associated systems, the containment temperature and pressure following any LOCA.

GDC 50, Containment design basis, specifies that the reactor containment, including access openings, penetrations, and the Containment Heat Removal System (CHRS) shall be designed to accommodate, without exceeding (with sufficient margin) the design leakage rate resulting from a design basis LOCA.

SRP Section 6.2.1, Containment Functional Design; Section 6.2.1.1.C, Pressure-Suppression Type Boiling-Water Reactor (BWR) Containments; 6.2.1.3, Mass and Energy Release Analysis for Postulated LOCA; and Section 6.2.2, Containment Heat Removal Systems, provide review guidance in the area of containment safety analysis.

-70.:

3.13.1.2 Engineered Safety Features Heating, Ventilation and Air Conditioning Systems The NRC's regulations and guidance specify criteria for control room habitability and post-accident fission product control and removal.

GDC 4, Environmental and dynamic effects design basis, requires that SSCs important to safety be designed to accommodate the effects of and to be compatible with the environmental conditions associated with postulated accidents, including the effects of the release of post-accident fission products and toxic gases.

GDC 19, Control room, requires adequate radiation protection be provided to permit access and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of five rem whole body, or its equivalent, to any part of the body, for the duration of the accident.

GDC 41, Containment atmosphere cleanup, requires that systems to control fission products released into the reactor containment be provided to reduce the concentration and quality of fission products released to the environment following postulated accidents.

GDC 60, Control of releases of radioactive materials to the environment, requires that the plant design include means to control the release of radioactive gaseous and liquid effluents for normal operation and anticipated operational occurrences (defined in 10 CFR Part 50 Appendix A).

GDC 64, Monitoring radioactivity releases, requires that means shall be provided for monitoring effluent discharge paths and the plant environs for radioactivity that may be. released from normal operations, including anticipated operational occurrences, and postulated accidents.

Guidance in SRP Sections 6.4, Control Room Habitability System; 6.5.2, Containment Spray as a Fission Product Cleanup System; 9.4.1, Control Room Area Ventilation System; 9.4.2, Spent Fuel Pool Area Ventilation System; 9.4.3, Auxiliary and Radwaste Area Ventilation System; 9.4.4, Turbine Area Ventilation. System; and 9.4.5, Engineered Safety Feature Ventilation System, and in Regulatory Guide (RG) 1.52 and 1.78, contain specific review criteria.

3.13.2 Technical Evaluation 3.13.2.1 Containment Appendix G to NEDC 32938P, Revision 1, "Licensing Topical Report: Generic Guidelines and Evaluations for General Electric BWR Thermal Power Optimization," dated May 2003, outlines the methods, approach and scope.for the TPO uprate contain merit evaluation for LOCA. This report discusses that the previous containment evaluations are bounding for the TPO uprate because they have previously considered ;::: 2 percent power uncertainty as required by previous methodology. Although the nominal operating conditions will be increased slightly because of this TPO uprate, the required bounding conditions for the limiting analytical cases remain the same as previously documented.

Containment systems are described in the Fermi 2 UFSAR Section 6.2. The accident response analysis is also discussed in the Fermi UFSAR Subsection 6.2.1, wherein plant response to

various large and small LOCAs is evaluated and the short and long-term containment pressure and temperature responses are presented.

Containment LOCA Pressure and Temperature Response The effects of TPO uprate conditions on LOCA containment pressure and temperature are bounded by the previous analysis performed at,:: 102 percent of CLTP. Short-term containment pressure and temperature response analyses have been performed using the GE code, M3CPT. The power uprate methods approved by the NRC permit the use of the M3CPT computer code to calculate the short-term containment response. As documented in NEDC 32938P, Revision 1, the licensee performed long-term containment pool heatup analysis for the limiting UFSAR events to show acceptable pool temperatures considering limits due to:

Containment design temperature

Net positive suction head

Equipment design or qualification temperatures These analyses have been performed using the GE computer code Super Hex (SHEX). The NRC has accepted the use of the SHEX code for Mark I containment long-term analysis. Decay heat inputs are conservatively based on ANSI/ANS 5.1-1979, "Decay Heat Power in Light Water

Reactors," results.

Containment Dynamic Loads The short-term containment pressure, temperature, and vent flow have been calculated for up to 102 percent of CLTP with M3CPT. Since the previous analysis bounds the TPO conditions for LOCA dynamic design loads, they are not affected by the TPO uprate.

The Safety-Relief Valve (SRV) opening setpoint pressures are not increased for TPO uprate

and therefore, the SRV loads associated with SRV actuations following initiation of an event are unchanged by TPO uprate. There is also no change in limiting case break flow conditions due to TPO uprate because the analysis basis is bounded by the previous cases, therefore, the analytical or experimental basis for the LOCA subcompartment pressurization dynamic loads and the basis for suppression chamber/wetwell loads remain consistent with the evaluation for CL TP conditions.

The subcompartment pressurization loads continue to remain within allowable structural limits for a TPO power uprate because the changes are within existing margins because of the very small changes to operating conditions associated with TPO uprate with no dome pressure change. There is no significant change because TPO uprate only changes system operating temperatures and pressures slightly: < 1oF (recirculation lines), < 2°F (FW lines only), < 5 psi (FW lines only), and*< 1 psi (recirculation discharge lines only) due to slightly higher pressure drops at TPO flow rates. Vessel dome pressure and other portions of the primary coolant pressure boundary remain at current operating pressure (or lower, e.g., main steam line).

Therefore, subcompartment pressurization will not significantly change.

The NRC staff has reviewed the licensee's key assumptions and methodology used for the containment systems performance analyses. The staff has confirmed that the current analysis of record documented in the UFSAR is bounding for the MUR power upratewith respect to

containment pressure and temperature response, dynamic loads due to LOCA and SRV actuation.

Generic Letter (GL) 96-06 I . .

The licensee reviewed its response to GL 96~06, "Assurance of Equipment Operability and Containment Integrity during Design-Basis Accident Conditions," as a result of the MUR power uprate impact. The containment design temperatures and pressures in the current GL 96-06 evaluations are not exceeded under post-accident conditions for the MUR power uprate. Therefore, the NRC staff concludes that its response to GL 96-06 remains valid under MUR power uprate conditions.

Emergency Core Cooling System Net Positive Suction Head The licensee evaluated the ECCS performance to the Net Positive Suction Head (NPSH) to the CHRS pumps. The current containment analyses were based on 102 percent CLTP, and therefore there is no change in available NPSH for systems using Suppression Pool (SP) water.

The TPO uprate does not affect compliance with the ECCS pump NPSH requirements.

Therefore, the NRC staff concludes that all safety aspects of the ECCS NPSH are within previous evaluations and the requirements are unchanged for the TPO uprate conditions.

3.13.2.2 Heating, Ventilation and Air Conditioning Systems Main Control Room Atmosphere Control System Control room habitability following a postulated accide.nt is bounded by the current evaluation of the main control room atmosphere control system at two percent above the current power level.

Therefore, the licensee asserts, and the NRC staff agrees, that the Heating, Ventilation and Air Conditioning (HVAC) analyses described above have been performed at a power level that bounds operation at the uprated conditions.

Standby Gas Treatment System The Standby Gas Treatment System (SGTS) minimizes the offsite and control room dose rates during venting and purging of the containment atmosphere under abnormal circumstances. The capability of the SGTS is not changed by the uprate conditions. The licensee stated that the SGTS can accommodate design-basis-accident conditions at 102 percent of CLTP. Therefore, the NRC staff concludes that the SGTS remains capable of performing its saf~ty function for the uprate conditions.

Post-LOCA Combustible Gas Control System The original licensing basis of the Combustible Gas Control System (CGCS) was to maintain the post-LOCA concentration of oxygen or hydrogen in the containment atmosphere below the.

flammable limit. This evaluation is no longer applicable to Fermi 2 as the hydrogen combining function requirements of the system have been deleted from the TS in accordance with the 10 CFR 50.44 guidelines. Therefore, this system is not applicable to the Fermi 2 MUR.

3.13.3 Conclusion The current containment and HVAC analyses have been performed at a power level that bounds operation at the uprated power level. Therefore, the NRC staff finds the proposed uprate acceptable with respect to the containment and HVAC systems. Fermi 2 remains in compliance with GDCs 4, 16, 19, 38, 41, 50, 60, 61, and 64 and the applicable SRP guidance at MUR conditions.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Michigan State official was notified of the proposed issuance of the amendments. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendment changes requirements with respect to the installation or use of the facilities components located within the restricted area as defined in 10 CFR Part 20 and change surveillance requirements. The NRC staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration, and there has been no public comment on such finding (78 FR 35069 dated June 11, 2013). Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.

6.0 REGULATORY COMMITMENTS To support the proposed Fermi 2 MUR power up~ate, the licensee made the following commitments (as stated). A requirement is included in Section 3 ofthe NRC license amendment associated with this SE, to ensure that these Regulatory Commitments are incorporated coincident with the licensee's implementation of the amendment COMMITTED ONE-TIME ON-GOING COMMITMENT DATE OR ACTION COMMITMENT OUTAGE.

(Yes/No) (Yes/No)

Limitations regarding operation with Prior to Startup an inoperable LEFM system will be from the No ' Yes included in the TRM. Sixteenth Refueling Outage*

A process will be implemented to Prior to Startup use the LEFM feedwater flow to from the adjust or correct the existing No Yes Sixteenth feedwater flow venturi-based Refueling Outage signals.

COMMITTED ONE-TIME ON-GOING COMMITMENT DATE OR ACTION COMMITMENT OUTAGE (Yes/No) (Yes/No)

Plant maintenance and calibration procedures will be revised to incorporate Cameron's Prior to Startup maintenance and calibration from the No Yes requirements. Initial preventive Sixteenth scope and frequ~ncy Vl(ill be based Refueling Outage.

on vendor recommendations Modifications for the power uprate Prior to Startup will be implemented. from the Yes No Sixteenth Refueling Outage Necessary procedure revisions for Prior to Startup the power uprate will be completed. from the No Yes Sixteenth Refueling Outage The plant simulator will be modified Prior to Startup for the uprated conditions and the from the changes will be validated in Yes No Sixteenth accordance with plant configuration Refueling Outage control processes.

Operator training will be completed Prior to Startup prior to implementation of the from the.

' Yes No proposed power uprate changes. Sixteenth Refueling Outage Plant testing for the proposed power uprate changes will be completed as described in As described Yes No Enclosure 7, Section 10.4, "Testing."

Plant-specific analyses for all Prior to Startup potentially limiting events will be from the performed on a cycle-specific basis Yes No Sixteenth as part of the reload licensing Refueling Outage process.

7.0 CONCLUSION

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

Principal Contributors: G. Lapinsky, NRR J. Dozier. NRR N. Iqbal, NRR C. Hunt, NRR W. Jessup, NRR G. Bedi, NRR M. Hardgrove, NRR C. Sydnor, NRR S. Basturescu, NRR S. Wyman, NRR G. Purciarello; NRR B. Lee, NRR Date: February 10, 2014

LIST OF ACRONYMS A ampere AOO anticipated operational occurrences AC alternating current AOP abnormal operating procedure AOR analysis of record AOT allowed outage time APRM average power range monitor ARL Alden Research Laboratories ART adjusted reference temperature ASME American Society of Mechanical Engineers AST alternative source term ATWS anticipated transient without SCRAM BHP brake horsepower BOP balance-of-plant BWR boiling water reactor

BWRVIP Boiling Water Reactor Vessel and Internals Project CHRS containment heat removal system CLTP current licensed thermal power CLB current licensing basis CPPU constant pressure power operation CTP . core thermal power CUF cumulative usage factor DBA design-basis accident DC direct current DBA design-basis accident ECCS emergency core cooling system EDG emergency diesel generator EDMG extreme damage mitigation system EFPY effective full power years EOL end-of-license EOP emergency operating procedure EQ environmental qualification EPU extended power uprate F degrees *Fahrenheit FAC flow accelerated corrosion FHA fuel handling accident

. FIV flow induced vibration FW feedwater GDC General Design Criterion gpm gallons per minute GL Generic Letter HELB high energy line break HP horsepower HVAC heating, ventilation and air conditioning 1ST inservice testing kV kiloVolt LAR license amendment request LEFM leading edge flow meter LOCA loss-of-coolant accident

LSSS limiting safety system settings MCPR minimum critical power ratio MSIV main steam isolation valve MSLB main steam line break MUR measurement uncertainty recapture MVA mega-voltamperes MW megawatt MWe megawatts-electric MWt megawatts-thermal NIST National Institute of Standards and Technology NRC U.S. Nuclear Regulatory Commission NSSRV nuclear system safety relief valves OLMCPR open limit minimum critical power ratio OLTP original licensed thermal power OPRM oscillation power range monitor Psi a pounds per square inch, absolute Psig pounds per square inch, gauge P/T pressure-temperature PTLR Pressure Temperature Limits Report RAI request for additional information RCP reactor coolant pump RCPB reactor coolant pressure boundary RCS reactor coolant. system RG Regulatory Guide RIPD reactor internal pressure difference RIS Regulatory Issue Summary RPS reactor protection system RPV reactor pressure vessel RTD resistance temperature detector RTP rated thermal power RTNDT reference temperature for non-ductile transition RVI reactor vessel internals RWCU reactor water cleanup system SAR safety analysis report SBO station blackout SE safety evaluation SER safety evaluation report SFP spent fuel pool SRP NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear.Power Plants SRSS square-root-of-the-sum-of-the-squares SRV safety relief valve SSCs structures, systems, and components TLTR Generic Guidelines and Evaluations for General Electric Boiling Water Reactor Thermal Power Optimization

TPO thermal power optimization TPSAR Thermal Power Safety Analysis Report TRM Technical Requirements Manual TRMLCO Technical Requirements Manual Limiting Condition for Operation TS * . Technical Specification UFM ultrasonic flow meter

UHS Ultimate Heat Sink UFSAR updated final safety analysis report USE upper shelf energy v volt

ML13364A131 *By Memo Dated OFFICE LPL3-1/PM LPL3-1/LA DE/EICB DSS/STSB DE/EPNB NAME TWengert MHenderson JThorp* REIIiott TLupold*

DATE 01/27/14 01/27/14 01/08/14 01/27/14 09/24/13 OFFICE DE/ESGB DE/EMCB DRNAHPB DSS/SBPB DRNAFPB NAME GKulesa* AMcMurtray* UShoop* GCasto* AKiein*

DATE 08/27/13 11/26/13 05/16/13 10/07/13 07/22/13 OFFICE DRNAADB DE/EVIB DSS/SRXB DE/EEEB DSS/SCVB NAME TTate* SRosenberg* CJackson* JZimmerman* RDennig*

DATE 09/06/13 12/17/13 12/02/13 12/04/13 10/03/13 OFFICE LPL3-1/BC OGC-NLO DORUD LPL3-1/PM w/comments NAME RCarlson BHarris MEvans TWengert DATE 02/06/14 01/30/14 02/07/14 02/10/14

v t e Letter Sequence Approval
TAC:MF0650, Clarify Application of Setpoint Methodology for LSSS Functions (Approved, Closed)
Initiation Request, Request, Request Acceptance Supplement Administration Withholding Request Acceptance Questions 8 May 2013 3 June 2013 5 June 2013 Answers 7 June 2013 15 July 2013 Results Approval, Approval Other:
MONTHYEAR2013-06-05 [Table View]

ML13364A131

Text

SUBJECT:

Dear Mr. Plena:

Enclosures:

Attachment:

1.0 INTRODUCTION

2.0 BACKGROUND

1.3 NRC Staff Conclusion

NRC Staff Evaluation

NRC Staff Evaluation

NRC Staff Evaluation

4.0 STATE CONSULTATION

5.0 ENVIRONMENTAL CONSIDERATION

7.0 CONCLUSION

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