ML18096A542

Issuance of Amendment Measurement Uncertainty Recapture Power Uprate (CAC MF9930; EPID L-2017-LLS-0002)

ML18096A542
Person / Time
Site:	Hope Creek
Issue date:	04/24/2018
From:	Lisa Regner Plant Licensing Branch IV
To:	Sena P Public Service Enterprise Group
Regner L, NRC/NRR/DORL/LPL4, 415-1906
References
CAC MF9930, EPID L-2017-LLS-0002
Download: ML18096A542 (117)

Text

OFFICIAL USE ONLY PROPRIETARY INFORMATION NU WASHINGTON. tLC 20555-00D1 April 24, 2018 Mr. Peter P. Sena, Ill President and Chief Nuclear Officer PSEG Nuclear LLC - N09 P.O. Box 236 Hancocks Bridge, NJ 08038

SUBJECT:

HOPE CREEK GENERATING STATION - ISSUANCE OF AMENDMENT NO. 212 RE: MEASUREMENT UNCERTAINTY RECAPTURE POWER UPRATE (CAC NO. MF9930; EPID L-2017-LLS-0002)

Dear Mr. Sena:

The U.S. Nuclear Regulatory Commission (NRC; Commission) has issued the enclosed Amendment No. 212 to Renewed Facility Operating License No. NPF-57 for the Hope Creek Generating Station. This amendment consists of changes to the Technical Specifications in response to your application dated July 7, 2017, 1 as supplemented by additional letters.2 The amendment revises the Renewed Facility Operating License and Technical Specifications to implement a measurement uncertainty recapture power uprate. Specifically, the amendment authorizes an increase in the maximum licensed thermal power level from 3,840 megawatts thermal to 3,902 megawatts thermal, which is an increase of approximately 1.6 percent, which does not exceed 120 percent of the original licensed thermal power.

The NRC staff evaluated the proposed license conditions 2.C.(28) and 2.C.(29) in your application and determined that they are not necessary since the referenced license amendment requests were issued by NRC letters dated August 4, and December 14, 2017. 3 transmitted herewith contains sensitive unclassified information.

When seoarated from Enclosure 3. this document is decontrolled.

1 Agencywide Document Access and Management System (ADAMS) Accession Package No. ML17188A259 2 Letters dated November 1 (ADAMS Accession No. ML173058270); November 27 (ADAMS Accession No. ML17333A853); December 14 (ADAMS Accession No. ML173488289); December 19 (4 letters: ADAMS Accession Nos. ML17353A744, ML17353A778, ML17353A831, and ML17353A926); December 22, 2017 (ADAMS Accession No. ML17356A139); January 22, 2018 (ADAMS Accession No ML18022A147) 3 ADAMS Accession Nos. ML17216A022 and ML17324A840, respectively OFFICIAL USE ONLY PROPRIETARY INFORMATION

P. Peter OFFICIAL USE ONLY PROPRIETARY INFORMATION The NRC has determined that the related safety evaluation (SE) contains proprietary information pursuant to Title 10 of the Code of Federal Regulations, Section 2.390, "Public inspections, exemptions, requests for withholding." Accordingly, the NRC staff has also prepared a non-proprietary version of the SE, which is provided in Enclosure 2. The proprietary version of the SE is provided in Enclosure 3. The Notice of Issuance will be included in the Commission's next biweekly Federal Register notice.

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

Docket No. 50-354

Enclosures:

1.

Amendment No. 212 to Renewed License No. NPF-57

2.

Non-Proprietary Safety Evaluation

3.

Proprietary Safety Evaluation cc:w/o Enclosure 3 Lisa M. Regner, Senior Project Manager Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation OFFICb\\L USE ONLY PROPRIETARY INFORMATION

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 PSEG NUCLEAR LLC DOCKET NO. 50-354 HOPE CREEK GENERATING STATION AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 212 Renewed License No. NPF-57

1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment filed by PSEG Nuclear LLC dated July 7, 2017, as supplemented by letters dated November 1, November 27, December 14, December 19 (4 letters), December 22, 2017, and January 22, 2018, 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 set forth in 1 O CFR Chapter I; 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 Renewed Facility Operating License No. NPF-57 is hereby amended to read as follows:

(2)

Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 212, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated into the renewed license. PSEG Nuclear LLC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

The license amendment is effective as of its date of issuance and shall be implemented within 120 days.

Attachment:

Changes to the Renewed Facility Operating License and Technical Specifications Dateoflssuance: April 24, 2018 FOR THE NUCLEAR REGULATORY COMMISSION Tara Inverso, Acting Deputy Director Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

ATTACHMENT TO LICENSE AMENDMENT NO. 212 HOPE CREEK GENERATING STATION RENEWED FACILITY OPERATING LICENSE NO. NPF-57 DOCKET NO. 50-354 Replace the following pages of the Renewed Facility Operating License with the revised pages.

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

Remove Insert Replace the following pages of the 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 1-6 2-4 3/4 1-4 3/4 1-16 3/4 3-59 3/4 4-1 3/4 4-2a 3/4 10-2 Insert 1-6 2-4 3/4 1-4 3/4 1-16 3/4 3-59 3/4 4-1 3/4 4-2a 3/4 10-2 reactor operation, as described in the Final Safety Analysis Report, as supplemented and amended; (4)

PSEG Nuclear LLC, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (5)

PSEG Nuclear LLC, 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)

PSEG Nuclear LLC, 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. Mechanical disassembly of the GE14i isotope test assemblies containing Cobalt-60 is not considered separation.

(7)

PSEG Nuclear LLC, pursuant to the Act and 10 CFR Part 30, to intentionally produce, possess, receive, transfer, and use Cobalt-60.

C.

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

(1)

(2)

Maximum Power Level PSEG Nuclear LLC is authorized to operate the facility at reactor core power levels not in excess of 3902 megawatts thermal ( 100 percent rated power) in accordance with the conditions specified herein.

Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 212, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the renewed license. PSEG Nuclear LLC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

Renewed License No. NPF-57 Amendment No. 212

(7)

(8) Fire Protection (Section 9.5.1.8, SSER No. 5: Section 9.5.1, SSER No. 6)

PSEG Nuclear LLC shall implement and maintain in effect all provisions of the approved fire protection program as described in the Final Safety Analysis Report for the facility through Amendment No. 15 and as described in its submittal dated May 13, 1986, and as approved in the SER dated October 1984 (and Supplements 1 through 6) subject to the following provision:

PSEG Nuclear LLC may make changes to the approved fire protection program without prior approval of the Commission only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire.

Solid Waste Process Control Program (Section 11.4.2, SER:

Section 11.4, SSER No. 4)

DELETED (9)

Emergency Planning (Section 13.3, SSER No. 5)

DELETED (10)

Initial Startup Test Program (Section 14, SSER No. 5)

DELETED

( 11)

Partial Feedwater Heating (Section 15.1, SER: Section 15.1, SSER No. 5:

Section 15.1, SSER No. 6)

The facility shall not be operated with a rated thermal power feedwater temperature less than 331.5°F for the purpose of extending the normal fuel cycle.

( 12)

Detailed Control Room Design Review (Section 18.1, SSER No. 5)

Renewed License No. NPF-57 Amendment No. 212

DEFINITIONS PROCESS CONTROL PROGRAM 1.33 The PROCESS CONTROL PROGRAM (PCP) shall contain the current formulas, sampling, analyses, test, and determinations to be made to ensure that processing and packing of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR Parts 20, 61, and 71, State regulations, burial ground requirements, and other requirements governing the disposal of solid radioactive waste.

PURGE - PURGING 1.34 PURGE or PURGING shall be the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such manner that replacement air or gas is required to purify the confinement.

RA TED THERMAL POWER 1.35 RA TED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 3902 MWt.

REACTOR PROTECTION SYSTEM RESPONSE TIME 1.36 REACTOR PROTECTION SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its trip setpoint at the channel sensor until de-energization of the scram pilot valve solenoids. The response time may be measured by any series of sequential, overlapping or total steps such that the entire response time is measured.

REPORTABLE EVENT 1.37 A REPORTABLE EVENT shall be any of those conditions specified in Section 50. 73 to 10 CFR Part 50.

ROD DENSITY 1.38 ROD DENSITY shall be the number of control rod notches inserted as a fraction of the total number of control rod notches. All rods fully inserted is equivalent to 100% ROD.

DENSITY.

HOPE CREEK 1-6 Amendment No. 212

TABLE 2.2.1-1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS FUNCTIONAL UNIT TRIP SETPOINT

1. Intermediate Range

~ 120/125 divisions of Monitor, Neutron Flux-High full scale

2. Average Power Range Monitor:

a. Neutron Flux-Upscale (Setdown)

b. Simulated Thermal Power-Upscale**

1) Flow Biased-Two Recirculation Loop Operation

2) Flow Biased-Single Recirculation Loop Operation

c. Neutron Flux - Upscale

d. Inoperative

e. 2-0ut-Of-4 Voter

f. OPRM Upscale

3. Reactor Vessel Steam Dome Pressure - High

4. Reactor Vessel Water Level -

Low, Level 3

5. Main Steam Line Isolation Valve - Closure See Bases Figure B 3/4 3-1.

~ 17% of RATED THERMAL POWER

~ 0. 56w + 58%** (a) with a maximum of~ 113.5% of RATED THERMAL POWER

~ 0.56(w-10.8%) + 58%**(a) with a maximum of

~ 113.5% of RATED THERMAL POWER

~ 116.3% of RATED THERMAL POWER NA NA See CORE OPERATING LIMITS REPORT

~ 1037 psig

~ 12.5 inches above instrument zero*

o::; 8% closed ALLOWABLE VALUES

~ 122/125 divisions of full scale

o::; 19% of RATED THERMAL POWER

~ 0.56w + 60%** with a maximum of

~ 115.5%

of RATED THERMAL POWER

~ 0.56(w-9%) + 60%**

with a maximum of

~ 115.5% of RATED THERMAL POWER

~ 118.3% of RATED THERMAL POWER NA NA NA

o::; 1057 psig

~ 11.0 inches above instrument zero

o::; 12% closed The Average Power Range Monitor Scram function varies as a function of recirculation loop drive flow (w).

(a) When the Automated BSP Scram Regions Setpoints are implemented in accordance with Action 10 of Table 3.3.1-1, the Simulated Thermal Power-Upscale Flow Biased Setpoint will be adjusted per the CORE OPERATING LIMITS REPORT HOPE CREEK 2-4 Amendment No. 212

REACTIVITY CONTROL SYSTEMS LIMITING CONDITION FOR OPERATION (Continued}

ACTION (Continued)

d.

One or more BPWS groups with four or more inoperable control rods*****, within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, restore control rod(s) to OPERABLE status.

Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

e.

With more than 8 control rods inoperable, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

f.

With one or more scram discharge volume (SDV) vent or drain lines*** with one valve inoperable, isolate the associated line within 7 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.****

g.

With one or more SDV vent or drain lines*** with both valves inoperable, isolate the associated line within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.****

SURVEILLANCE REQUIREMENTS 4.1.3.1.1 The scram discharge volume drain and vent valves shall be demonstrated OPERABLE in accordance with the Surveillance Frequency Control Program by:

a.

Verifying each valve to be open,* and

b.

Cycling each valve through at least one complete cycle of full travel.

These valves may be closed intermittently for testing under administrative controls.

May be rearmed intermittently, under administrative control, to permit testing associated with restoring the control rod to OPERABLE status.

Separate Action entry is allowed for each SDV vent and drain line.

An isolated line may be unisolated under administrative control to allow draining and venting of the SDV.

Not applicable when THERMAL POWER is greater than 8.5% RA TED THERMAL POWER HOPE CREEK 3/4 1-4 Amendment No. 212

REACTIVITY CONTROL SYSTEMS 3/4.1.4 CONTROL ROD PROGRAM CONTROLS ROD WORTH MINIMIZER LIMITING CONDITION FOR OPERATION 3.1.4.1 The Rod worth minimizer (RWM) shall be OPERABLE.

APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2*#, when THERMAL POWER is less than or equal to 8.5% of RATED THERMAL POWER, minimum allowable low power setpoint.

ACTION:

a.

With the RWM inoperable after the first 12 control rods are fully withdrawn, operation may continue provided that control rod movement and compliance with the prescribed control rod pattern is verified by a second licensed operator or other technically qualified member of the unit technical staff who is present at the reactor control console.

b.

With the RWM inoperable before the first twelve (12) control rods are fully withdrawn, one startup per calendar year may be performed provided that the control rod movement and compliance with the prescribed control rod pattern are verified by a second licensed operator or other technically qualified member of the unit technical staff who is present at the reactor control console.

c.

Otherwise, control rod movement may be only by actuating the manual scram or placing the reactor mode switch in the Shutdown position.

SURVEILLANCE REQUIREMENTS 4.1.4.1 The RWM shall be demonstrated OPERABLE:

a.

In OPERATIONAL CONDITION 2 within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> prior to withdrawal of control rods for the purpose of making the reactor critical, and in OPERATIONAL CONDITION 1 within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> prior to RWM automatic initiation when reducing THERMAL POWER, by verifying proper indication of the selection error of at least one out-of-sequence control rod.

Entry into OPERATIONAL CONDITION 2 and withdrawal of selected control rods is permitted for the purpose of determining the OPERABILITY of the RWM prior to withdrawal of control rods for the purpose of bringing the reactor to criticality.

See Special Test Exception 3.10.2.

HOPE CREEK 3/41-16 Amendment No. 212

TABLE 3.3.6-2 CONTROL ROD BLOCK INSTRUMENTATION SETPOINTS TRIP FUNCTION TRIP SETPOINT

1. ROD BLOCK MONITOR

a. Upscale<a>

i.) Low Trip Setpoint {LTSP)<bl ii.) Intermediate Trip Setpoint (ITSP)(c) iii.) High Trip Setpoint {HTSP)<dl

b. Inoperative

c. Downscale

2. APRM

a. Simulated Thermal Power - Upscale*

1) Flow Biased - Two Recirculation Loop Operation

2) Flow Biased - Single Recirculation Loop Operation

b. Inoperative

c. Downscale

d. Simulated Thermal Power - Upscale (Setdown)

3. SOURCE RANGE MONITORS

a. Detector not full in

b. Upscale

c. Inoperative

d. Downscale

4. INTERMEDIATE RANGE MONITORS

a. Detector not full in

b. Upscale

c. Inoperative

d. Downscale

5. SCRAM DISCHARGE VOLUME

a. Water Level-High (Float Switch)

6. Deleted

7. REACTOR MODE SWITCH SHUTDOWN POSITION NA s 0.56w + 53.1%* with a maximum of s 108% of RATED THERMAL POWER s 0.56(w-10.8%) + 53.1%* with a maximum of s 108% of RATED THERMAL POWER NA

4% of RATED THERMAL POWER s 11% of RATED THERMAL POWER NA s 1.0 x 105 cps NA

3 cps NA s 108/125 divisions of full scale NA

5/125 divisions of full scale 109'1" (North Volume) 108'11.5" (South Volume)

NA

• The rod block function is varied as a function of recirculation loop flow (w).

• • Refer to the CORE OPERATING LIMITS REPORT for these values ALLOWABLE VALUE NA s 0.56w + 55.1 %* with a maximum of s 111 % of RATED THERMAL POWER s 0.56(w-9%} + 55.1 %* with a maximum of s 111% of RATED THERMAL POWER NA

2% of RATED THERMAL POWER s 13% of RA TED THERMAL POWER NA s 1.6 x 105 cps NA

1.8 cps NA s 110/125 divisions of full scale NA

3/125 divisions of full scale 109'3" (North Volume) 109'1.5" (South Volume)

NA

a. Each upscale trip level is applicable over its specified rated power range. All RBM trips are automatically bypassed below the low power setpoint {LPSP). The upscale L TSP is applied between the LPSP and the intermediate power setpoint (IPSP). The upscale ITSP is applied between the IPSP and the high power setpoint (HPSP). The HTSP is applied above the HPSP.

b. APRM Simulated Thermal Power is ;:: 28% and < 63% RTP

c. APRM Simulated Thermal Power is ;:: 63% and < 83% RTP

d. APRM Simulated Thermal Power is ;:: 83%

HOPE CREEK 3/4 3-59 Amendment No. 212

3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM RECIRCULATION LOOPS LIMITING CONDITION FOR OPERATION 3.4.1.1 Two reactor coolant system recirculation loops shall be in operation.

APPLICABILITY: OPERATIONAL CONDITIONS 1* and 2*.

ACTION:

a.

With one reactor coolant system recirculation loop not in operation:

1.

Within 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />s:

a)

Place the recirculation flow control system in the Local Manual mode, and b)

Reduce THERMAL POWER to s 59.89% of RATED THERMAL POWER, and c)

Increase the MINIMUM CRITICAL POWER RATIO (MCPR)

Safety Limit per Specification 2.1.2, and d)

Reduce the AVERAGE PLANAR LINEAR HEAT GENERATION RA TE{APLHGR) limit to a value specified in the CORE OPERATING LIMITS REPORT for single loop operation, and e)

Reduce the LINEAR HEAT GENERATION RATE (LHGR) limit to a value specified in the CORE OPERATING LIMITS REPORT for single loop operation, and f)

Limit the speed of the operating recirculation pump to less than or equal to 90% of rated pump speed, and g)

Perform surveillance requirement 4.4.1.1.2 if THERMAL POWER is s 38% of RA TED THERMAL POWER or the recirculation loop flow in the operating loop is s 50% of rated loop flow.

2.

Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, reduce the Average Power Range Monitor (APRM) Scram Trip Setpoints and Allowable Values to those applicable for single recirculation loop operation per Specification 2.2.1; otherwise, declare the APRM channel INOPERABLE and take the action of RPS Instrumentation TS 3.3.1 ACTION a.

3.

Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, reduce the APRM Control Rod Block Trip Setpoints and Allowable Values to those applicable for single recirculation loop operation per Specification 3.3.6; otherwise declare the APRM channel INOPERABLE and take the action of Control Rod Block Instrumentation TS 3.3.6 ACTION a and b.

See Special Test Exception 3.10.4.

HOPE CREEK 3/4 4-1 Amendment No. 212

REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS 4.4.1.1.1 With one reactor coolant system recirculation loop not in operation in accordance with the Surveillance Frequency Control Program verify that:

a.

Reactor THERMAL POWER is s 59.89% of RATED THERMAL POWER, and

b.

The recirculation flow control system is in the Local Manual mode, and

c.

The speed of the operating recirculation pump is less than or equal to 90% of rated pump speed.

4.4.1.1.2 With one reactor coolant system recirculation loop not in operation, within no more than 15 minutes prior to either THERMAL POWER increase or recirculation loop flow increase, verify that the following differential temperature requirements are met if THERMAL POWER is s 38% of RATED THERMAL POWER or the recirculation loop flow in the operating recirculation loop is s 50% of rated loop flow:

a.

s 145°F between reactor vessel steam space coolant and bottom head drain line coolant, and

b.

s 50°F between the reactor coolant within the loop not in operation and the coolant in the reactor pressure vessel, and

c.

s 50°F between the reactor coolant within the loop not in operation and the operating loop.

The differential temperature requirements or Specifications 4.4.1.1.2b and 4.4.1.1.2c do not apply when the loop not in operation is isolated from the reactor pressure vessel.

4.4.1.1.3 DELETED.

HOPE CREEK 3/4 4-2a Amendment No. 212

SPECIAL TEST EXCEPTIONS 3/4.10.2 ROD WORTH MINIMIZER LIMITING CONDITION FOR OPERATION 3.10.2 The sequence constraints imposed on control rod groups by the rod worth minimizer (RWM) per Specification 3.1.4.1 may be suspended for the following tests provided that control rod movement prescribed for this testing is verified by a second licensed operator or other technically qualified member of the unit technical staff present at the reactor console:

a.

Shutdown margin demonstrations, Specification 4.1.1.

b.

Control rod scram, Specification 4.1.3.2.

c.

Control rod friction measurements.

APPLICABILITY:

OPERATIONAL CONDITIONS 1 and 2 when THERMAL POWER is less than or equal to 8.5% of RA TED THERMAL POWER ACTION:

With the requirements of the above specification not satisfied, verify that the RWM is OPERABLE per Specifications 3.1.4.1.

SURVEILLANCE REQUIREMENTS 4.10.2 When the sequence constraints imposed by the RWM are bypassed, verify:

a.

That movement of the control rods from 75% ROD DENSITY to the RWM low power setpoint is limited to the approved control rod withdrawal sequence during scram and friction tests.

b.

That movement of control rods during shutdown margin demonstrations is limited to the prescribed sequence per Specification 3.10.3.

c.

Conformance with this specification and test procedures by a second licensed operator or other technically qualified member of the unit technical staff.

HOPE CREEK 3/4 10-2 Amendment No. 212

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 212 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-57 PSEG NUCLEAR LLC HOPE CREEK GENERATING STATION DOCKET NO. 50-354 This document contains proprietary information pursuant to Title 10 of the Code of Federal Regulations (10 CFR) Section 2.390.

Proprietary information is identified by underlined text (in bold font) enclosed within double brackets as shown here ((example proprietary text)).

OFFICIAL USE ONLY PROPRIETARY INFORMATION i

OFFICIAL USE ONLY PROPRIETARY INFORMATION TABLE OF CONTENTS INTRODUCTION........................................................................................................

1.1 Background................................................................................................................

1.1.1 General Site Information.............................................................................................

1.1.2 Previous Power Uprates for Hope Creek Generating Station......................................

1.1.3 General Approach for MUR Power Uprates................................................................

1.2 Licensees Approach..................................................................................................

1.3 Method of NRC Review..............................................................................................

REGULATORY EVALUATION..................................................................................

2.1 General Design Criteria..............................................................................................

2.2 Technical Specification Requirements........................................................................

2.3 Other Regulatory Requirements.................................................................................

2.4 Guidance Documents.................................................................................................

TECHNICAL EVALUATION.....................................................................................

3.1 Overview of the Thermal Power Optimization Safety Analysis Report (TSAR)..........

3.2 TSAR Section 2.0 - Reactor Core and Fuel Performance.........................................

3.2.1 TSAR Section 2.1 - Fuel Design and Operation........................................................

3.2.2 TSAR Section 2.2 - Thermal Limits Assessment......................................................

3.2.3 TSAR Section 2.3 - Reactivity Characteristics.........................................................

3.2.4 TSAR Section 2.4 - Thermal Hydraulic Stability.......................................................

3.2.5 TSAR Section 2.5 - Reactivity Control.....................................................................

3.2.6 TSAR Section 2.6 - Additional Limitations and Conditions Related to Reactor Core and Fuel Performance........................................................................

3.3 TSAR Section 3.0 - Reactor Coolant and Connected Systems................................

3.3.1 TSAR Section 3.1 - Nuclear System Pressure Relief/Overpressure Protection.................................................................................................................

3.3.2 TSAR Section 3.2 - Reactor Vessel.........................................................................

3.3.3 TSAR Section 3.3 - Reactor Internals......................................................................

3.3.4 TSAR Section 3.4 - Flow-Induced Vibration..............................................................

3.3.5 TSAR Section 3.5 - Piping Evaluation......................................................................

3.3.6 TSAR Section 3.6 - Reactor Recirculation System...................................................

3.3.7 TSAR Section 3.7 - Main Steam Line Flow Restrictors.............................................

3.3.8 TSAR Section 3.8 - Main Steam Isolation Valves.....................................................

OFFICIAL USE ONLY PROPRIETARY INFORMATION ii OFFICIAL USE ONLY PROPRIETARY INFORMATION 3.3.9 TSAR Section 3.9 - Reactor Core Isolation Cooling..................................................

3.3.10 TSAR Section 3.10 - Residual Heat Removal System.............................................

3.3.11 TSAR Section 3.11 - Reactor Water Cleanup System..............................................

3.4 TSAR Section 4.0 - Engineered Safety Features......................................................

3.4.1 TSAR Section 4.1 - Containment System Performance...........................................

3.4.2 TSAR Section 4.2 - Emergency Core Cooling Systems (ECCSs)............................

3.4.3 TSAR Section 4.3 - Emergency Core Cooling System Performance........................

3.4.4 TSAR Section 4.4 - Main Control Room Atmosphere Control System......................

3.4.5 TSAR Section 4.5 - Standby Gas Treatment System................................................

3.4.6 TSAR Section 4.6 - Main Steam Isolation Valve Leakage Control System...............

3.4.7 TSAR Section 4.7 - Post-LOCA Containment Atmosphere Control System..............

3.5 TSAR Section 5.0 - Instrumentation and Control......................................................

3.5.1 TSAR Section 5.1 - Nuclear Steam Supply System Monitoring and Control......................................................................................................................

3.5.2 TSAR Section 5.2 - Balance-of-Plant Monitoring and Control..................................

3.5.3 TSAR Section 5.3 - Technical Specification Instrument Setpoints............................

3.5.4 Thermal Power Measurement Uncertainty................................................................

3.6 TSAR Section 6.0 - Electrical Power and Auxiliary Systems.....................................

3.6.1 TSAR Section 6.1 - Alternating Current (AC) Power.................................................

3.6.2 TSAR Section 6.2 - Direct Current (DC) Power........................................................

3.6.3 TSAR Section 6.3 - Fuel Pool...................................................................................

3.6.4 TSAR Section 6.4 - Water Systems..........................................................................

3.6.5 TSAR Section 6.5 - Standby Liquid Control System.................................................

3.6.6 TSAR Section 6.6 - Power-Dependent Heating, Ventilation, and Air Conditioning.............................................................................................................

3.6.7 TSAR Section 6.7 - Fire Protection...........................................................................

3.6.8 TSAR Section 6.8 - Systems Not Affected by TPO Uprate.......................................

3.7 TSAR Section 7.0 - Power Conversion Systems......................................................

3.8 TSAR Section 8.0 - Radwaste and Radiation Sources.............................................

3.8.1 TSAR Section 8.1 - Liquid and Solid Waste Management.......................................

3.8.2 TSAR Section 8.2 - Gaseous Waste Management..................................................

3.8.3 TSAR Section 8.3 - Radiation Sources in the Reactor Core.....................................

3.8.4 TSAR Section 8.4 - Radiation Sources in the Reactor Coolant.................................

3.8.5 TSAR Section 8.5 - Radiation Levels.......................................................................

3.8.6 TSAR Section 8.6 - Normal Operation Off-Site Doses.............................................

OFFICIAL USE ONLY PROPRIETARY INFORMATION iii OFFICIAL USE ONLY PROPRIETARY INFORMATION 3.9 TSAR Section 9.0 - Reactor Safety Performance Evaluations..................................

3.9.1 TSAR Section 9.1 - Anticipated Operational Occurrences.......................................

3.9.2 TSAR Section 9.2 - Design Basis Accidents............................................................

3.9.3 TSAR Section 9.3 - Special Events..........................................................................

3.10 TSAR Section 10.0 - Other Evaluations....................................................................

3.10.1 TSAR Section 10.1 - High Energy Line Break...........................................................

3.10.2 TSAR Section 10.2 - Moderate Energy Line Break...................................................

3.10.3 TSAR Section 10.3 - Environmental Qualification.....................................................

3.10.4 TSAR Section 10.4 - Testing...................................................................................

3.10.5 TSAR Section 10.5 - Operator Training and Human Factors....................................

3.10.6 TSAR Section 10.6 - Plant Life................................................................................

3.10.7 TSAR Section 10.7 - NRC and Industry Communications........................................

3.10.8 TSAR Section 10.8 - Plant Procedures and Programs.............................................

3.10.9 TSAR Section 10.9 - Emergency Operating Procedures..........................................

3.10.10 TSAR Section 10.10 - Individual Plant Examination...............................................

3.11 License and Technical Specification Changes..........................................................

3.11.1 RFOL Paragraph 2.C.(1) - Maximum Power Level...................................................

3.11.2 RFOL Paragraph 2.C.(11) - Partial Feedwater Heating............................................

3.11.3 TS 1.35 - Definitions................................................................................................

3.11.4 TS 2.2 - Limiting Safety System Settings.................................................................

3.11.5 TS Limiting Condition for Operation (LCO) 3.1.3.1 Control Rod Operability, TS 3.1.4.1 LCO Rod Worth Minimizer, and TS 3.10.2 LCO Rod Worth Minimizer................................................................................................

3.11.6 Changes to TS 3.3.6 Control Rod Block Instrumentation..........................................

3.11.7 Changes to TS 3.4.1.1 LCO and TS 4.4.1.1.1.a. SR for the Recirculation System.....................................................................................................................

3.12 Technical Evaluation Conclusion..............................................................................

STATE CONSULTATION........................................................................................

ENVIRONMENTAL CONSIDERATION....................................................................

CONCLUSION.........................................................................................................

APPENDICES Appendix A - List of Acronyms............................................................................................. - A1-

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION INTRODUCTION By application dated July 7, 2017 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML17188A259), as supplemented by letters dated November 1, 2017 (ADAMS Accession No. ML17305B270), November 27, 2017 (ADAMS Accession No. ML17333A853), December 14, 2017 (ADAMS Accession No. ML17348B289),

December 19, 2017 (four letters) (ADAMS Accession Nos. ML17353A744, ML17353A778, ML17353A831, and ML17353A926), December 22, 2017 (ADAMS Accession No. ML17356A139), and January 22, 2018 (ADAMS Accession No. ML18022A147), PSEG Nuclear LLC (PSEG, the licensee), submitted a license amendment request (LAR) for Hope Creek Generating Station (HCGS). The amendment would revise the Renewed Facility Operating License (RFOL) and Technical Specifications (TSs) to implement a measurement uncertainty recapture (MUR) power uprate. Specifically, the amendment would authorize an increase in the maximum licensed thermal power level from 3,840 megawatts thermal (MWt) to 3,902 MWt, which is an increase of approximately 1.6 percent. This increase does not exceed 120 percent of the original licensed thermal power.

The supplemental letters provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the U.S. Nuclear Regulatory Commission (NRC or the Commission) staffs original proposed no significant hazards consideration determination as published in the Federal Register on October 3, 2017 (82 FR 46098). Portions of the supplemental letters dated November 27, December 19 and 22, 2017, contain sensitive unclassified non-safeguards information, and accordingly, have been withheld from public disclosure pursuant to Section 2.390 of Title 10 of Code of Federal Regulations (10 CFR).

1.1 Background

1.1.1 General Site Information The HCGS site is located in Lower Alloways Creek Township, Salem County, New Jersey, on the southern part of Artificial Island on the east bank of the Delaware River. The site is 15 miles south of the Delaware Memorial Bridge, 18 miles south of Wilmington, Delaware, 30 miles southwest of Philadelphia, Pennsylvania, and 7.5 miles southwest of Salem, New Jersey.

The HCGS site is a boiling-water reactor (BWR) plant of the BWR/4 design with Mark I containments. The unit began commercial operation in 1986, and the RFOL expires in 2046.

1.1.2 Previous Power Uprates for Hope Creek Generating Station The NRC issued a full power operating license for HCGS, on July 25, 1986 (ADAMS Accession No. ML011760205), with an original license thermal power level of 3,293 MWt.

By Amendment No. 131 dated July 30, 2001 (ADAMS Accession No. ML011910345), the NRC approved an approximate 1.4 percent MUR power uprate that authorized an increase in the maximum thermal power level from 3,293 MWt to 3,339 MWt.

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION By Amendment No. 174 dated May 14, 2008 (ADAMS Accession No. ML081230540), the NRC approved an approximate 15 percent extended power uprate (EPU) that authorized an increase in the maximum thermal power level from 3,339 MWt to 3,840 MWt.

In addition to the power uprates noted above, by Amendment No. 163 dated February 8, 2006 (not publicly available), the NRC approved a change to the TSs to allow an expanded operating domain resulting from the implementation of the average power range monitor, rod block monitor TSs and maximum extended load line limit analysis (ARTS/MELLLA). The ARTS/MELLLA operating domain increased startup and operating flexibility by updating thermal limits requirements and improving plant instrumentation accuracy.

1.1.3 General Approach for MUR Power Uprates Nuclear power plants are licensed to operate at a specified maximum core thermal power, often called rated thermal power (RTP). Appendix K, [Emergency Core Cooling System] ECCS Evaluation Models, of 10 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 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 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 feedwater (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.

Since substantial conservatism remains 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 LAR, which must be reviewed and approved by the NRC staff.

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, errors can be introduced in the flow rate

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION due to venturi fouling and, to a lesser extent, flow nozzle fouling, the transmitter, and the analog-to-digital converter. To reduce flow instrumentation uncertainty and enable plant operation 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) to be a viable alternative. The UFMs are based on computer-controlled electronic transducers that do not have differential pressure elements that are susceptible to fouling.

The uprate for HCGS will be achieved by the use of improved FW flow and temperature measurement hardware, which reduces the uncertainty associated with measuring those parameters. The licensee includes allowances for FW flow and temperature measurement uncertainty in the plant safety analyses. When the uncertainty is reduced, the licensee is able to increase the HCGS thermal power output by a marginal amount without significant modifications to the plant, and without adverse impact on public health and safety. This type of power uprate, achieved with improved FW measurement hardware, is called an MUR and may also be called a thermal power optimization (TPO).

1.2 Licensees Approach As stated in Section 1.0 of LAR Enclosure 1 Description and Evaluation of the Proposed Change, the TPO request is based on the increased FW flow and temperature measurement accuracy of the Cameron International (formerly Caldon) CheckPlusTM Leading Edge Flow Meter (LEFM) UFM instrumentation which provides a more accurate calculation of reactor thermal power. When the amendment was submitted, the CheckPlusTM LEFM system was not installed at HCGS, but it was scheduled for the spring 2018 refueling outage.

As stated in Section 3.1 of LAR, Enclosure 1, the LEFM system reduces the core thermal power measurement uncertainty to a maximum of 0.374 percent of the MUR uprate power level of 3,902 MWt. As such, the licensee stated this will support an increase in RTP of approximately 1.6 percent (i.e. 2.00 - 0.374), from 3,840 MWt to 3,902.4 MWt, which is conservatively rounded down to the requested value of 3,902 MWt.

Licensees seeking guidance on how to request an MUR power uprate using improved FW flow measurement may refer to the NRCs 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 provided guidance 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 Section 3.1 of Enclosure 1 to its application, that the scope and content of the evaluations performed and described in the LAR are consistent with the guidance contained in RIS 2002-03. Enclosure 4 to PSEGs application provides a cross-reference between the contents of the LAR and the guidance in RIS 2002-03.

In RIS 2002-03 it also states that where the NRC has approved a specific methodology (e.g.,

topical report) for the type of measurement uncertainty recapture power uprate being requested, licensees should follow the format prescribed for that specific methodology and provide the information called for in that methodology and the NRCs letter and safety evaluation (SE) approving the methodology. By letter and SE dated April 1, 2003, the NRC approved a General Electric (GE; now GE-Hitachi) Licensing Topical Report (LTR) that provides a

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION methodology acceptable to the staff for justifying an MUR power uprate for a BWR. The accepted version of this TPO LTR (referred to as the TLTR), NEDC-32938P-A, Revision 2, was issued in May 2003 (ADAMS Accession No. ML17076A201).

to PSEGs application dated July 7, 2017, GE - Hitachi Nuclear Energy (GEH),

Safety Analysis Report for Hope Creek Generating Station Thermal Power Optimization, NEDC-33871P, Revision 0, dated April 2017 (Proprietary information. Not publicly available.),

summarizes the evaluations performed for HCGS, in accordance with the content and format specified in the TLTR. This report is referred to as the TSAR (i.e., Thermal Power Optimization Safety Analysis Report). A public version of the TSAR, GEH report NEDO-33871, is contained in Enclosure 8 to PSEGs application (ADAMS Accession No. ML17188A263).

Enclosures 9 and 11 to PSEGs application dated July 7, 2017 (Proprietary information. Not publicly available.), provide Cameron Documents ER-1123P Bounding Uncertainty Analysis for Thermal Power Determination at Hope Creek Unit 1 Nuclear Generating Station Using the LEFM [Check-Plus] System, Revision 2, and ER-1132P Meter Factor Calculation and Accuracy Assessment for Hope Creek Nuclear Generating Station, Revision 2. Public versions of these assessments are contained in Enclosures 10 and 12 (ADAMS Accession Nos.

ML17188A265 and ML17188A267, respectively). Enclosure 14 is Calculation SC-BB-0525, Hope Creek Heat Balance Uncertainty Calculation, Revision 6 (ADAMS Accession No. ML17188A269). Enclosure 15 is the LEFM Flow Meter Installation Location Drawings (ADAMS Accession No. ML17188A270.

1.3 Method of NRC Review The NRC staff evaluated the licensees assessment of the impact of the proposed MUR on the applicable HCGS design-basis analyses. The NRC staff reviewed the licensees application and supplements.

The NRC staff reviewed the LAR to ensure 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 Commissions 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.

In areas where the licensee and its contractors used NRC-approved or widely accepted methods in performing analyses related to the LAR, the NRC staff reviewed relevant material to ensure that the licensee/contractor used the methods consistent with the limitations and conditions placed on the methods. In addition, the NRC staff considered the effects of the changes in plant operating conditions on the use of these methods to ensure that the methods are appropriate for use at the proposed MUR conditions.

Details of the NRC staffs technical evaluation are provided in Section 3.0 of this SE. The technical evaluation generally follows the format of the technical review areas contained in Sections 2.0 through 10.0 of the TSAR. As noted above, the TSAR was provided in (proprietary, non-public) and Enclosure 8 (public) to the licensees application.

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION REGULATORY EVALUATION 2.1 General Design Criteria The construction permit for HCGS was issued by the NRC on November 4, 1974 (ADAMS Accession No. ML011760627). As stated in Section 3.1 Conformance with NRC General Design Criteria, in the Updated Final Safety Analysis Report (UFSAR) (ADAMS Accession No. ML18053A189), the design bases for HCGS were measured against the 64 General Design Criteria (GDC) for Nuclear Power Plants specified in 10 CFR Section 50, Appendix A. The licensee concluded that HCGS is in compliance with the NRC GDC.

The NRC staff identified the following GDC as applicable to the LAR:

GDC 3, Fire protection, which requires, in part, that that structures, systems, and components (SSCs) important to safety be designed and located to minimize the probability and effect of fires; noncombustible and heat resistant materials be used; and fire detection and fighting systems be provided and designed to minimize the adverse effects of fires on SSCs important to safety.

GDC 4, Environmental and Dynamic Effects Design Bases, which requires, in part, that SSCs important to safety be protected against dynamic effects.

GDC 10, Reactor design, which requires, in part, that the reactor protection system be designed to assure that specified, acceptable fuel design limits are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences (AOOs).

GDC 12, Suppression of reactor power oscillations, which requires that the reactor core and associated coolant, control, and protection systems be designed to assure that power oscillations, which can result in conditions exceeding specified, acceptable fuel design limits are not possible or can be reliably and readily detected and suppressed.

GDC 14, Reactor coolant pressure boundary, which requires that the reactor coolant pressure boundary (RCPB) be designed, fabricated, erected, and tested so as to have an extremely low probability of rapidly propagating fracture and of gross rupture.

GDC 15, Reactor coolant system design, which requires that the reactor coolant system (RCS) and certain associated systems must be designed with sufficient margin to ensure that the design conditions of the RCPB are not exceeded during normal operating conditions, including anticipated operational occurrences.

GDC 16, Containment design, which requires that the reactor containment be provided to establish an essentially leaktight barrier against the uncontrolled release of radioactivity to the environment.

GDC 17, Electric power systems, which requires, in part, 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.

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION GDC 19, Control room, which requires, in part, that 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 5 roentgen equivalent man (rem) whole body, or its equivalent to any part of the body, for the duration of the accident.

GDC 20, Protection System Functions, which requires, in part, that the protection system 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.

GDC 27, Combined reactivity control systems capability, which requires that the reactivity control systems be designed to have a combined capability, in conjunction with poison addition by the emergency core cooling system, of reliably controlling reactivity changes under postulated accident conditions, with appropriate margin for stuck rods, to assure the capability to cool the core is maintained.

GDC 31, Fracture prevention of reactor coolant pressure boundary, which requires, in part, that the RCPB be designed with sufficient margin to assure that when stressed under specified conditions, it will behave in a nonbrittle manner, and the probability of rapidly propagating fracture is minimized.

GDC 33, Reactor coolant makeup, which requires to assure that specified acceptable fuel design limits are not exceeded as a result of reactor coolant loss due to leakage from the reactor coolant pressure boundary and rupture of small piping or other small components which are part of the boundary. The system is required to assure that for onsite electric power system operation (assuming offsite power is not available) and for offsite electric power system operation (assuming onsite power is not available) the system safety function to maintain coolant inventory during normal reactor operation is accomplished.

GDC 34, Residual heat removal, which requires to remove decay heat and other residual heat from the reactor core at a rate such that specified acceptable fuel design limits and the design conditions of the RCPB are not exceeded.

GDC 35, Emergency Core Cooling, which requires that a system to provide abundant emergency core cooling be provided to transfer heat from the reactor core following any LOCA.

GDC 38, Containment heat removal, which requires that a containment heat removal system be provided, and that its function shall be to rapidly reduce the containment pressure and temperature following a LOCA and maintain them at acceptably low levels.

GDC 50, Containment design basis, which requires that the containment and its associated heat removal systems be designed so that the containment structure can accommodate, without exceeding the design leakage rate and with sufficient margin, the calculated temperature and pressure conditions resulting from any loss-of-coolant accident.

GDC 60, Control of releases of radioactive materials to the environment, which requires, in part, that the plant design include means to control the release of radioactive effluents.

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION GDC 61, Fuel storage and handling and radioactivity control, which requires, in part, that systems that contain radioactivity be designed with appropriate confinement.

2.2 Technical Specification Requirements In 10 CFR 50.36, Technical specifications, the NRC established its regulatory requirements related to the content of TSs. Pursuant to 10 CFR 50.36, TSs are required to include items in the following categories: (1) safety limits, limiting safety system settings, and limiting control settings; (2) limiting conditions for operation (LCOs); (3) surveillance requirements (SRs);

(4) design features; and (5) administrative controls. The regulation does not specify the particular requirements to be included in a plants TSs.

As stated in 10 CFR 50.36(c)(1), safety limits for nuclear reactors are limits upon important process variables that are found to be necessary to reasonably protect the integrity of certain of the physical barriers that guard against the uncontrolled release of radioactivity. If any safety limit is exceeded, the reactor must be shut down.

As stated in 10 CFR 50.36(c)(2), LCOs are the lowest functional capability or performance level of equipment required for safe operation of the facility. When LCOs are not met, the licensee shall shut down the reactor or follow any remedial action permitted by the TSs until the LCOs can be met.

As stated in 10 CFR 50.36(c)(3), SRs are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the LCOs will be met.

2.3 Other Regulatory Requirements The NRC staff identified the following regulatory requirements as being applicable to the LAR:

10 CFR Part 20, Standards for Protection Against Radiation, which, in part, establishes requirements for radioactivity in liquid and gaseous effluents released to unrestricted areas and contains limits for occupational and public radiation doses.

10 CFR 50.44, Combustible gas control for nuclear power reactors, which requires, in part, that plants be provided with the capability for controlling combustible gas concentrations in the containment atmosphere.

10 CFR 50.46, Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors, which, in part, establishes standards for the calculation of ECCS accident performance and acceptance criteria for that calculated performance.

10 CFR 50.62, Requirements for reduction of risk from anticipated transients without scram (ATWS) events for light-water-cooled nuclear power plants, which requires, in part, that:

1. Each BWR has an alternate rod injection system that is designed to perform its function in a reliable manner and be independent (from the existing reactor trip system) from sensor output to the final actuation device.

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION

2. Each BWR has a standby liquid control (SLC) system with the capability of injecting into the reactor vessel a borated water solution with reactivity control at least equivalent to the control obtained by injecting 86 gallons per minute (gpm) of a 13 weight-percent sodium pentaborate decahydrate solution at the natural boron-10 isotope abundance into a 251-inch inside diameter reactor vessel.

3. Each BWR has equipment to trip the reactor coolant recirculation pumps automatically under conditions indicative of an ATWS. ATWS is defined as an AOO followed by the failure of the reactor trip portion of the protection system.

10 CFR 50.63, Loss of all alternating current power, which requires, in part, that the plant withstand and recover from a station blackout (SBO) event of a specified duration.

10 CFR 50.67, Accident source term, which, in part, sets limits for the radiological consequences of a postulated design-basis accident (DBA) using an alternative source term.

The NRC approved a full scope implementation of an alternative source term methodology for HCGS by License Amendment No. 134 on October 3, 2001 (ADAMS Accession No. ML012600176).

10 CFR Part 50, Appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants, which provides quality assurance requirements for the design, fabrication, construction, and testing of SSCs.

10 CFR Part 50, Appendix G, Fracture Toughness Requirements, provides fracture toughness requirements for ferritic materials in the RCPB, including requirements on the upper-shelf energy (USE) values used for assessing the safety margins of the reactor vessel materials against ductile tearing and requirements for calculating pressure-temperature (P-T) limits for the plant. These P-T limits are established to ensure the structural integrity of the ferritic components of the RCPB during any condition of normal operation, including anticipated operational occurrences and hydrostatic tests.

10 CFR Part 50, Appendix H, Reactor Vessel Material Surveillance Program Requirements, provides for monitoring changes in the fracture toughness properties of materials in the reactor vessel beltline region.

10 CFR Part 50, Appendix I, Numerical Guides for Design Objectives and Limiting Conditions for Operation to Meet the Criterion As Low as is Reasonably Achievable for Radioactive Material in Light-Water-Cooled Nuclear Power Reactor Effluents, which, in part, sets numerical guides to meet the as low as is reasonably achievable criterion.

10 CFR Part 50, Appendix K, ECCS Evaluation Models, which, in part, establishes required and acceptable features of evaluation models for heat removal by the ECCS after the blowdown phase of a LOCA.

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION 10 CFR Part 50, Appendix R, Fire Protection Program for Nuclear Power Facilities Operating Prior to January 1, 1979, require the development of a fire protection program to ensure, among other things, the capability to safely shut down the plant.

2.4 Guidance Documents The guidance that the NRC staff considered in its review of this LAR included the following:

NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition (hereinafter referred to as the SRP) (ADAMS Accession No. ML070660036).

Regulatory Guidance (RG) 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence (ADAMS Accession No. ML010890301), describes methods and assumptions acceptable to the NRC staff for determining the pressure vessel neutron fluence with respect to the GDC contained in Appendix A to 10 CFR Part 50.

RG 1.54, Revision 3, Service Level I, II, and III Protective Coatings Applied to Nuclear Power Plants, dated September 7, 2016 (ADAMS Accession No. ML16070A091).

RG 1.82, Revision 4, Water Sources for Long-Term Recirculation Cooling Following a Loss-of-Coolant Accident, dated March 2012 (ADAMS Accession No. ML111330278).

RG 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials, dated May 1988 (ADAMS Accession No. ML003740284), provides guidance on the general procedures acceptable to the NRC staff for calculating the effects of neutron radiation embrittlement of the low-alloy steels used for light-water reactor vessels.

RG 1.105, Revision 3, Setpoints for Safety-Related instrumentation, dated December 1999 (ADAMS Accession No. ML993560062), describes a method acceptable to the NRC staff for complying with the NRC regulations for assuring that setpoints for safety-related instrumentation are initially within and remain within the TS limits.

RG 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, dated July 2000 (ADAMS Accession No. ML003716792).

RG 1.189, Fire Protection for Nuclear Power Plants, Revision 2 (ADAMS Accession No. ML092580550), describes an approach that is acceptable to the NRC to meet the regulatory requirements of 10 CFR Section 50.48, Fire Protection.

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, dated August 2006, discusses issues that could occur during testing of limiting safety system settings and, therefore, may have an adverse effect on equipment operability (ADAMS Accession No. ML051810077).

RIS 2007-21, Adherence to Licensed Power Limits, Revision 1, and the Nuclear Energy Institute (NEI) Position Statement, Guidance to Licensees on Complying with the Licensed Power Limit, dated June 23, 2008 (ADAMS Accession No. ML081750537). The NEI Position Statement was endorsed by NRC SE, Safety Evaluation Regarding Endorsement of NEI

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION Guidance for Adhering to the Licensed Thermal Power Limit, issued on October 8, 2008 (ADAMS Accession No. ML082690105).

RIS 2014-11 Information on Licensing Applications for Fracture Toughness Requirements for Ferritic Reactor Coolant Pressure Boundary Components, dated October 2014 (ADAMS Accession No. ML14149A165), provides guidance to addressees on the scope and detail of information that should be provided in reactor vessel fracture toughness and associated P-T limits licensing applications to facilitate an NRC staff review.

The Human Factors review used the guidance in Chapter 13 of the Standard Review Plan (SRP) as supplemented by information in:

NUREG-0700, Revision 2, Human-System Interface Design Review Guidelines, dated May 2002, provides the guidelines for the NRC staffs review of the physical and functional characteristics of human-system interfaces (ADAMS Accession No. ML021700373).

NUREG-0711, Revision 3, Human Factors Engineering Program Review Model, provides the methodology for the NRC staffs review of human factors engineering programs (ADAMS Accession No. ML12324A013).

NUREG-1764, Revision 1, Guidance for the Review of Changes to Human Actions, dated September 30, 2017, provides guidance for NRC staff for the level of review for LARs (ADAMS Accession No. ML072640413).

Generic Letter (GL) 89-08, Erosion/Corrosion-Induced Pipe Wall Thinning, dated May 2, 1989 (ADAMS Accession No. ML031200731), provides expectations for licensees to ensure that flow-accelerated corrosion (FAC) will be monitored to prevent failure of piping.

GL 89-10, Consideration of Valve Mispositioning in Pressurized-Water Reactors, dated June 28, 1989 (ADAMS Accession No. ML031150300), provides licensees with information on ensuring the operability of safety-related motor-operated valves (MOVs).

GL 89-16, Installation of a Hardened Wetwell Vent, dated September 1, 1989 (ADAMS Accession No. ML031140220), provided licensees with NRC staff expectations concerning plant modifications to enhance the capability to prevent and mitigate the consequences of severe accidents.

GL 95-07, Pressure Locking and Thermal Binding of Safety-Related Power-Operated Gate Valves, dated August 17, 1995 (ADAMS Accession No. ML031070145), requesting licensees confirm that safety-related, power-operated gate valves will perform their function considering the susceptibility to pressure locking and thermal binding.

GL 96-03, Relocation of the Pressure Temperature Limit Curves and Low Temperature Overpressure Protections System Limits, dated January 31, 1996 (ADAMS Accession No. ML031110004), provides guidance to licensees on an acceptable method to relocate the P-T limit curves to a licensee-controlled document.

GL 96-05, Periodic Verification of Design-Bases Capability of Safety-Related Motor-Operated Valves, dated September 18, 1996 (ADAMS Accession No. ML031110010), requested actions to be taken by licensees to ensure the operability of safety-related MOVs.

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION GL 96-06, Assurance of Equipment Operability and Containment Integrity During Design-Bases Accident Conditions, dated September 30, 1996 (ADAMS Accession No. ML031110021),

requested licensee actions as appropriate to address safety-significant impacts due to water hammer and two-phase flow during accident conditions.

SECY-11-0014, Enclosure 1, The Use of Containment Accident Pressure in Reactor Safety Analysis, dated January 31, 2011 (ADAMS Accession No. ML102110167), provides guidance to licensees on an acceptable way to use containment accident pressure in reactor safety analyses.

NRC Order EA-13-109 Issuance of Order to Modify Licenses with Regard to Reliable Hardened Containment Vents Capable of Operation Under Severe Accident Conditions, dated June 6, 2013 (ADAMS Accession No. ML13143A321), requested licensees to modify plants to ensure the continued operability of containment during accident conditions.

TECHNICAL EVALUATION 3.1 Overview of the Thermal Power Optimization Safety Analysis Report (TSAR)

As stated in SE Section 1.2, the TSAR summarizes the plant-specific evaluations performed to support the proposed MUR power uprate for HCGS. The TSAR contains information divided into the following sections:

TSAR Section 1.0 - Introduction TSAR Section 2.0 - Reactor Core and Fuel Performance TSAR Section 3.0 - Reactor Coolant and Connected Systems TSAR Section 4.0 - Engineered Safety Features TSAR Section 5.0 - Instrumentation and Control TSAR Section 6.0 - Electrical Power and Auxiliary Systems TSAR Section 7.0 - Power Conversion Systems TSAR Section 8.0 - Radwaste and Radiation Sources TSAR Section 9.0 - Reactor Safety Performance Evaluations TSAR Section 10.0 - Other Evaluations TSAR Section 11.0 - References The TSAR also contains Appendices A and B, which address continued applicability of the limitations and conditions, described in the NRC SEs for two GEH LTRs at MUR RTP conditions, including plant conditions based on the previously approved extended power uprate (EPU) and ARTS/MELLLA amendments. The limitations and conditions addressed in TSAR Appendices A and B, apply to the following NRC-approved LTRs, respectively (note that the versions provided below incorporate the NRC staffs SEs):

NEDC-33173P-A, Revision 4, Applicability of GE Methods to Expanded Operating Domains, dated November 7, 2012 (ADAMS Accession No. ML123130130). This LTR is referred to as the Methods LTR.

NEDC-33075P-A, Revision 8, General Electric Boiling Water Reactor Detect and Suppress Solution-Confirmation Density (DSS-CD LTR); this is the proprietary version.

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION A redacted, publicly available version is titled NEDO-33075-A, Revision 8, and is found at ADAMS Accession No. ML13324A099).

As stated in SE Section 1.3, the NRC staffs technical evaluation generally follows the format of the technical review areas contained in Sections 2.0 through 10.0 of the TSAR. These technical review areas are addressed in SE Sections 3.2 through 3.10.

3.2 TSAR Section 2.0 - Reactor Core and Fuel Performance The reactor core and fuel system consists of arrays of fuel rods, burnable poison rods, spacer grids and springs, end plates, channel boxes, and reactivity control rods. The NRC staff reviewed the mechanical and nuclear design of the fuel system and the thermal and hydraulic design of the core. The NRC staff reviewed the fuel system mechanical design to ensure that (1) the fuel system is not damaged as a result of normal operation and AOOs, (2) fuel system damage is never so severe as to prevent control rod insertion when it is required, (3) the number of fuel rod failures is not underestimated for postulated accidents; and (4) the coolable geometry of the reactor core is always maintained. The NRC staff reviewed the nuclear, thermal and hydraulic design to confirm that the design (1) has been accomplished using acceptable analytical methods, (2) is equivalent to or a justified extrapolation from proven designs, (3) provides acceptable margins of safety from conditions which would lead to fuel damage during normal reactor operation and AOOs, and (4) is not susceptible to thermal-hydraulic instability.

The NRC staffs review covered fuel system damage mechanisms, limiting values for important parameters, and performance of the fuel system during normal operation, AOOs, and postulated accidents. Further, the review covered core power distribution, reactivity coefficients, reactivity control requirements and control provisions, control rod patterns and reactivity worths, criticality, burnup, and vessel irradiation. The review also covered hydraulic loads on the core and RCS components during normal operation and DBA conditions and core thermal-hydraulic stability under normal operation and ATWS events.

The licensee provided an assessment of the MUR that was largely consistent with the TLTR with additional discussion focusing on the applicability of the Methods LTR. The general approach followed by the licensee was to apply generic dispositions in specific areas where possible and to limit new analyses to those topics for which the new thermal power level was not bounded by existing safety analyses. This approach is consistent with that described in RIS 2002-03.

3.2.1 TSAR Section 2.1 - Fuel Design and Operation The licensee stated that the reactor core and fuel design are adequate for the proposed MUR.

To support this claim, the licensee provided tables and plots of key parameters for the proposed MUR core design. The licensee also provided, for comparison, parameters associated with recent HCGS operating cycles, as well as with other plants in the GEH experience base. Refer to Table 2-2 and Figures 2-1 through 2-17 of the TSAR.

The information described above is provided in accordance with Limitation 24 of the Methods LTR. Section 8.5 of the NRC staff SE approving the Methods LTR, notes that the core thermal-hydraulic conditions for operation in expanded power-to-flow operating domains, including

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION uprated core thermal power, can be gauged by the review of key parameters including peak bundle power, coolant flow, and exit void fraction; exit void fractions for the peak channel (which may not be the same as the peak-power bundle) and core-wide average.

In the NRC staffs SE for the Methods LTR, the NRC staff noted that reviewing these parameters would provide insight into the plant-specific core conditions relative to the existing experience base. This information, in turn, enables the NRC staff to confirm that plant-specific applications of the GEH fuel design methods are within the demonstrated qualification base.

This verification effort is necessary to determine whether the computational design and analysis methods are being used in such a way that their results are valid, and that uncertainties, which are quantified on a generic basis, are applicable to those results. This information, in turn, provides the requisite assurance that HCGS, as operated in MUR uprate conditions, will meet the GDC pertinent to the fuel design.

This verification effort is simple when the parameters for the plant in question fall well within those associated with other plants and operating cycles. However, the HCGS-specific trend lines in the following LAR figures sit noticeably higher than most, if not all, of the other trend lines shown.

Figure 2-1, Power of Peak Bundle versus Cycle Exposure Figure 2-3, Exit Void Fraction for Peak Power Bundle versus Cycle Exposure Figure 2-4, Maximum Channel Exit Void Fraction versus Cycle Exposure The figures identified above show only one plant with peak bundle power as high as those predicted for HCGS TPO conditions. Also, the predicted exit void fractions for both the peak power bundle and the maximum channel approach 89 percent, which is noticeably higher than any of the other trend lines shown in Figures 2-3 and 2-4. The information shown in these figures suggests that the TPO conditions expected for HCGS exceed the operating plant experience base shown. ((

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The fact that HCGS is predicted to operate at more severe conditions than those shown for the other plants does not mean that these conditions are unacceptable. Rather, it indicates that a more detailed review is needed in order to confirm that the methods are applicable. The NRC staff performed such a review, addressing the following topics: (1) the qualification of GEH analytic methods at high-power, high void fraction conditions; (2) the effect that operating at high-power, high void fraction conditions has on the predictive capability of the methods; and (3) the impact that a significant increase in predicted core exit void fraction have on the dispositions available in the TLTR.

In its letter dated November 27, 2017, the licensee provided the following information:

Reasoning for the higher maximum bundle exit void fraction with the introduction of the GNF2 fuel than associated with the EPU review

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION Justification for the application of the nuclear and thermal-hydraulic design uncertainties in the GESTAR-II analytic process without the penalties required of plants that operate in the high bundle power, high maximum channel exit void fraction conditions indicated for HCGS Evidence that the dispositions contained in the TLTR relative to fuel design, operation, and thermal margin assessment are applicable in light of the significantly different operating conditions between EPU implementation, GNF2 introduction, and the MUR analysis The licensee summarized several design characteristics of GNF2 fuel, which tend to enable the fuel to operate in higher steady-state void conditions than GE14 fuel; this information explained the general trends shown in the TSAR figures. The licensee also showed cycle operating characteristics for a plant operating with a cycle length and power-to-flow domain that are similar to HCGS, indicating that there is similar operating experience to that shown for HCGS, and that at least at one other plant with GE14 fuel is presently operated in bundle exit void conditions that are similar to those associated with the HCGS MUR cycle design.

The NRC staff evaluated the licensees description of the GNF2 fuel design characteristics by comparing it to the information described in the NRC staff SE approving the Methods LTR that addresses GNF2 conformance to the conditions and limitations associated with the Methods LTR. The NRC staff concluded that the licensees description was similar to the information considered in the NRC staff SE. For example, in the SE, the NRC staff qualified maximum channel exit void fraction as being in the range of 89 to 92 percent. The licensees November 27, 2017, supplement showed that HCGS TPO conditions are expected to remain below 90 percent maximum channel exit void fraction.

As stated above, the channel exit void conditions in particular provide a means to assess the application of GEH analytic methods. The Methods LTR SE states that, because there has been a historical lack of high void data against which to qualify the analytic methods, operation in these conditions may challenge the validity of the uncertainties that are determined based on the qualification data. In certain conditions (e.g., plants using MELLLA+), the NRC staff has previously imposed interim penalties on safety and operating limits to ensure that any potential for increased analytic uncertainty is bounded, pending GEHs submittal of additional qualification data in the high-void conditions. However, the licensees supplemental response dated November 27, 2017, confirmed that the HCGS operating conditions, while extending beyond the generic qualification data set shown in the TSAR figures, remains within the acceptable conditions for both uprated plants and for plants using GNF2 fuel.

Thus, the information provided by the licensee, when evaluated in the context of the Methods LTR supplements that apply to GNF2 fuel, demonstrated that the HCGS TPO will be accomplished with operating characteristics that are in the range of conditions exhibited in current operating experience, and considered in the Methods LTR review. Based on these two considerations, the NRC staff determines that the proposed MUR is acceptable with regard to disposition of the Methods LTR limitations and conditions. More specifically, the additional information provided by the licensee on November 27, 2017, indicated that the generic uncertainties applied in the GEH fuel design and safety analysis methods are appropriate for application to HCGS given implementation of the proposed MUR.

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION The licensee also provided a comparison of pre-and post-MUR transient analysis results in its November 27, 2017, response, which is addressed in Section 3.2.2 of this SE.

In summary, based on the NRC staffs evaluation of the material provided by the licensee, including the LAR, TSAR, and the request for additional information response, the NRC staff determines that the proposed uprate is consistent with the TLTR and Methods LTR conditions and limitations. Based on this consistency, the NRC staff concludes that the proposed MUR is acceptable with respect to fuel design and operation.

3.2.2 TSAR Section 2.2 - Thermal Limits Assessment This section of the TSAR addresses the safety limit minimum critical power ratio (SLMCPR), the operating limit MCPR (OLMCPR), the maximum average planar linear heat generation rate (MAPLHGR) and maximum linear heat generation rate (LHGR) limits, and the power-to-flow ratio.

The SLMCPR and OLMCPR work together to meet the intent of GDC 10. The SLMCPR is a specified acceptable fuel design limit (SAFDL) that provides assurance that 99.9 percent of the fuel rods in the core remain in nucleate boiling. If the fuel rods remain in nucleate boiling (i.e.,

do not depart from nucleate boiling), there is assurance that the fuel cladding is cooled adequately enough to maintain its mechanical integrity. The SLMCPR includes uncertainty allowances for the correlation used to determine the MCPR, the instruments used to infer the core MCPR while operating, and the interplay of pin-and bundle-wise radial power peaking, which cannot be readily measured. The OLMCPR is a margin that is determined from the reactor transient analyses and applied to the SLMCPR to provide margin for AOOs. As long as the core operates above the OLMCPR, there is reasonable assurance that the SAFDL is met with adequate margin to allow for AOOs, as GDC 10 specifies.

The NRC staffs assessment of the acceptability of the proposed MUR is driven by two key considerations. First, the licensee stated that the impacts on both SLMCPR and OLMCPR are minimal for the MUR as evidenced by substantial operating experience demonstrating that the effects of such uprates are small. Second, the NRC staff is aware that the licensee uses NRC-approved safety analysis methods to determine or confirm these parameters on a cycle-specific basis prior to the start of each cycle.

In its review, the NRC staff observed that the thermal-hydraulic conditions predicted to exist for the MUR appear to be noticeably different, in some cases, than those reviewed during the EPU amendment request. This issue is stated in greater detail in SE Section 3.2.1. Based on these observed differences, the NRC staff requested that the licensee provide additional evidence to confirm that the effect of the MUR is as minimal as described in the TLTR.

In its response dated November 27, 2017, the licensee provided a table of changes in the critical power ratio results for the two potentially limiting pressurization transients, the turbine trip with no bypass, and the generator load rejection with no bypass. These events were analyzed at mid-cycle and end-of-cycle (EOC), in both full-power, rated core flow and increased core flow conditions. The results of the events, given the various initial conditions, did not vary from pre-to post-MUR conditions more than +/-0.0124. The NRC staff concludes that this result demonstrates that the effect of the MUR on the transient analyses is ((

)).

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION The LHGR limits (i.e., MAPLHGR and maximum LHGR) are established for several reasons.

The MAPLHGR limits ensure that the plant operates within the bounds established by the safety analysis. Primarily, in the case of MAPLHGR, the applicable safety analysis is the ECCS evaluation, because the MAPLHGR has a direct impact on the predicted peak cladding temperature (PCT), which is a key result of the ECCS evaluation. The maximum LHGR limit ensures that the fuel does not melt as a result of normal operation or AOOs. The LHGR limits ensure that the plant is operated in compliance with the acceptance criteria contained in 10 CFR 50.46, and with the requirements of GDC 10.

The licensee noted that the LHGR limits are fuel dependent, and the MAPLHGR limits are determined in accordance with the ECCS evaluation. The NRC staff determined that the licensee is correct in that LHGR limits are fuel dependent; the MUR effect on ECCS performance is evaluated in SE Section 3.4.3.

3.2.3 TSAR Section 2.3 - Reactivity Characteristics The reload core analysis ensures that the minimum shutdown margin requirements are met for each core design. All minimum shutdown margin requirements apply to cold shutdown conditions and are maintained without change. Checks of cold shutdown margin based on SLCS boron injection capability and shutdown using control rods with the most reactive control rod stuck out are made for each reload. The TPO uprate has no significant effect on these conditions; the shutdown margin is confirmed in the reload core design.

The licensee stated that all minimum shutdown margin requirements apply to cold shutdown conditions and are maintained without change. Further, the shutdown margin is confirmed in the reload design. This information is consistent with the disposition provided in TLTR Section 5.7.1. In the SE approving the TLTR, the NRC staff concluded that, since the reload analysis assures that adequate shutdown margin is maintained during any particular fuel cycle, a plant-specific evaluation supporting an MUR requested in accordance with the TLTR would not be necessary.

Based on the above considerations, the NRC staff determines that the disposition for reactivity characteristics was in accordance with the TLTR and hence the proposed MUR is acceptable with respect to reactivity characteristics.

3.2.4 TSAR Section 2.4 - Thermal Hydraulic Stability Detect and Suppress Solution - Confirmation Density To ensure that reactor core instability events are readily detected and suppressed per the specifications of GDCs 10 and 12, the licensee stated in TSAR Section 2.4 that it uses the DSS-CD long term stability solution consistent with the DSS-CD LTR (ADAMS Accession No. ML13324A097). The NRC issued Amendment No. 206 to the HCGS facility operating license on August 4, 2017 (ADAMS Accession No. ML17216A022), which approved implementation of the DSS-CD stability solution.

The licensee was approved to implement DSS-CD with a higher oscillation power range monitor (OPRM) amplitude discriminator setpoint than the generic value, following the process delineated in the DSS-CD LTR. For the MUR application, the licensee provided updated,

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION plant-specific information in TSAR Table 2-6. The analysis was performed using a conservative approach to bound applicable uncertainties, as set forth in NRC-approved LTR NEDE-33147P-A, Revision 4, DSS-CD TRACG Application dated August 12, 2013 (ADAMS Accession No. ML13224A319).

The licensee information confirmed that the primary DSS-CD algorithm provides a reactor trip with adequate margin to the SLMCPR in the event of anticipated transients that lead to unstable conditions. In addition, the licensee provided updated applicability information for the DSS-CD LTR in TSAR Tables 2-4 and 2-5. The NRC staff reviewed the information provided by the licensee and confirmed that (1) the applicability envelope for DSS-CD continues to cover the uprated conditions; (2) the plant-specific analytic results demonstrate adequate margin to prevent the onset of SAFDL-challenging instabilities; and (3) the supporting analyses were accomplished in accordance with NRC-approved methods. Based on these considerations, the NRC staff determined that the proposed MUR is acceptable with respect to its effect on reactor stability, and on conformance with GDCs 10 and 12, insofar as they pertain to reactor stability.

Thermal Limits Monitoring Threshold

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OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION

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Armed Region The licensee stated that the OPRM armed region would be maintained consistent with the existing TLMT. The NRC staff noted that this approach is consistent with the DSS-CD LTR, wherein Table 8-1 indicates that the applicable mode for requiring OPRM system operability should be a plant-specific value that is 5 percent below the lower boundary of the DSS-CD Armed Region. At HCGS, this value is 19 percent, which is 5 percent below the TLMT. The adequacy of the TLMT is evaluated in the Thermal Limits Monitoring Threshold section of this SE. Since the licensee is consistent with the DSS-CD LTR, the NRC staff determined that the proposed MUR is acceptable with respect to the OPRM armed region.

Backup Stability Protection (BSP)

The licensee stated that BSP would be implemented in accordance with the DSS-CD LTR.

Furthermore, the adequacy of the BSP approach, the manual BSP regions, and the automatic BSP setpoints are all confirmed on a cycle-specific basis and re-established as necessary in accordance with the DSS-CD LTR. In evaluating this disposition, the NRC staff considered that the licensees maximum rod line does not change as a result of the proposed uprate. Thus, while the severity of potential power oscillations if not suppressed may increase marginally as a result of the proposed uprate, the susceptibility does not appreciably increase because the maximum rod line remains fixed.

Therefore, the NRC staff concludes that the proposed MUR is acceptable with respect to BSP because the applicable approach is consistent with the DSS-CD LTR methodology and setpoints are within the scope of the cycle-to-cycle reload analysis.

3.2.5 TSAR Section 2.5 - Reactivity Control The TLTR establishes that the effect of an MUR uprate is negligible on the control rod drive system, and that although the increased power level will have a small effect on control blade lifetime, shutdown margin capability is confirmed in each cycle reload evaluation. The licensee stated in TSAR Section 2.5 that HCGS is consistent with the TLTR in this regard; therefore, the licensee concluded that no further assessment of the effect of an MUR on reactivity control is necessary. Based on the consistency between the TSAR and the generic disposition provided in the TLTR, the NRC staff concludes that the proposed MUR is acceptable with respect to reactivity control.

3.2.6 TSAR Section 2.6 - Additional Limitations and Conditions Related to Reactor Core and Fuel Performance As stated in TSAR Section 2.6, the licensee addressed limitations and conditions in the NRC staffs SE for the Methods LTR relating to the reactor core and fuel design.

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION TGBLA/PANAC Version The Methods LTR Limitation and Condition 9.1 states that the most recent TGBLA and PANAC computer code versions will be used for cycle-specific analyses at TPO uprated conditions. The licensee stated that HCGS will adhere to the Methods LTR limitation and condition requiring use of the latest versions of the TGBLA and PANAC code system. This approach is consistent with the Methods LTR; therefore, this limitation and condition is satisfied.

LHGR and Exposure Qualification The Methods LTR Limitation and Condition 9.12 states that once the PRIME computer code and its application are approved by the NRC, future license applications should Reference these thermal-mechanical methods. The licensee stated that the HCGS fuel thermal-mechanical analyses are performed using the NRC-approved PRIME methodology. This approach is consistent with the Methods LTR Limitation and Condition on the phase-out of the legacy code GESTR-M and implementation of the PRIME methodology. This approach is consistent with the Methods LTR; therefore, this limitation and condition is satisfied.

3.3 TSAR Section 3.0 - Reactor Coolant and Connected Systems The following provides the NRC staffs technical review of the topics in Section 3.0 of the TSAR.

3.3.1 TSAR Section 3.1 - Nuclear System Pressure Relief/Overpressure Protection As described in UFSAR Section 5.2.2, the overpressure protection for the reactor coolant pressure boundary (RCPB) is provided by the nuclear pressure relief system, which includes the following:

a) main steam line safety relief valves (SRVs),

b) SRV discharge lines and T-quenchers, c) discharge line vacuum relief valves, and d) check valves and pneumatic accumulators.

The nuclear pressure relief system protects the RCPB from damage due to overpressure during abnormal operational transients. To protect against overpressurizing the system, the main steam SRVs are provided to discharge steam from the reactor pressure vessel (RPV) to the suppression pool. As a part of the nuclear pressure relief system, the automatic depressurization system (ADS) also acts to automatically lower pressure in the nuclear steam supply system (NSSS) in the event of a LOCA in which the high pressure coolant injection (HPCI) system fails to maintain RPV water level. The depressurization of the NSSS by the ADS allows the low pressure core cooling systems to inject enough cooling water to adequately cool the fuel.

The generic evaluation contained in the TLTR, is applicable at the TPO uprate power because of the following:

a) The methodology is not changed.

b) The analysis of record performed at a power level greater than 102 percent CLTP

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION demonstrated that the RPV conforms to the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (ASME Code) and TS requirements.

c) The nominal operating pressure of the RCPB is not increased.

d) SRV setpoints are not changed.

e) SRV out-of-service options are not changed.

f) The limiting overpressure event is the same.

g) The fuel re-load analysis for each re-load confirms the capability of the system to meet the ASME Code design criteria.

Based on the above evaluation, the nuclear system pressure relief and overpressure protection system continues to meet the requirements of GDC 15 because the design conditions of the RCPB are not exceeded during normal operating conditions at the TPO uprate power.

3.3.2 TSAR Section 3.2 - Reactor Vessel The NRC staffs review of the proposed MUR power uprate regarding RPV integrity addressed:

(1) neutron fluence calculations, (2) RPV material surveillance program, (3) RCS P-T limits, (4) USE evaluation, (5) RPV circumferential and axial welds, and (6) RPV structural evaluation.

Neutron Fluence Calculations The guidance provided in RG 1.190 indicates that the following elements comprise an acceptable fluence calculation: (1) determination of the geometrical and material input data, (2) determination of the core neutron source, (3) propagation of the neutron fluence from the core to the vessel and into the cavity, and (4) qualification of the calculational procedure. The NRCs review was performed to establish that these elements of the calculational method adhere to the regulatory positions set forth in RG 1.190.

Reactor Pressure Vessel Material Surveillance Program The RPV material surveillance program provides a means for monitoring the fracture toughness of RPV beltline materials to support analyses for ensuring the structural integrity of the RPV.

The NRC staffs review addressed the effects of the proposed MUR on the licensees RPV surveillance capsule withdrawal schedule. The NRCs acceptance criteria for RPV material surveillance programs are based on: (1) GDC 14, which requires that the RCPB be designed, fabricated, erected, and tested so as to have an extremely low probability of rapidly propagating fracture and of gross rupture; (2) GDC 31, which requires, in part, that the RCPB be designed with sufficient margin to assure that when stressed under specified conditions, it will behave in a non-brittle manner, and the probability of rapidly propagating fracture is minimized; (3) 10 CFR Part 50, Appendix H, which establishes requirements for monitoring changes in the fracture toughness properties of materials in the RPV beltline region; and (4) 10 CFR 50.60, which, in part, requires compliance with the material surveillance program requirements for the RCPB set forth in 10 CFR Part 50, Appendix H.

As an alternative to a plant-specific RPV material surveillance program, 10 CFR Part 50, Appendix H, allows for the implementation of an integrated surveillance program (ISP). An ISP

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION is defined in 10 CFR Part 50, Appendix H, as occurring when the representative materials chosen for surveillance for a reactor are irradiated in one or more reactors that have similar design and operating features.

The licensee stated the impact of the proposed MUR on the HCGS RPV material surveillance program in TSAR Section 3.2.1. The licensee identified that HCGS actively participates in the Boiling-Water Reactor Vessel and Internals Project (BWRVIP) ISP. The licensee stated that prior to the ISP, the original surveillance program consisted of three capsule holders, each containing six capsules, located at the 30, 120 and 300 degree azimuthal locations in HCGS.

The licensee identified that flux wires from the first surveillance capsule were removed from a 30 degree capsule and tested at 6.0 effective full power years (EFPY). A second capsule, from the 120 degree location, was removed and flux wires were tested at 24.1 EFPY. Both of these capsules were tested prior to the implementation of the BWRVIP ISP. The licensee indicated that HCGS is designated as a host plant for capsules under the ISP; there is one additional representative capsule that will be removed and tested around 2036 in accordance with the ISP withdrawal schedule. The licensee determined that the increase in the projected neutron fluence for the proposed MUR will not affect the existing ISP capsule withdrawal schedule or continued ISP implementation. Electric Power Research Institute (EPRI) Topical Report, BWRVIP-86NP, Revision 1-A, dated May 2013 (ADAMS Accession No. ML13176A097),

establishes the ISP requirements for RPV materials (base metal and weld metal) in all operating BWRs for the first 40-year operating period and for the first 20-year period of extended operation. The NRCs final SE for BWRVIP-86NP, Revision 1-A, dated October 20, 2011 (ADAMS Accession No. ML112780511), documents that the BWRVIP ISP is compliant with the ISP requirements established in 10 CFR Part 50, Appendix H, for the original 40-year license terms and the first 20-year license renewal (LR) terms.

The BWRVIP ISP provides for a number of surveillance capsules to be removed from specified BWRs and to be available for testing during the LR period for the BWR fleet. The ISP establishes acceptable technical criteria for capsule withdrawal and testing. The NRC staff identified that HCGS is a designated host plant for surveillance capsules under the BWRVIP ISP. The NRC staff verified that the small increase in neutron fluence for MUR conditions will not invalidate the estimated projected irradiation of the representative ISP capsule materials being hosted at HCGS at the specified future withdrawal times, as approved by the staff in BWRVIP-86NP, Revision 1-A. Therefore, the staff determined that the licensees ISP capsule withdrawal schedule for HCGS will remain acceptable under MUR conditions. The NRC staff identified that ISP capsule materials will continue to be appropriately handled and tested in accordance with the BWRVIP ISP criteria, as approved by the NRC, for meeting the requirements of 10 CFR Part 50, Appendix H. Therefore, the NRC staff determined that the licensees RPV material surveillance program for HCGS is acceptable for satisfying the requirements of 10 CFR Part 50, Appendix H, for MUR conditions.

Reactor Coolant System Pressure-Temperature Limits Appendix G to 10 CFR Part 50 provides fracture toughness requirements for ferritic materials in the RCPB, including requirements for the Charpy USE for protecting RPV beltline materials against non-brittle failure and requirements for calculating RCS P-T limits for protection against brittle fracture. The RCS P-T limits are specifically established to ensure the structural integrity of the ferritic components of the RCPB (in particular the RPV) during any condition of normal operation, including AOOs and hydrostatic tests. The NRC staffs review of USE and P-T limits addressed the licensees current licensing basis (CLB) methodologies for USE (or the mandated

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION alternative analyses stated below) and P-T limits, and its plant-specific evaluation for demonstrating that 10 CFR Part 50, Appendix G requirements will continue to be satisfied following implementation of the proposed MUR, specifically considering neutron embrittlement of RPV beltline materials for MUR conditions.

The NRCs acceptance criteria for USE and P-T limits are based on: (1) GDC 14, which requires that the RCPB be designed, fabricated, erected, and tested so as to have an extremely low probability of rapidly propagating fracture and of gross rupture; (2) GDC 31, which requires, in part, that the RCPB be designed with sufficient margin to assure that when stressed under specified conditions, it will behave in a non-brittle manner, and the probability of rapidly propagating fracture is minimized; (3) 10 CFR Part 50, Appendix G, which provides fracture toughness requirements for ferritic materials in the RCPB; and (4) 10 CFR 50.60, which, in part, requires compliance with the fracture toughness requirements for the RCPB set forth in 10 CFR Part 50, Appendix G.

Section IV.A.1 of 10 CFR Part 50, Appendix G, provides requirements for maintaining acceptable levels of USE for RPV beltline materials throughout the licensed operating terms of nuclear power reactors. The rule requires that RPV beltline materials have Charpy USE in the transverse direction for base material and along the weld for weld material greater than or equal to 75 foot-pounds (ft-lbs) in the unirradiated condition. The rule also requires that RPV beltline materials must maintain Charpy USE greater than or equal to 50 ft-lbs throughout the operating life of the RPV, unless it is demonstrated in a manner approved by the NRC that lower values of USE would provide margins of safety against fracture equivalent to those required by the ASME Code,Section XI, Appendix G. The analysis to demonstrate acceptable margins of safety against fracture is generally referred to as an equivalent margins analysis (EMA). The rule also requires that the methods used to calculate projected USE values or perform EMAs must account for the effects of neutron radiation on the USE values or EMA results for the materials and must incorporate any credible RPV surveillance capsule data that is reported through implementation of a plants 10 CFR Part 50, Appendix H, RPV material surveillance program.

The NRC staffs recommended guidelines for calculating the effects of neutron radiation on the USE values for the RPV beltline materials are provided in NRC RG 1.99, Revision 2.

Section IV.A.2 of 10 CFR Part 50, Appendix G, requires that the P-T limits for operating reactors be at least as conservative as limits obtained by following the methods of analysis and the margins of safety of the ASME Code,Section XI, Appendix G. The rule also requires that the P-T limits calculations account for the effects of neutron radiation on the material properties of the RPV beltline materials and that P-T limits calculations incorporate any applicable RPV surveillance capsule data that is reported as part of the licensees implementation of its 10 CFR Part 50, Appendix H, RPV materials surveillance program. The ASME Code,Section XI, Appendix G, specifies a procedure for calculating P-T limits that is based on the principles of linear elastic fracture mechanics. The key material property input to this linear elastic fracture mechanics procedure is the critical stress intensity factor, KIC, also referred to as the fracture toughness. The KIC is a function of the difference in metal temperature and the Reference nil-ductility temperature (RTNDT) for the material. Therefore, for a given RTNDT value, KIC is a single-valued function of metal temperature, referred to as the KIC curve. Neutron irradiation of RPV beltline materials will increase their RTNDT values, thereby causing a shift to the KIC versus temperature curve, which directly corresponds to a conservative shift in the RPV beltline P-T limit curve. The NRC staffs recommended guidelines for calculating the effects of neutron radiation on the RTNDT for RPV beltline materials, whereby licensees calculate the adjusted RTNDT (ART) value due to neutron radiation, are specified in RG 1.99, Revision 2. Finally,

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION guidelines published in RIS 2014-11, provide additional NRC staff expectations for evaluations of P-T limits in licensing applications, including guidance on the definition of the beltline region and consideration of structural discontinuities in the development of P-T limits. As stated in the RIS, structural discontinuities could potentially result in more bounding P-T limits than those defined by the RPV beltline shell material with the most limiting fracture toughness. Therefore, structural discontinuities need to be considered in the development of P-T limit curves.

In Section 3.2.1 of the TSAR, the licensee identified that the projected ART values for the RPV beltline materials have been evaluated for the proposed MUR, and new P-T limit curves were developed using the NRC-approved BWROG methodology stated below. A listing of the 56 EFPY ART values (pertaining to 60-years of operation) and all input parameters, which include material property inputs and projected neutron fluence, are provided in Table 3-4 of the TSAR.

By letter dated December 14, 2017 (ADAMS Accession No. ML17324A840), the NRC staff issued Amendment No. 209 and its accompanying SE for HCGS regarding the relocation of the P-T limits from the TSs to a new licensee-controlled document called the P-T Limits Report (PTLR), consistent with the guidance in GL 96-03, Relocation of the Pressure Temperature Limit Curves and Low Temperature Overpressure Protections System Limits. The amendment also added new administrative controls via TS 6.9.1.10, Reactor Coolant System (RCS)

Pressure and Temperature Limits Report (PTLR), which provides requirements for the control of future changes to the plant-specific P-T limits and for submittal of PTLR revisions to the NRC.

Under TS 6.9.1.10, it is required that the analytical methods used to determine the P-T limits shall be those previously approved by the NRC as described in BWROG-TP-11-022-A, Revision 1, Pressure-Temperature Limits Report Methodology for Boiling Water Reactors, dated August 2013 (ADAMS Accession No. ML13277A557. The NRC staffs SE regarding these TS changes documents the technical bases for acceptance of the licensees implementation of the BWROG-TP-11-022-A, Revision 1, PTLR methodology for generating current and future plant-specific P-T limits in accordance with the TS 6.9.1.10 administrative controls. Any change to the analytical methods to determine the P-T limits that deviates from the BWROG PTLR methodology, as cited in TS 6.9.1.10, would require prior NRC approval by license amendment pursuant to 10 CFR 50.90.

As documented in its SE for the HCGS PTLR amendment, the NRC staff verified that the licensees P-T limit calculations based on the BWROG-TP-11-022-A, Revision 1, PTLR methodology were adequately addressed and were bounding for all ferritic RPV beltline and non-beltline components. For its evaluation of the specific effects of the MUR on the HCGS P-T limits, the NRC staff reviewed and independently verified the licensees RPV beltline material ART calculations for MUR conditions, as provided in Table 3-4 of the TSAR. The NRC staff confirmed that the 56 ART values for the RPV beltline plates, welds, and nozzles were correctly recalculated for MUR conditions in accordance with RG 1.99, Revision 2, taking into consideration the slight increase in the projected neutron fluence that would result from the MUR. The NRC staff verified that all RPV beltline material property inputs to the ART calculations, as listed in Table 3-4 of the TSAR, are consistent with the CLB. The staff also determined that the neutron fluence inputs to the ART calculations are acceptable for the proposed MUR conditions based on the fact that the licensee used NRC-approved neutron fluence calculational methods that are consistent with RG 1.190.

The NRC staff identified that TS 6.9.1.10 provides the necessary administrative controls for updating the PTLR to account for new neutron fluence-based operating periods and for

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION submitting these PTLR revisions to the NRC. This includes, among other things, any necessary revisions to P-T limits based on new neutron fluence periods that are calculated for power uprate amendment applications. The staff also verified that the projected ARTs provided in Table 3-4 would ensure that the P-T limits remain in compliance with 10 CFR Part 50, Appendix G for MUR conditions Therefore, the staff finds that the PTLR controls in TS 6.9.1.10 are acceptable for ensuring that changes to the neutron fluence, ARTs and P-T limits - or EFPY operating term for these limits - will be implemented following the staffs approval of the MUR.

Therefore, the staff determined that the licensees evaluation of its P-T limits and RPV beltline region ART values for MUR conditions is acceptable for satisfying the requirements of 10 CFR Part 50, Appendix G.

Upper Shelf Energy Evaluation The licensee stated that beltline material upper shelf energy (USE) value calculations were updated by incorporating the results of updated fluence values on end-of-license USE values.

The licensee also stated that the fluence projections used in the present calculation assumed a 1.6 percent power uprate for all future fuel cycles.

The licensee determined the RPV neutron fluence in accordance with the NRC-approved radiation analysis modeling application (RAMA) fluence methodology, which is documented in LTR BWRVIP-114NP-A, BWR Vessel and Internals Project - RAMA Fluence Methodology Theory Manual (ADAMS Accession No. ML092650376). By letter dated December 14, 2017, the NRC staff determined that the licensee had acceptably implemented the RAMA fluence methodology insofar as it would support the calculation of RPV P-T limits. The NRC staffs SE documented the basis for the NRC staffs determination that the fluence calculations adhere to RG 1.190 guidance and are therefore acceptable.

Because the licensee calculates neutron fluence in accordance with this generic methodology, the most prevalent effects of the proposed MUR are changes to the characteristics of the core neutron source and moderator density. The licensee stated the present calculations assume a 1.6 percent power uprate for all future cycles, therefore, the NRC staff concluded that the fluence calculations properly account for the TPO effects. Thus, the fluence calculations remain adherent to RG 1.190, and the NRC staff concludes that the licensee has acceptably accounted for the effect of the proposed MUR on RPV neutron fluence.

In TSAR Section 3.2.1, the licensee stated that the USE will remain greater than 50 ft-lb for the design life of the RPV. The licensee identified that all of the RPV beltline materials at HCGS have unirradiated USE data, yet an EMA was still performed to determine how the surveillance data effects USE reductions for the limiting beltline plate and weld materials. The USE values and EMA calculations are documented in Tables 3-1, 3-2 and 3-3 of the TSAR. The tables show calculations of projected USE and EMA results through 56 EFPY for MUR conditions, based on RG 1.99, Revision 2.

RG 1.99, Revision 2, recommends that projected USE values for RPV beltline materials be calculated based on the projected neutron fluence at a postulated flaw depth corresponding to one-quarter of the RPV beltline wall thickness from the clad/base metal interface of the RPV (1/4T location), the weight percentage (wt. %) of copper (Cu) in the material, and the preservice (unirradiated) USE value for the material. The 1/4T projected neutron fluence and the wt. % Cu are used to calculate the projected percentage decrease in the USE for the material using Figure 2 of RG 1.99; and when available, credible surveillance data are applied to this

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION calculation in accordance with Position 2.2 of the RG. If valid unirradiated USE data is available, the projected percentage decrease in USE directly determines the projected USE at the end of the licensed operating term. If projected USE values do not meet the 50 ft-lbs acceptance criterion from 10 CFR Part 50, Appendix G, or if plant-specific unirradiated USE data for the material is not available to demonstrate projected USE greater than 50 ft-lbs, an EMA shall be performed for the licensed operating term.

Certified RPV material test reports for many older BWRs do not include adequate Charpy USE data for the unirradiated condition. Therefore, some BWR plants are unable to demonstrate that the USE for their RPV beltline materials will be maintained greater than 50 ft-lbs throughout the licensed operating terms, and they have been required to perform EMAs for 40-year license terms and for 20-year LR terms to demonstrate compliance with 10 CFR Part 50, Appendix G.

TR NEDO-32205-A, Revision 1 (Legacy Library No. 9403280161), documents the original (40-year) generic EMA methodology and EMA results that were approved by the NRC-staff for generic application to all U.S. BWRs for demonstrating compliance with 10 CFR Part 50, Appendix G. These generic EMA results for BWRs were later updated for LR via the BWRVIP-74-A report (ADAMS Accession No. ML031710354), which was reviewed and approved by the NRC staff on October 18, 2001 (ADAMS Accession No. ML012920549), for referencing in LR applications. The BWRVIP-74-A report identifies that LR applicants without sufficient unirradiated USE data must demonstrate that their RPV beltline materials satisfy the applicable generic EMA acceptance criteria for 60-year license terms. These BWR LR applicants addressed the BWRVIP-74-A criteria by (1) performing plant-specific RG 1.99, Revision 2, calculations of the projected percentage decrease in the USE for the limiting RPV beltline plate and weld materials for 20-year LR terms; and (2) then comparing the projected percentage decrease in USE to the applicable EMA acceptance criterion for the BWR plant categories and plate and weld material types specified in BWRVIP-74-A. However, some newer BWR plants like HCGS do have adequate unirradiated Charpy USE data in their certified material test reports, which can be used to directly calculate the projected end-of-license USE based on their projected percentage decrease in USE, per RG 1.99, Revision 2.

For all power uprates, licensees must reevaluate the projected percentage decrease in the USE for the RPV beltline materials based on the wt. % Cu content for the material and the projected neutron fluence at the 1/4T location. In Table 3-1 of the TSAR, the licensee provided the unirradiated USE, wt. % Cu content, 1/4T neutron fluence, projected percent decrease in USE, and projected USE at the 1/4T location for all RPV beltline plates, welds and nozzles at HCGS for 60 years of operation (56 EFPY) at MUR conditions. The NRC staff reviewed these calculations and verified that all of the RPV beltline materials are projected to maintain greater than 50 ft-lbs USE for MUR conditions, based on the unirradiated USE values listed in TSAR Table 3-1. The NRC staff identified that no EMA is needed for the non-surveillance RPV beltline materials. For the surveillance plate and weld materials represented in the ISP, the licensee applied the BWRVIP-74-A EMA results using ISP surveillance data from their previously pulled 30 degree and 120 degree capsules, representing the limiting plate (Heat No. 5K3238/1) and the limiting weld (Heat No. D53040), and documented the results in TSAR Tables 3-2 and 3-3 respectively. The staff independently verified that the licensee correctly applied this surveillance data for calculating the projected percent decrease in USE in accordance with Position 2.2 of RG 1.99, Revision 2. Additionally, the staff determined that the licensee used the correct EMA acceptance criteria from BWRVIP-74-A for demonstrating that the projected percent decrease in USE is acceptable. The NRC staff confirmed that the wt. % Cu values used for all USE and EMA calculations are consistent with CLB values that were previously approved in HCGS

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION License Amendment No. 209 to implement the PTLR. Unirradiated USE values reported in the TSAR Table 3-1 are consistent with CLB values that were approved for LR and in the FSAR.

The staff also determined that the neutron fluence inputs for all calculations of projected percent decrease in USE are acceptable for the proposed MUR conditions based on the fact that the licensee used NRC-approved neutron fluence calculational methods that are consistent with RG 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence" (Reference 9). Therefore, the staff determined that the licensees USE evaluation and EMA results for the RPV beltline materials are acceptable for satisfying the requirements of 10 CFR Part 50, Appendix G, for MUR conditions.

Reactor Pressure Vessel Circumferential and Axial Welds Inservice inspection (ISI) of ASME Code Class 1, 2, and 3 components is performed in accordance with Section XI of the ASME Code and applicable Addenda, as required by 10 CFR 50.55a(g). The ASME Code,Section XI, Table IWB-2500-1, requires ISI of all RPV welds every 10-year ISI interval. Pursuant to 10 CFR 50.55a(z), plant-specific alternatives to the Code may be used when authorized by the NRC if (1) the proposed alternatives would provide an acceptable level of quality and safety, or (2) compliance with the specified requirements would result in hardship or unusual difficulty, without a compensating increase in the level of quality and safety.

In Section 3.2.1 of the TSAR, the licensee stated that the limiting 56 EFPY beltline circumferential and axial weld mean RTNDT values will remain bounded by the criteria in NRC-approved Topical Report BWRVIP-05 (Legacy Library No. 9808040037). The staff notes that this analysis also addresses BWRVIP-74-A acceptance criteria for periods of extended operation, as stated below. The licensee provided the results of the limiting RPV beltline axial weld and circumferential weld evaluations in Tables 3-5 and 3-6 of the TSAR, respectively.

Code Alternatives allowing for permanent elimination of RPV circumferential shell weld exams were authorized pursuant to 10 CFR 50.55a(z)(1) for the duration of all BWR plants original 40-year licensed operating terms, based on the licensees application of the NRC-approved BWRVIP probabilistic fracture mechanics (PFM) methodologies that are described in the staffs SE for BWRVIP-05. The BWR licensees must re-apply for these ASME Code alternatives upon entering the period of extended operation to obtain NRC authorization for implementation of the staff-approved BWRVIP PFM methods to justify elimination of RPV circumferential shell weld exams for 60 year extended license terms. The NRC-approved BWRVIP-74-A report provides the technical basis for application of the PFM methods for periods of extended operation. The NRC staffs SE accompanying the BWRVIP-74-A report provides the NRC staffs specific RPV weld embrittlement acceptance criteria for plant-specific application of these PFM results for elimination of BWR RPV circumferential shell examinations for 20-year LR terms. These acceptance criteria must be satisfied for both RPV circumferential and axial welds1. Licensees must demonstrate in their 10 CFR 50.55a(z)(1) submittals that the projected embrittlement of 1 The NRC staffs 1998 SE for BWRVIP-05 identified a need to further evaluate the higher conditional failure probability levels for RPV axial welds. Therefore, the staff performed a review of supplemental BWRVIP correspondence regarding the axial weld failure probabilities. The staffs March 7, 2000, supplemental SE for BWRVIP-05 (ADAMS Accession No. ML003690281) concluded that the RPV failure frequency due to failure of the limiting axial welds in the BWR fleet are acceptable on a generic basis; however, these generic axial weld results are only applicable to 40-year license terms. Therefore, consideration of BWR axial welds for renewed license terms would require a plant-specific treatment, based on plant-specific evaluation of limiting axial weld embrittlement for periods of extended operation.

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION their limiting RPV circumferential and axial welds are bounded by the acceptance criteria documented in the BWRVIP-74-A SE for periods of extended operation. As documented in the BWRVIP-74-A SE, the projected embrittlement is determined based on the mean RTNDT values for the limiting circumferential and axial welds, which is equal to the unirradiated RTNDT value, plus the projected mean value of the shift in RTNDT caused by neutron embrittlement during the period of extended operation. The mean value of the shift in the RTNDT is calculated at the interface of the RPV clad and weld metal in accordance with the procedures of RG 1.99, Revision 2.

The HCGS site is still in the 40-year license term; and will therefore, require staff authorization of this Code alternative for a 60-year extended license when it enters the period of extended operation in 2026. However, since HCGS holds a renewed operating license, the licensee had performed LR time-limited aging analyses in accordance with 10 CFR 54.21(c), for demonstrating that the projected embrittlement of the limiting RPV circumferential and axial welds would remain bounded by the BWRVIP-74-A acceptance criteria for the 20-year period of extended operation. Thus, the licensee correctly determined that these analyses require update for MUR conditions. The staff verified that Tables 3-5 and 3-6 of the TSAR provide the necessary updates for the limiting RPV circumferential and axial weld mean RTNDT values for MUR conditions. The NRC staff reviewed the licensees mean RTNDT calculations for MUR conditions and verified that they are correct and represent the most limiting RPV circumferential and axial weld materials at HCGS. The staff reviewed the initial RTNDT and CF values used as inputs for the mean RTNDT calculations and verified that they are consistent with those previously approved by the staff in HCGS License Amendment No. 209. The staff also determined that the neutron fluence inputs to the mean RTNDT calculations are acceptable for the proposed MUR conditions based on the licensees use of NRC-approved neutron fluence calculational methods that are consistent with RG 1.190. Finally the staff verified that the licensee applied the correct NRC acceptance criteria for the circumferential and axial weld mean RTNDT values from BWRVIP-74-A for determining that the welds will continue to satisfy the applicable RPV PFM results supporting the elimination of the RPV circumferential shell weld exams for the period of extended operation under MUR conditions. Therefore, the staff determined that the licensees evaluation of its limiting RPV circumferential and axial welds for MUR conditions is acceptable for ensuring that its analytical basis for permanent elimination of RPV circumferential shell weld exams will be valid for the period of extended operation.

However, the staff notes that, since HCGS is still operating in the 40-year license term, its CLB Code alternative for elimination of the RPV circumferential welds exams will expire upon entering the period of extended operation; and it will have to reapply for this Code alternative for the 20-year extended license term in order to obtain NRC-authorization for permanent elimination of the RPV circumferential weld exams for 60 years.

Reactor Pressure Vessel Structural Evaluation In TSAR Section 3.2.2, the licensee provided a structural evaluation of the RPV. This is a bounding reconciliation for stresses for normal, upset, emergency, and faulted conditions, as well as fatigue for normal and upset conditions. To address concerns pertaining to GEH safety communication (SC) letters GEH SC 12-20 and SC 13-08, the licensee showed that the shroud support attachment to the RPV component is within the allowable limits for acoustic loads and is therefore structurally qualified for operation at TPO uprate conditions.

The NRC staff reviewed the licensees submittal. The licensees reconciliation is a bounding reconciliation for stress and fatigue of the reactor vessel because the evaluations performed by

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION the licensee in 2008 for the EPU power level of 3840 MWt ((

)) for a 60-year plant life. The NRC staff noted that there are no RPV components that ((

)) Based on a review of the licensees evaluation, the staff concludes that the reactor vessel will maintain its structural integrity allowable limits per the applicable ASME Code 1968 Edition, including Winter 1969 Addenda, from stress and fatigue considerations for normal and upset conditions, as well as from stress considerations for emergency and faulted conditions at the proposed TPO uprate.

Reactor Vessel Conclusion Based on the considerations stated above, the NRC staff concludes that there is reasonable assurance that the structural integrity of the RPV will continue to be maintained under TPO uprate conditions, consistent with the regulatory requirements set forth in 10 CFR Part 50, Appendix G; 10 CFR Part 50, Appendix H; 10 CFR 50.60; and 10 CFR 50.55a.

3.3.3 TSAR Section 3.3 - Reactor Internals In TSAR Section 3.3, the licensee described the reactor internals, including the (1) core support structure components (shroud, shroud support, core plate, top guide, control rod drive housing, control rod guide tube, and orificed fuel support); and (2) non-core support structure components (FW sparger, jet pump assembly, core spray (CS) line and sparger, access hole cover, steam dryer, shroud head and steam separator assembly, core differential pressure and liquid control lines, fuel channel and low pressure coolant injection (LPCI) coupling). The evaluation for reactor internal pressure differences (RIPDs) was addressed in TSAR Section 3.3.1 and the structural evaluation of reactor internals in TSAR Section 3.3.2. The structural evaluation or reconciliation of the steam dryer for TPO conditions was provided in a licensee supplement dated December 19, 2017.

The NRC staff reviewed the licensees submittal and supplements. The reactor internals are not ASME Code components, but the licensee utilized ASME Code requirements as a guide in the structural evaluations of the internals. The RIPDs for the TPO uprate for normal and upset operating conditions are slightly higher from the CLTP. However, the reconciliation evaluations demonstrated that all reactor internals are within the allowable limits. The limiting stresses and fatigue usage factors for all RPV internals are shown to be acceptable. The emergency and faulted evaluations of RIPDs for the TPO uprate were conservatively bounded by the EPU analysis performed at 102 percent of EPU, and therefore are acceptable.

The licensee also addressed annulus pressurization and jet reaction acoustic-and flow-induced loads pertaining to GEH SC letters GEH SC 12-20 SC 14-02 and SC 14-03 in the TSAR, Section 3.3.2. The stresses of the RPV internals that were affected by the SCs were reconciled for the increase of the acoustic load to show that adequate stress margins still exist and the stresses remain within the allowable limits. All the RPV internals were shown to be within the allowable limits. Therefore, the RPV internal components are demonstrated to be structurally qualified for operation at TPO conditions.

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION

))

The NRC staff noted that MASR is greater than 1.0 demonstrating that the steam dryer components meet the ASME Code acceptance criteria for high cycle fatigue. In addition to the high cycle fatigue assessment, the steam dryer was also evaluated by the licensee to show compliance with the structural requirements of the ASME Boiler and Pressure Vessel Code (ASME Code),Section III, Subsection NG. The assessment shows that the dryers meet the stress and fatigue usage limits of the ASME Code for the duty cycles due to implementation of MUR for normal operation, upset conditions, emergency conditions, and faulted conditions.

Based on the review as outlined above, the staff concludes that the steam dryers meet the applicable allowable limits and therefore, maintain their structural integrity for the proposed MUR conditions.

3.3.4 TSAR Section 3.4 - Flow-Induced Vibration The licensee performed flow-induced vibration (FIV) evaluations for reactor internals in TSAR Section 3.4. This discussion also included evaluations for piping components, thermowells in main steam (MS), feedwater (FW), and reactor recirculation system (RRS) piping.

The NRC staff reviewed the licensees evaluation that determined the effects of FIV on the reactor internals at 105 percent of rated core flow and at a power level of 3,905 MWt (101.7 percent of CLTP). The vibration levels for the TPO conditions were estimated from measured vibration data during startup tests on HCGS, and the BWR prototype plant. The licensee estimated expected vibration levels for TPO by extrapolating the measured vibration data based on GEH BWR operating experience. These expected vibration levels were compared with vibration acceptance limits.

For the proposed rated TPO thermal power, there is an increase in FW flow of approximately 2 percent, that leads to a less than 4 increase in FIV stresses in the FW sparger, jet pumps, shroud, shroud head and separator. By extrapolation of measured data, the licensee showed that the resulting stresses are within acceptable limits.

Since there is no change in core flow from CLTP to the TPO uprate power, there is no change in stresses for the control rod guide tube and the in-core guide tube.

For CS piping and sparger, based on ASME, Appendix N criteria, the licensee demonstrated that there is adequate separation between the structural natural frequencies and the vortex shedding frequency; therefore, no resonance or change in stress occurs in the TPO region. For jet pump sensing lines, the licensee also showed that no resonance occurs at the vane passing frequency range of the recirculation pumps due to TPO. The HCGS fuel assembly is acceptable for TPO conditions because bounding values were used for operating conditions.

By conservatively assuming a lock-in condition, the maximum stresses in the guide rods were calculated and shown to be within the GEH conservative acceptance limit of 10,000 psi. It is noted that the GEH acceptance limit of 10,000 psi is more conservative compared with the ASME Code high cycle fatigue stress limit of 13,600 psi corresponding to 1 x 1011 cycles for stainless steel.

The flow rates in safety-related MS and FW piping increase by less than 2 percent due to the TPO uprate. The RRS flow rate is essentially unchanged at TPO. The safety-related portions of the MS and FW piping thermowells experience increased vibration levels, approximately proportional to the increase in the square of the flow velocities and in proportion to any change

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION in fluid density. As the result of a roughly 2 degrees Fahrenheit (°F) increase in FW temperature, there is an insignificant decrease in fluid density for TPO conditions. Analytical evaluations have shown that the safety related piping components, and thermowells, in the MS, FW, and RRS piping are structurally adequate for TPO conditions.

Based on a review of the licensees evaluations as summarized above, the staff concludes that the reactor internals, MS, FW, & RRS thermowells and piping components will maintain their structural integrity from FIV considerations for the proposed MUR power uprate.

3.3.5 TSAR Section 3.5 - Piping Evaluation The licensees evaluations for the RCPB piping and its supports are included in TSAR Section 3.5.1 and for the balance of plant (BOP) piping and its supports in TSAR Section 3.5.2.

The licensee described that there is no change in methods used for plant specific piping and pipe support evaluations from those used in the 2008 EPU evaluations for HCGS. The effect of the TPO uprate without a nominal vessel dome pressure increase is negligible for the RCPB portion of all piping except for portions of the FW lines, MS lines, and piping connected to the FW and MS lines. The steam flow in the MS lines at TPO RTP is approximately 2 percent higher than at CLTP. There is an increase of roughly 2 percent in FW flow and pressure, and a 2 degree F increase FW temperature.

The NRC staff reviewed the licensees submittal. The current analysis bounds the TPO conditions for recirculation piping due to the negligible nature of the changes. There is a small change in core pressure drop and recirculation fluid temperature, negligible changes in piping stresses, and negligible impacts on pipe supports. For the RPV bottom head line, reactor core isolation cooling (RCIC) piping, HPCI piping, LPCI piping, CS piping, standby liquid control (SLC) system piping, and reactor water cleanup piping, there is negligible change in piping stresses and pipe support loads due to the small change in core pressure and recirculation fluid temperature.

For the MS and FW lines, supports, and connected lines, factors were applied by the licensee to determine the percentage increases in applicable ASME Code stresses, displacements, cumulative usage factors (CUF), and pipe interface component loads (including supports, anchors, equipment nozzle loads) as a function of the percentage increase in pressure (where applicable), temperature, and flow due to TPO conditions. The licensee applied the percentage increases to the highest calculated stresses, displacements, and CUF at applicable piping system node points to conservatively determine the maximum TPO calculated stresses, displacements, and usage factors. This approach is conservative because the TPO does not affect certain loads such as weight and all building filtered loads (i.e., seismic loads are not affected by the TPO). The factor approach used by the licensee is conservative as it increases contributions from certain loads such as dead weight, and seismic loads that are unaffected by TPO. The NRC staff determines, therefore, that MS, FW, and attached piping and pipe supports meet acceptance criteria, and the CLB evaluation conclusions remain valid for TPO conditions.

The NRC staff concludes that the licensee adequately addressed the effect of minor changes in MS flow, FW flow, FW pressure, and FW temperature due to the TPO RTP on piping stresses and pipe supports, and showed the acceptability using a factored approach for the TPO conditions. For other RCPB, as well as BOP piping and supports, there is a negligible impact and the CLB analysis bounds the TPO conditions. Therefore, the NRC staff concludes that the

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION RCPB piping and supports, as well as BOP piping, will remain acceptable for the proposed TPO uprate.

Flow-Accelerated Corrosion Flow-accelerated corrosion (FAC) is a corrosion mechanism that occurs in carbon steel components exposed to either single-phase or two-phase water flow. Components made from stainless steel are not affected by FAC, and FAC is significantly reduced in components containing a small amount 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 FAC program is based on GL 89-08, Erosion/Corrosion-Induced Pipe Wall Thinning (ADAMS Accession No. ML031200731) and the acceptance criteria is based on a structural evaluation of the minimum acceptable wall thickness for the components undergoing degradation by FAC.

In TSAR Section 3.5.1, the licensee stated that there is a program to monitor FAC for pipe wall thinning in both single and two-phase high-energy carbon steel piping in the MS, FW, and BOP systems. The licensee stated that the HCGS FAC program implements the recommendations of GL 89-08 and a predictive method is used to calculate potential wall thinning of components susceptible to FAC. In the LAR, the licensee stated that operation at the proposed MUR power uprate conditions results in changes to parameters (e.g. flow velocity, temperature, moisture content) in certain systems. The licensee stated its evaluation of predicted wall thinning indicates a minimal effect for the MS, FW and attached systems, but that the FAC monitoring program will continue to consider adjustments related to predicted material loss. The licensee stated that high-energy piping systems will continue to be monitored in order to provide confidence in the integrity of these systems. In addition, the licensee stated that no changes to inspection frequencies are required to maintain adequate margin for certain piping systems.

However, for piping systems impacted by the changing process conditions, the piping inspection frequency will be changed appropriately. In addition, the licensee stated that the continued use of existing procedures will provide reasonable assurance that any effects from the MUR power uprate on FAC-induced wall thinning will be monitored and addressed.

For the MS, FW and associated systems, the licensee stated that no changes to the piping inspection scope or frequency would be required due to the MUR power uprate because the current program is adequate to ensure that potential adverse effects from the MUR power uprate will be monitored and addressed. Based on the licensees current program to monitor FAC, which incorporated the recommendations from GL 89-08, EPRI NSAC-202L Recommendations for an Effective Flow-Accelerated Corrosion Program (ADAMS Accession No. ML030100368), and a predictive method to calculate wall thinning of components susceptible to FAC, the NRC staff has reasonable assurance that potential adverse effects from FAC due to the MUR power uprate will continue to be appropriately monitored and managed.

The licensee stated that predicted material loss rates may impact the piping inspection frequency for the BOP piping systems. The NRC staff reviewed the predicted material loss rates for the BOP piping systems provided by the licensee. The material loss rates for the BOP piping systems appear to be consistent with the expected material loss rates for an MUR power uprate. In conjunction with the licensees monitoring program, the staff has reasonable assurance that the potential effects from FAC will be appropriately managed with respect to the

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION MUR power uprate.

The NRC staff reviewed the licensees 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 provide reasonable assurance the loss of material by FAC will be predicted, and will ensure timely repair or replacement of affected components following implementation of the proposed MUR power uprate. The NRC staff has found that the FAC program will provide reasonable assurance that components susceptible to FAC will be managed appropriately post MUR power uprate implementation.

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

3.3.6 TSAR Section 3.6 - Reactor Recirculation System As described in UFSAR Section 5.4.1.3, the reactor recirculation system consists of two reactor recirculation loops external to the reactor vessel. The pump and piping in each loop provide the path and driving flow of water to the reactor vessel jet pumps. Each external loop contains one high capacity variable speed motor driven recirculation pump and two motor operated gate valves for pump maintenance. Each loop also contains a flow measuring system. These recirculation loops are part of the RCPB and are located inside the drywell structure. The jet pumps, however, are part of the reactor vessel internals. The primary function of the reactor recirculation system is to vary the core flow and power during normal operation.

The licensee evaluation in TSAR Section 3.6 states that the TPO uprate does not require an increase in the maximum core flow capability and there is no significant reduction in the maximum flow capability because the core pressure drop increases by less than one psi. The effect on recirculation pump net positive suction head (NPSH) at TPO uprate conditions is negligible. The licensee also confirmed that no significant increase in the reactor recirculation system vibration occurs at the TPO operating conditions. The cavitation protection interlock for the recirculation pumps and jet pumps is not affected by the TPO uprate.

Based on the consideration stated above, the NRC staff concludes that there is reasonable assurance that the changes associated with the TPO uprate will not impact the capability of the reactor recirculation system form performing its intended functions.

3.3.7 TSAR Section 3.7 - Main Steam Line Flow Restrictors As described in UFSAR Section 5.4.4.2, a flow restrictor assembly is provided for each of the four MSLs in the NSSS. The flow restrictor is located in the drywell and is a venturi type nozzle welded into the MSL. The flow restrictor has no moving parts; however, in the event of a MSL break outside the containment, the restrictor limits the coolant blowdown rate from the reactor vessel. Its mechanical structure can withstand the velocities and forces associated with a MSL break.

As stated in the TSAR Section 3.7, the licensee applied the generic evaluation provided in the TLTR Appendix J.2.3.7. The licensee stated that there are no changes to the main steam line flow restrictors requirements. Since there is no change in operating pressure, there would be no change in steam flow rate if a break occurred.

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION Based on the above, the NRC staff concludes that the TLTR Appendix J.2.3.7 is applicable because there is no change in the MSL break flow rate. Therefore, the staff finds the MSL flow restrictors are acceptable in the TPO uprate conditions.

3.3.8 TSAR Section 3.8 - Main Steam Isolation Valves As described in UFSAR Section 5.4.5.2, the MSL isolation valves (MSIVs) are a part of the main steam isolation system. Two MSIVs are welded in the horizontal run of each of the four main steam lines with one valve as close as possible to the inside of the drywell, and the other as close as possible to the outside of the primary containment. In the event of a MSL break inside or outside the containment, the MSIVs are closed and form a flow barrier.

As stated in the TSAR Section 3.8, the licensee applied the generic evaluation from the TLTR Appendix J.2.3.7. The licensee stated that the all safety and operational aspects of the MSIVs are within the previous evaluations for the EPU uprate.

Based on the above, the NRC staff concludes that the TLTR Appendix J.2.3.7 is applicable because there is no change in the requirements for the MSIVs under the TPO uprate conditions.

Therefore, the staff finds the MSIVs are acceptable in the TPO uprate conditions.

3.3.9 TSAR Section 3.9 - Reactor Core Isolation Cooling As described in UFSAR Section 5.4.6, the RCIC is a safety-related system consisting of a steam turbine, turbine driven pump, piping, valves, controls, and instrumentation designed to ensure that sufficient reactor water inventory is maintained in the reactor vessel to allow for adequate core cooling. The system prevents reactor fuel from overheating during the conditions (1) when the vessel is isolated and maintained in the hot standby condition, (2) when the vessel is isolated with a loss of coolant flow from the reactor FW system, and (3) when the plant is shutdown with a loss of the normal FW system and before the reactor is depressurized to the level for the operation of the shutdown cooling (SDC) system.

As stated in the TSAR Section 3.9, the licensee applied the generic evaluation from the TLTR Section 5.6.7. The licensee stated that the TPO uprate does not affect the RCIC system operation, initiation, or capability requirements.

The NRC staff determined that the generic evaluation is applicable because the ability of the RCIC to perform required design functions has been demonstrated with previous analyses based on 102 percent of CLTP that bounds the TPO uprate conditions. Therefore, based on the above, the RCIC system continues to meet the requirements of GDC 33 because it will perform its safety function under the TPO uprate conditions so that the fuel design limits are not exceeded under the conditions (a), (b), and (c), described above.

3.3.10 TSAR Section 3.10 - Residual Heat Removal System As described in UFSAR Section 5.4.7.1, the residual heat removal (RHR) system consists of four independent loops A, B, C and D. Each loop contains a motor driven pump, piping, valves, instrumentation, and controls. Each loop takes suction from the suppression pool and is capable of discharging water to the reactor vessel via separate low pressure coolant injection (LPCI) nozzles, or back to the suppression pool via a full flow test line. Loops A and B have heat exchangers that are cooled by an independent loop of the safety auxiliaries cooling system

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION (SACS). In addition, the loops A and C pump discharge headers and the loops B and D pump discharge headers are each cross-tied via two manual isolation valves. The purpose of these cross-ties is to permit the use of C pump with RHR heat exchanger A and the use of D pump with RHR heat exchanger B for alternate decay heat removal. The two RHR heat exchanger loops can also take suction from the reactor recirculation system suction or the fuel pool and can discharge into the reactor recirculation pump discharge, fuel pool cooling discharge, or to the suppression pool and drywell spray spargers. The system has five subsystems or modes of operation; each having its own functions. These modes are as follows: (1) SDC, (2) LPCI, (3) suppression pool cooling (SPC), (d) containment spray cooling (CSC), and (e) alternate shutdown cooling (ASDC).

The licensee stated that the SDC and the ASDC modes of the RHR system are not affected by the TPO uprate because a small increase in the decay heat is within the heat removal capability of the RHR equipment. By letter dated December 22, 2017, the licensee confirmed that the increase in the decay heat for the TPO uprate had been quantitatively verified to be within the RHR equipment heat removal capability in the SDC and the ASDC modes of the RHR system.

In its response, the licensee stated that assuming a decay heat associated with 102 percent of 3,840 MWt (or 3,917 MWt) which bounds the TPO power level of 3,902 MWt, the SDC analysis demonstrated that the reactor coolant temperature could be reduced to 200°F within 13.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> of the plant shutdown using one heat exchanger in the SDC; and therefore, meets the TS 24-hour requirement with sufficient margin. The analysis also demonstrated that if the reactor cooldown is conducted using the ASDC mode with one RHR heat exchanger in operation, cold shutdown can be achieved in approximately 60 hours6.944444e-4 days <br />0.0167 hours <br />9.920635e-5 weeks <br />2.283e-5 months <br />, and therefore, meets the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> cold shutdown requirement for the Appendix R Fire event.

The SPC mode of the RHR system maintains the suppression pool temperature below the TS limit during normal operation and is not affected for the TPO uprate because the reasons for pool heat-up, such as leakby of the safety relief valve (SRV), are not power dependent. For LOCA and special events, the SPC and LPCI modes of the RHR system are evaluated in SE Section 3.4.2.

The CSC mode of the RHR system sprays water into the containment during a large or small break LOCA to reduce the post-accident containment pressure and temperature. This mode is used during accidents and is not used during special events. In the analysis of record (the most recent analysis), this mode was evaluated at 102 percent CLTP which bounds the TPO uprate power. Therefore, the NRC staff determines that the CSC mode of the RHR system is not affected by the TPO uprate.

Based on the evaluation of the RHR system under the TPO uprate conditions, the system in its SPC mode continues to meet the requirements of GDC 38 because it will maintain the suppression pool temperature below its TS limit during normal operation. The system in its CSC mode continues to meet the requirements of GDC 38 because it will reduce the containment pressure and temperature following LOCAs and maintain them at less than the design conditions. The system in its SDC and ASDC modes continues to meet the requirements of GDC 38 because it will bring the reactor to cold shutdown condition as required in the TS.

3.3.11 TSAR Section 3.11 - Reactor Water Cleanup System In TSAR Section 3.11 the licensee stated it applied the generic evaluation of the reactor water

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION cleanup (RWCU) system provided in the TLTR Section 5.6.6 and Appendix J.2.3.4. The licensee stated that there would not be a significant change in the operating pressure and operating temperature in the high pressure portion of the system for the TPO uprate. Further, the licensee stated that RWCU flow will not change.

The NRC staff determines that the TLTR Section 5.6.6 and Appendix J.2.3.4 are applicable because there is not a significant change in the operating pressures for RWCU system under the TPO uprate conditions. Therefore, the staff finds the RWCU system acceptable for the TPO uprate conditions for the system requirements.

In addition to the system requirements, the licensee stated it met the requirements of the TLTR to confirm that safety and operational aspects of water chemistry performance are not affected. The NRC staff reviewed the impacts from the proposed MUR power uprate on the ability of the RWCU system to maintain water purity and the operating parameters of the system. The NRC staff verified the RWCU system operation and performance with respect to water chemistry.

Because the impact of the proposed MUR power uprate is expected to be minimal, the staff determines that the RWCU system will continue to maintain RCS inventory and water chemistry.

Thus, the NRC staff concludes that the licensee adequately addressed the impacts on the MUR power uprate on the RWCU system, with respect to water chemistry performance.

3.4 TSAR Section 4.0 - Engineered Safety Features The following provides the NRC staffs technical review of the topics in Section 4.0 of the TSAR.

3.4.1 TSAR Section 4.1 - Containment System Performance Appendix G to the TLTR, outlines the methods, approach, and scope of plant-specific containment analyses that have been used by the licensee in support of the TPO uprate. The TLTR states that the previous LOCA containment analyses are bounding for the TPO uprate because they considered 2 percent uncertainty in the RTP as required by the previous methodology. Although the TPO uprate will increase the nominal operating conditions slightly, the required bounding conditions for the limiting analytical cases do not change from the previously documented bounding conditions. The TLTR states that there is no effect of the TPO uprate on the containment pressure and temperature response and dynamic loads due to LOCA and SRV actuation.

As stated in TSAR Section 4.1, the following containment analyses were performed at 102 percent of the CLTP of 3,840 MWt, which bounds the TPO power level of 3,902 MWt:

Short term response for peak containment pressure Short term response for drywell gas temperature Long term suppression pool temperature response for bulk pool temperature Long term suppression pool temperature response for local temperature with SRV discharge

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION Subcompartment pressurization loads LOCA dynamic loads SRV discharge loads The licensee stated that although the nominal operating conditions change slightly because of the TPO uprate, the initial conditions and inputs remain the same as in the current analysis at the CLTP in the license amendment for the EPU.

The NRC staff concludes that the above analyses are not affected for the TPO power uprate because they were performed assuming 102 percent CLTP which bounds the TPO power uprate.

The NRC staff noted during its review that information was missing from TSAR Section 4.1, regarding the containment temperature response concerning equipment environmental qualification. By letter dated December 22, 2017, the licensee provided the results of its analysis of the limiting break size under the TPO uprate conditions stating that the plant-specific containment analyses for CLTP, including the small steam line break analysis, were performed at 102 percent of the CLTP.

The NRC staff concludes that the containment temperature response for equipment environmental qualification acceptable at the TPO uprate condition because it was determined at 102 percent of CLTP power which bounds the TPO power uprate.

During its review, the NRC staff noted that information regarding the peak containment wall temperature for structural analysis was missing from TSAR Section 4.1. By letter dated December 22, 2017, the licensee provided the results of this analysis including the limiting break size under the TPO uprate condition stating that the current plant-specific containment analyses was performed at 102 percent of the CLTP.

The NRC staff concludes that the current peak containment wall temperature acceptable at the TPO uprate condition because it was determined at 102 percent of CLTP power which bounds the TPO power uprate.

The TPO uprate LOCA containment pressure and temperature response, LOCA subcompartment pressurization, LOCA dynamic loads, and SRV loads evaluation by the licensee concluded that re-performing a structural analysis was not needed because the bounding events remain unchanged from the CLTP. The containment pressure and temperature and dynamic loads have margins that are not affected by the TPO uprate.

The NRC staff concludes that the containment system performance for the TPO is acceptable because the containment system continues to meet the requirements of: (1) GDC 4 because the dynamic loads due to LOCA subcompartment pressurization, LOCA dynamic loads on safety related containment structures and components, SRV piping loads, and SRV discharge loads on containment are bounded by the dynamic loads in the current analysis; (2) GDC 16 because the LOCA containment pressure and temperature under the TPO uprate conditions is bounded by the current analysis for containment pressure and temperature; and therefore, it will be maintained as a leaktight barrier to a release of radioactivity to the environment; and (3) GDC 50 because the LOCA containment pressure and temperature under the TPO uprate

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION conditions are bounded by the current analysis; and therefore, in association with its heat removal system, the design leakage rate will not be exceeded.

TSAR Section 4.1.1 - Generic Letter (GL) 89-10 The NRC GL 89-10 Safety-Related Motor-Operated Valve Testing and Surveillance (ADAMS Accession No. ML031150300), extended the scope of the program outlined in IE Bulletin 85-03, Motor-Operated Valve Common Mode Failures during Plant Transients Due to Improper Switch Settings, and its Supplement 1, to include all safety-related motor-operated valves (MOVs) as well as all position-changeable MOVs. The licensees were requested to develop and implement a program to ensure that the switch settings (torque, torque bypass, position limit, and overload) on the safety-related MOVs are selected, set, and maintained correctly to accommodate the maximum differential pressures expected on these valves during both normal and abnormal events within the design basis during the life of the plant.

As stated in TSAR Section 4.1.1, the licensee stated that the current analyses for the GL 89-10 MOVs at HCGS were either based on 102 percent of CLTP or are consistent with the plant conditions expected at the TPO uprate. The operating pressure and temperature for the safety-related MOVs within the GL 89-10 program do not increase, except for FW valves, which have a small increase in operating temperature not requiring modification. Therefore, the GL 89-10 program for the safety-related MOVs remains unchanged following TPO uprate and the MOVs remain capable of performing their safety-related functions.

The NRC staff considers the evaluation of the GL 89-10 MOVs acceptable because the current analyses bound the operating conditions at the TPO uprate, and no changes are necessary to meet the functional requirements of these MOVs at the TPO uprate conditions.

TSAR Section 4.1.2 - GL 96-05 The NRC GL 96-05 Periodic Verification of Design-Basis Capability of Safety-Related Motor-Operated Valves (ADAMS Accession No. ML031110010), stated the periodic verification of the capability of safety-related MOVs to perform their safety functions consistent with the current licensing bases of nuclear power plants. This GL provides more complete guidance regarding periodic verification of safety-related MOVs and supersedes GL 89-10 and its supplements with regard to MOV periodic verification.

As stated in TSAR Section 4.1.2, the licensee stated that the current evaluation of the GL 96-05 program was reviewed and determined to have no effects related to the TPO uprate. In a supplement dated December 22, 2017, the licensee justified why the TPO uprate would not affect the current evaluation of GL 96-05. In its response, the licensee stated that GL 96-05 addressed the same valves as addressed in GL 89-10. Therefore, for the same reasons that the response to GL 89-10 is unaffected, the response to GL 96-05 is not affected by the TPO.

As a part of the TPO uprate program, the licensee reviewed all the plant modifications post-EPU for possible impacts on MOV performance, and determined that no modifications change the conclusions from the previous evaluation.

The NRC staff considers the licensees evaluation of the above reasonable; and therefore, determines that the GL 96-05 valves would not be affected by the TPO uprate.

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION TSAR Section 4.1.3 - Generic Letter 95-07 The NRC GL 95-07 Pressure Locking and Thermal Binding of Safety-Related Power-Operated Gate Valves (ADAMS Accession No. ML031070145) requested that licensees perform or confirm that they previously performed, (1) evaluations of operational configurations of safety-related, power-operated (including motor-, air-, and hydraulically-operated) gate valves for susceptibility to pressure locking and thermal binding, and (2) further analyses, and any needed corrective actions, to ensure that safety-related power-operated gate valves that are susceptible to pressure locking or thermal binding are capable of performing the safety functions within the current licensing bases of the facility.

In TSAR Section 4.1.3, the licensee stated that the criteria for susceptibility to pressure locking or thermal binding were reviewed at the TPO uprate and it was determined that the expected slight changes in operating or environmental conditions from the TPO uprate would have no effect on the functioning of power-operated gate valves within the scope of GL 95-07, and the valves remain capable of performing their safety-related functions.

The NRC staff agrees that the licensees evaluation with respect to GL 95-07 remains valid at the TPO uprate conditions because there is reasonable assurance that minor changes in the operating or environmental conditions would not affect the operation of the power-assisted safety related gate valves.

TSAR Section 4.1.4 - Generic Letter 96-06 The NRC GL 96-06 Assurance of Equipment Operability and Containment Integrity During Design-Basis Accident Conditions (ADAMS Accession No. ML031110021) identifies the following potential problems with equipment operability and containment integrity during design basis accident (DBA) conditions: (1) cooling water systems serving the containment air coolers may be exposed to water hammer during postulated accident conditions, (2) cooling water systems serving the containment air coolers may experience two-phase flow conditions during postulated accident conditions, and (3) thermally induced over-pressurization of isolated water-filled piping sections in containment could jeopardize the ability of accident-mitigating systems to perform their safety functions and could also lead to a breach of containment integrity via bypass leakage. The NRC GL 96-06 questioned whether the higher heat loads at accident conditions could potentially cause steam bubbles, water hammer, and two-phase flow due to the higher outlet temperatures from cooled components, particularly the containment fan coolers.

As stated in TSAR Section 4.1.4, the licensee stated that containment design temperatures and pressures in the current GL 96-06 evaluation are not exceeded under post-accident conditions for the TPO uprate. In its supplement dated December 22, 2017, the licensee clarified the containment design temperature and pressure at which the current GL 96-06 evaluation was performed. The licensee stated that the GL 96-06 problems identified in item (1) through (3) above are not based on a specific design pressure or temperature and are also valid for post-EPU or TPO conditions.

The licensee evaluated the above three problems identified in GL 96-06 that could affect containment integrity and the operability of safety-related equipment during accident conditions:

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1. For problem (1), the licensee mitigated the problem (i.e., did not rely on an evaluation) by restricting the use of the drywell coolers to avoid steam voids to remove the possibility of water hammer.

2. For problem (2), the problem was determined not to be applicable because the licensee does not credit drywell coolers for mitigation of an accident.

3. For problem (3), the licensee determined the problem was applicable to five drywell penetrations. The licensee installed pressure relief valves in these lines to prevent over-pressurization. Therefore a change in maximum containment pressure during an accident would not affect the potential for thermally induced over-pressurization of the piping passing through these penetrations.

The NRC staff determines that the evaluation with respect to GL 96-06 is acceptable because it was performed using parameters corresponding to operation at 102 percent 3,952 MWt, which bounds the TPO power of 3,902 MWt.

Generic Letter 89-16 The NRC GL 89-16 Installation of a Hardened Wetwell Vent (ADAMS Accession No. ML031140220) stated the advantages of installing a hardened containment (wetwell) vent and requested information from licensees on installation of such a vent. This was a result of the NRCs BWR Mark I Containment Performance Improvement Program. The HCGS TSAR did not provide an evaluation of the hardened wetwell vent at the TPO uprate power.

In its supplement dated December 22, 2017, the licensee provided quantitative results of the evaluation of the vent at the TPO uprate power level. In its response, the licensee stated that the hardened containment vent system (HCVS) capacity was calculated in response to the requirement in NRC Order EA-13-109 dated February 12, 2015 (ADAMS Accession No. ML14332A154) which is the same requirement of a vent installed per GL 89-16. The licensees calculation demonstrated that the HCVS has a capacity to vent saturated steam at a flow rate equivalent to one percent of 3,917 MWt at the lesser of primary containment design pressure and the primary containment pressure limit. The UFSAR Table 6.2-1 provides the containment internal design pressure of 62 psig which is less than the primary containment pressure limit.

The NRC staff interim evaluation of the HCGS overall integrated plan for Phase 1 of Order EA-13-109 states that the HCVS is sized for venting steam/energy at a nominal one percent of reactor thermal power assuming a future MUR power uprate. Therefore, the NRC staff finds the licensees evaluation of GL 89-16 acceptable. The NRC staff concludes that the licensee demonstrated the HCVS capacity bounds the TPO or MUR uprate power level of 3,902 MWt.

TSAR Section 4.1.5 - Containment Coatings 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 reviewed the protective coating systems used inside containment for their suitability for and stability under design basis LOCA conditions, considering temperature, pressure, radiation, and chemical effects on the ECCS.

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION In TSAR Sections 4.1 and 4.1.5, the licensee stated that the nominal operating conditions change slightly, but the initial conditions for the containment analysis do not change from the CLB. The licensee also stated that the service level 1 coatings are qualified to 340 °F, 70 psi, and 1 x 109 rads. The licensee concluded that the containment coatings continue to bound DBA temperatures, pressures, and radiation at the uprated power conditions.

The NRC staff reviewed the post-uprate design basis LOCA conditions and compared them to the qualification of the service level 1 coatings, and determined that the qualifications of the service level 1 coatings in containment continue to bound the predicted conditions for the proposed uprated conditions. Therefore, the staff has reasonable assurance that the coatings will not be adversely impacted by the power uprate conditions and finds the MUR power uprate acceptable with respect to protective coatings. Thus, the protective coatings will continue to meet the requirements of 10 CFR Part 50, Appendix B.

3.4.2 TSAR Section 4.2 - Emergency Core Cooling Systems (ECCSs)

The ECCSs are provided to limit the fuel cladding temperature to less than the limits of 10 CFR 50.46 in the event of a LOCA. The systems are designed to cool the reactor core over the complete range of breaks sizes in the RCPB. The systems initiate automatically when required regardless of the availability of offsite power supply or the normal generating system of the station. The ECCSs are: (1) HPCI, (2) automatic depressurization system (ADS), (3) CS, and (4) LPCI. Evaluations for each of these systems under the TPO power uprate are stated below.

TSAR Section 4.2.1 - High Pressure Coolant Injection (HPCI)

The HPCI system provides and maintains an adequate coolant inventory in the reactor vessel to limit the fuel cladding temperature for postulated small breaks in the RCPB. The high pressure system is needed because the reactor vessel depressurizes slowly during small breaks. Higher pressures in the RCS prevent the LPCI from injecting coolant. The HPCI system flow is propelled by a reactor-steam driven turbine and high-pressure pump that injects coolant into the RCS.

TLTR Section 5.6.7 provides a generic evaluation of the ability of the HPCI system to perform its function. The generic evaluation is applicable at the proposed TPO uprate because the ability of the HPCI system function was demonstrated in the evaluation for EPU. This was conducted at 102 percent CLTP which bounds the TPO power.

The NRC staff determines that the licensees evaluation of the HPCI system is acceptable because no changes are necessary to meet the functional requirements and the current evaluation bounds the TPO uprate conditions.

Based on the above evaluation, the HPCI system continues to meet the requirements of GDC 33 under the TPO uprate conditions because it will perform its required function during leakage and small breaks in the RCPB so that the fuel cladding temperature is not exceeded beyond its design limit.

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION TSAR Section 4.2.2 - Core Spray The CS system consists of two independent pump loops that deliver coolant to the spray spargers over the reactor core. The system is automatically actuated by conditions indicating a break in the RCPB. However, the CS system can only deliver coolant to the core after the reactor is depressurized to a pressure below the CS pump shutoff head. The CS system continues to supply coolant for the long-term core cooling in the event of a LOCA.

TLTR Section 5.6.10 provides a generic evaluation of the ability of the CS system to perform its function at TPO conditions. The licensee stated that the generic evaluation is applicable at the proposed TPO uprate because the ability of the CS system function is demonstrated in the EPU evaluation at 102 percent CLTP which bounds the TPO power.

Based on the above evaluation, the NRC staff concludes that the CS system is acceptable at TPO conditions because it will continue to meet the requirements of GDC 35 by providing water to the reactor vessel for long-term cooling in the event of a LOCA.

TSAR Section 4.2.3 - Low Pressure Coolant Injection (LPCI)

The LPCI is one of the operating modes of the RHR system. The primary purpose of this mode is to provide reactor vessel coolant inventory makeup following large break LOCAs. In the event of a small break LOCA, the LPCI is initiated to provide reactor inventory makeup after the reactor is depressurized by the ADS. The RHR system is automatically lined-up and initiated in its LPCI mode on a low reactor vessel water level signal or a high drywell pressure signal. Four RHR pumps deliver water from the suppression pool to four separate reactor vessel nozzles and inject directly into the core shroud region.

TLTR Section 5.6.4 provides a generic evaluation of the ability of the LPCI mode of the RHR system to perform its function. The licensee stated that the generic evaluation is applicable at the proposed TPO uprate because the ability of the RHR system function in the LPCI mode is demonstrated in the EPU evaluation, which was 102 percent CLTP and bounds the TPO uprate power.

The NRC staff considers the evaluation of the LPCI mode of the RHR system acceptable because no changes are necessary to meet its functional requirements, and its current evaluation bounds the TPO uprate conditions. Therefore, the NRC staff concludes that the LPCI mode of the RHR system will continue to meet the requirements of GDC 35 under the TPO conditions because it will continue to provide water for core cooling following a LOCA.

TSAR Section 4.2.4 - Automatic Depressurization System The ADS allows depressurization of the RCS for small break LOCAs in case the HPCI system cannot maintain the reactor water level or it malfunctions. Sufficient depressurization of the coolant system allows the CS system and the LPCI mode of the RHR system to operate as a backup to the HPCI system for maintaining the reactor water level to protect the fuel cladding.

The ADS is independent of any other ECCS subsystem and uses selected SRVs for depressurization of the reactor vessel. Each of the SRVs used for ADS is equipped with an air accumulator with a check valve arranged to ensure that the SRVs can be held open following failure of the air supply to the accumulators.

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION TLTR Section 5.6.8 provides a generic evaluation of the ability of the ADS to perform its function. The licensee stated that the generic evaluation is applicable at the proposed TPO uprate because the ability of the ADS is demonstrated in the EPU evaluation at 102 percent CLTP which bounds the TPO power. The ADS initiation logic and SRV control is not affected by the TPO uprate because it is independent of the reactor thermal power.

The NRC staff determines that the licensees evaluation of the ADS is acceptable because no changes are necessary for the ADS system to meet its functional requirements, and the current evaluation bounds the TPO uprate conditions. Therefore, the NRC staff concludes that the ADS will continue to meet the requirements of GDC 34 and -35 under the TPO uprate conditions because it will depressurize the reactor vessel during small break LOCAs to allow the CS system and the LPCI mode to cool the core.

TSAR Section 4.2.5 - Emergency Core Cooling System Net Positive Suction Head The RHR and CS system pumps are ECCS pumps that draw water from the suppression pool during LOCAs and special events. The LPCI mode of the RHR system injects makeup water in the reactor in the short-term, and its SPC mode performs long term SPC cooling during these events. A net positive suction head (NPSH) analysis is performed to confirm positive NPSH margin, which is a measure of the pumps ability to avoid excessive cavitation so that it can perform its safety function(s). The special events to be analyzed at the TPO power are Appendix R Fire, station blackout (SBO), and anticipated transient without scram (ATWS). The margin is measured as the difference between the NPSH available (NPSHa) at the pump inlet and the NPSH required (NPSHr) by the pump to prevent cavitation. The NPSHa depends on the suppression pool temperature response and is time-dependent.

In the EPU analysis for HCGS, the licensee stated that the suppression pool temperature response was analyzed at 102 percent CLTP for the large break and small steam line break LOCAs and therefore the suppression pool temperature responses would bound the responses at the TPO power. For the SBO and ATWS events, the licensee stated that the suppression pool temperature response analysis was performed at greater than 102 percent CLTP, and therefore the responses would also bound the TPO power response.

In a supplement dated December 22, 2017, the licensee provided additional information concerning the EPU values, or the values for the most current analysis, for the limiting suppression pool temperature response transients for the LOCAs, SBO event, and ATWS event. The licensee stated that no changes have occurred since the EPU analysis for LOCAs.

The limiting break from an NPSH standpoint remains a double-ended pipe break of a recirculation suction line. The licensee also confirmed that the peak long-term suppression pool temperature of 212.3°F was calculated assuming 102 percent of EPU power. The peak short-term non-limiting suppression pool temperature that was conservatively calculated at 102 percent of 3,952 MWt was 167.3°F.

The licensee confirmed that for the limiting ATWS event, the 199°F peak suppression pool temperature remained the same for the TPO as in the EPU analysis because it was analyzed at a power level of 3,952 MWt which bounds the TPO power.

The licensee further stated that for an SBO event, the re-calculated peak suppression pool temperature was 204.6°F. The increase from the EPU peak temperature of 198.0°F was based

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION on the TPO power and incorporated information from the GEH Safety Communication (SC) 10-18 dated December 2010. The staff reviewed the information and determined that the change in peak suppression pool temperature is acceptable since it is still within the acceptance criteria.

The LOCA NPSH analysis for the CS and the RHR pumps for the EPU remains bounding because it was performed using a peak suppression pool temperature of 218°F which is greater than the SBO event revised peak suppression pool temperature of 204.6°F. However, during an SBO, since the reactor level is maintained using the available HPCI and RCIC systems, which take suction from the condensate storage tank, the SBO suppression pool temperature response would not affect the analysis.

In a supplement dated December 22, 2017, the licensee provided additional information including the analysis results for the RHR and CS pump limiting NPSHa at the pump inlet for the LOCA short-term (i.e., up to 600 seconds from its initiation in case the pumps operate with runout flows up to 600 seconds), LOCA long-term, SBO, and ATWS without crediting containment accident pressure (CAP) developed during these events. The limiting NPSH margin (NPSHa - NPSHr) and margin ratio (NPSHa/NPSHr) were calculated in accordance with the existing licensing basis using a conservative bounding suppression pool bulk temperature of 218°F for a double-ended recirculation suction line break in the long-term, without crediting CAP. For the limiting RHR pump D, the NPSHa is 5.67 ft, NPSHr is 4.0 ft, with a margin of 1.7 ft, and a margin ratio of 1.5. For both CS pumps, the NPSHa is 6.8 ft, NPSHr is 5.6 ft, margin of 1.2 ft, and a margin ratio of 1.2. The NPSHr values are taken from the original manufacturers pump curves at the maximum achievable flow. For RHR pump, the maximum achievable flow was determined during initial startup testing with the LPCI injection valve fully open and corrected for test conditions. For the CS pump, the maximum achievable flow was determined using configurations and assumptions that maximized flow through the pumps.

Following the NRC issuance of the HCGS EPU amendment, the Advisory Committee for Reactor Safeguards expressed concerns to the NRC staff regarding the quantification of conservatism in the suppression pool temperature response and the use of CAP in the NPSHa.

In response to these concerns, the NRC issued SECY-11-0014 which provided guidance in (ADAMS Accession No. ML102110167) to resolve the issues raised by the Advisory Committee for Reactor Safeguards. This guidance was approved by the Commission in SRM-SECY-11-0014 on March 15, 2011. Since the NPSH analysis did not credit CAP, and is also not required for the TPO uprate, the NRC staff concludes that the guidance in SECY-11-0014 is not applicable to HCGS MUR.

For the Appendix R event, the following statement in Section 6.7.1 of the EPU submittal stated that the limiting Appendix R fire event was analyzed assuming CLTP and constant pressure power uprate. Further, in a supplement to the EPU LAR dated March 30, 2007, the licensee stated that the Appendix R Fire evaluation was performed at 3,840 MWt, which is 100 percent of the EPU. The Appendix R Fire event NPSH analysis does not credit CAP; and therefore, the guidance in SECY-11-0014 is not applicable.

The NRC staff requested clarification as to the power level for which this event was analyzed.

In a supplement dated January 22, 2018, the licensee provided corrections to the EPU submittal (ADAMS Accession No. ML032690072) and the supplement dated March 30, 2007 (ADAMS Accession No. ML070960103), stating that the Appendix R Fire containment analysis was

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION performed at 102 percent of the EPU power (i.e., 1.02 x 3,840 = 3,916.8 MWt, versus the original statement of 3,840 MWt).

In its supplement dated January 22, 2018, the licensee stated the key assumptions and the result of the peak suppression pool temperature (205.9°F) for the limiting Appendix R Fire event analysis performed for the EPU. The licensee also provided the peak suppression pool temperature (206.3°F) from a subsequent re-analysis. In the re-analysis, the assumption of the initiation time of the SPC mode of the RHR system was conservatively changed from 20 minutes to 60 minutes into the event.

The licensee also stated that the multiple spurious operation (MSO) compliance assessment was completed in July 2012 which is dated after its analysis of record. The assessment was performed based on Nuclear Energy Institute (NEI) 00-01, Guidance for Post Fire Safe-Shutdown Circuit Analysis, (endorsed by the NRC in RG 1.189, Revision 2, with exceptions) and utilized the MSO expert panel method in developing credible MSO scenarios. The licensees review of the MSO scenarios showed that none of the scenarios affect the analysis key assumptions or challenge the RHR pump NPSH or result in the increase of maximum predicted suppression pool temperatures.

The NRC staff determines that the licensees Appendix R Fire event analysis was performed at 102 percent of the CLTP; and therefore, bounds the TPO uprate. The NRC staff also determines that the licensees MSO compliance assessment is acceptable because it was performed in accordance with NRC-accepted guidance using the MSO expert panel process in developing the possible MSO scenarios. The Appendix R Fire event NPSH analysis is bounded by the LOCA NPSH analysis for the RHR pump because the Appendix R Fire event peak suppression pool temperature of 206.3°F is less than the LOCA peak suppression pool temperature of 218°F, and, therefore, is acceptable to the NRC staff.

Based on the above evaluation, the NRC staff concludes that the LPCI and SPC modes of the RHR system continue to meet the requirements of GDC 34, 35, and 38 under the TPO uprate conditions because the RHR pumps will have adequate NPSH to perform their safety function for core cooling and containment heat removal during a LOCA, SBO, ATWS and Appendix R Fire events. Further, the NRC staff concludes that the CS pump NPSH analysis continues to meet the requirement of GDC 35 because it will have adequate NPSH for core cooling during a LOCA.

3.4.3 TSAR Section 4.3 - Emergency Core Cooling System Performance The ECCS is designed to provide protection against a postulated LOCA caused by ruptures in the primary system piping. As stated in TSAR Section 4.3, the current LOCA analysis for HCGS was performed at power levels assuming 102 percent of CLTP; and therefore, bound the TPO uprate conditions consistent with Appendix K to 10 CFR Part 50. The LOCA analysis results for the peak cladding temperature (PCT) and oxidation of fuel cladding for the GE14 and GNF2 fuel provided by the licensee in TSAR Table 4-1 are given in the Table 1 below. The HCGS Cycle 20 core has GE14 fuel; the Cycle 21 core will have residual GE14 and a first reload of GNF2 fuel.

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION Based on the above evaluation, the NRC staff concludes there is reasonable assurance that the ECCS performance analysis continues to meet the requirements of 10 CFR 50.46, 10 CFR Part 50 Appendix K, GDC 33, -34, and -35 at the TPO uprated conditions.

3.4.4 TSAR Section 4.4 - Main Control Room Atmosphere Control System As stated in Appendix J, Section J.2.3.8 of the TLTR, the main control room atmosphere control system was considered to assure continued habitability of the control room following a postulated accident at TPO uprate conditions. The TLTR states that compliance is unchanged if the system has been previously evaluated for accident conditions at 102 percent of CLTP.

TSAR Section 4.4 confirmed that the main control room envelope and atmosphere control system had been previously evaluated using radiation release characteristics based on 102 percent of CLTP. The assumed radiation release remains bounding for operation at TPO conditions, and, thus, the NRC staff concludes that control room atmosphere control system is acceptable for TPO conditions.

3.4.5 TSAR Section 4.5 - Standby Gas Treatment System As stated in Appendix J, Section J.2.3.9 of the TLTR, the standby gas treatment system (SGTS) is designed to minimize offsite and control room dose rates during venting and purging of the containment atmosphere. The TLTR states that the capability of the SGTS is unchanged at TPO conditions if the system has been previously evaluated for accident conditions at 102 percent of CLTP.

TSAR Section 4.5 confirmed that the filtration, recirculation, and ventilation system (referred to as the SGTS in the TLTR) maintains the secondary containment at a slightly negative pressure during venting and purging of the primary containment atmosphere under abnormal conditions.

The licensee determined that the system could accommodate conditions associated with a DBA at 102 percent of CLTP. Therefore, the NRC staff concludes that the system remains capable of performing its safety function at TPO conditions.

3.4.6 TSAR Section 4.6 - Main Steam Isolation Valve Leakage Control System HCGS does not have a MSIV leakage control system, so this review is not applicable.

3.4.7 TSAR Section 4.7 - Post-LOCA Containment Atmosphere Control System The original licensing basis of the containment atmosphere control system was to ensure a combustible mixture of oxygen and hydrogen would not develop in the containment atmosphere post-LOCA through the use of hydrogen recombiners. This evaluation is no longer applicable to HCGS because the containment is inerted with nitrogen during plant operation and the atmosphere control function of the system is no longer necessary to satisfy the requirements of 10 CFR 50.44, Combustible Gas Control for Nuclear Power Reactors. Therefore, review of this system is not applicable to the HCGS TPO uprate.

3.5 TSAR Section 5.0 - Instrumentation and Control The following provides the NRC staffs technical review of the topics in Section 5.0 of the TSAR.

The NRC staffs review in the area of instrumentation and controls covers the proposed plant-

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION 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 RIS 2002-03.Section I Feedwater flow measurement technique and power measurement uncertainty, and Section VIII Changes to technical specifications, protection system settings, and emergency system settings of Attachment 1, to RIS 2002-03 provide the criteria the NRC staff used to confirm that the licensees implementation of the proposed FW flow measurement device is consistent with staff-approved topical reports ER-80P, Revision 0, and ER-157P, Revision 8. The NRC staff also reviewed the power measurement uncertainty calculations to ensure that (1) the instrumentation and control signal ranges and analytical limits (ALs) for setpoints maintain adequate operating margins between plant operating parameters and trip values, (2) the conservatively proposed uncertainty value of 0.34 percent correctly accounts for all uncertainties associated with power level instrumentation errors, and (3) the uncertainty calculations meet the relevant requirements of 10 CFR Part 50, Appendix K.

3.5.1 TSAR Section 5.1 - Nuclear Steam Supply System Monitoring and Control Section 5.1 of the TSAR states that instrumentation and controls that directly interact with or control the reactor are usually considered within the nuclear steam supply system (NSSS). The NSSS monitoring and control systems evaluated in TSAR Section 5.1 are stated below.

TSAR Section 5.1.1.1 - Average Power Range Monitors, Intermediate Range Monitors, and Source Range Monitors As stated in the UFSAR Section 7.5, the average range power monitor (APRMs) channels provide the primary indication of neutron flux within the core. The APRM channels receive input signals from the local power range monitors (LPRMs) within the reactor core to provide an indication of the power distribution and local power changes. The APRMs are capable of generating a scram trip signal in response to average neutron flux increases resulting from abnormal operational transients in time to prevent fuel damage. To ensure that the APRMs are accurately indicating the true core average power, the APRMs are calibrated to the reactor power calculated from a heat balance.

The licensee stated in TSAR Section 5.1.1.1 that the APRMs are re-calibrated to indicate 100 percent at the TPO RTP level of 3,902 MWt. The relevant setpoints, which include the APRM neutron flux-upscale scram, APRM simulated thermal power (STP) upscale high flow clamped (scram) and STP upscale high flow clamp rod block setpoints, are expressed in units of percent of licensed power and remain unchanged. The flow-biased APRM trips, expressed in units of absolute thermal power (i.e., MWt), remain the same. The licensee stated that this approach follows the guidelines of the TLTR, Section 5.6.1 and Appendix F, which is consistent with the NRC-approved practice for GE BWR uprates in topical report NEDC-32424P-A Generic Guidelines for General Electric Boiling Water Reactor Extended Power Uprate.

The licensee stated that no adjustment is needed to ensure the intermediate range monitors (IRMs) have adequate overlap with the source range monitors and APRMs for the TPO.

However, normal plant surveillance procedures may be used to adjust the IRM overlap with the APRMs. The IRM channels have sufficient margin to the upscale scram trip on the highest range when the APRM channels are reading near their downscale alarm trip because the change in APRM scaling is very small for the TPO uprate.

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION The NRC staff reviewed the licensees LAR, including the TSAR report, and determines that the licensee provided sufficient justifications that the only change required to the scram setpoints is for the APRM STP flow-biased trips. Therefore, the NRC staff concludes that the licensee has followed the practice defined in the TLTR and has met the regulatory requirements of 10 CFR Part 50, Appendix K.

TSAR Section 5.1.1.2 - Local Power Range Monitors and Traversing In-Core Probes In TSAR Section 5.1.1.2, the licensee stated that the flux at some LPRMs will increase at the TPO RTP level. However, the small change in the power level is not a significant factor to the neutronic service life of the LPRM detectors and radiation level of the traversing in-core probes (TIPs). The increase in flux does not change the number of cycles in the lifetime of the detectors. The LPRM accuracy at the increased flux continues to be within specified limits, and the LPRMs are designed as replaceable components. The TIPs are stored in shielded chambers and the radiation protection program for normal plant operation can accommodate a small increase in radiation levels of the TIPs.

The NRC staff reviewed the licensees LAR, including the TSAR, and determines that the proposed TPO does not have any impact to the LPRMs and TIPs. Therefore, the NRC staff concludes that licensee has followed the practice defined in TLTR and has met the regulatory requirements of 10 CFR Part 50, Appendix K.

TSAR Section 5.1.1.3 - Rod Block Monitor The licensee stated in TSAR Section 5.1.1.3 that the rod block monitor (RBM) instrumentation is referenced to an APRM channel. Because the APRM has been rescaled, there is only a small effect on the RBM performance due to the LPRM performance at the higher average local flux.

The licensee concluded that the RBM instrumentation is not significantly affected by the TPO uprate conditions, and the RBM trip setpoint remains unchanged.

In a supplement dated December 19, 2017 (ADAMS Accession No. ML17353A831), the licensee clarified its statements in the TSAR concerning a small effect and not significantly affected. The licensee stated that this conclusion was based on the LTR NEDC-33004P-A, Constant Pressure Power Uprate (ADAMS Accession No. ML032170315), Section 5.1.1.3, LPRM performance refers to accuracy; and per Section 5.1.1.2, LPRM accuracy remains within analyzed limits at uprate conditions. Further, the licensee stated that not significantly affected means that the accuracy effect on the RBM due to the LPRMs is bounded by the analyzed performance of the RBM. Since NEDC-33004P-A is applicable to EPUs up to 120 percent of the OLTP, this statement is applicable to HCGS at TPO conditions.

The NRC staff reviewed the licensees response and concludes that the proposed changes are acceptable, the licensee has followed the practice defined in TLTR, and has met the regulatory requirements of 10 CFR Part 50, Appendix K.

TSAR Section 5.1.2 - Rod Worth Minimizer The licensee stated in TSAR Section 5.1.2 that the rod worth minimizer (RWM) does not perform a safety-related function. The function of the RWM is to support the operator by enforcing rod patterns until reactor power has reached appropriate levels. The NRC staffs

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION evaluation for the power-dependent setpoints for the RWM is stated in SE Section 3.11.5 (TSAR Section 5.3.8).

3.5.2 TSAR Section 5.2 - Balance-of-Plant Monitoring and Control Section 5.2 of the TSAR addresses various control and protective systems that may be affected by changes in operating P-Ts. These systems include the reactor pressure control system (PCS), the main turbine electro-hydraulic control (EHC) system, the FW control system, and various leak detection systems. The control systems are not safety-related, but excessive challenges to the systems may increase the frequency of transients. The leak detection systems provide a protective function, but, if actuated, also introduce plant transients.

The licensee stated that operation at the TPO RTP level has a minimal effect on the BOP system instrumentation and control devices. All instrumentation with control functions has sufficient range and adjustment capability for use at the TPO uprate conditions. The plant-specific instrumentation and control design and operating conditions are bounded by those used in the evaluations contained in the TLTR.

The licensee stated that the approximate 2 percent increase in FW flow associated with TPO uprate is within the current control margin of these systems. No changes in the operating reactor water level or reactor water level trip setpoints are required for the TPO uprate. Per the guidelines of TLTR Appendix L, the performance of the FW level control systems will be recorded at 95 percent and 100 percent of CLTP and confirmed at the TPO level during power ascension. These checks will demonstrate acceptable operational capability and will utilize the methods and criteria described in the original startup testing of these systems.

TSAR Section 5.2.1 - Pressure Control System Section 5.2.1 of the TSAR included a description of the reactor PCS. The PCS controls reactor pressure by modulating turbine control valve position and through use of the turbine bypass system when a rapid reduction in turbine steam flow occurs. The analysis presented in the TSAR indicated that no modification to the turbine bypass system was necessary, but the high pressure turbine would be modified to provide pressure control margin at the full TPO uprate power. Therefore, adequate reactor pressure control would be maintained at TPO uprate conditions. Additionally, no modifications would be required for the controls nor alarm annunciators provided in the main control room. The required adjustments are limited to tuning the control settings that may be required to operate optimally at the TPO uprate power level.

PCS tests, consistent with the guidelines in TLTR Appendix L, will be performed during the power ascension phase (stated in SE Section 3.10.4).

TSAR Section 5.2.2 - EHC Turbine Control System and Section 5.2.3 - Feedwater Control System Sections 5.2.2 and 5.2.3 of the TSAR addressed the operation of the EHC and FW control systems respectively. The analyses presented in the TSAR indicated that these systems had adequate control margin to accommodate the small changes in flow associated with operation at TPO uprate conditions. Therefore, adequate EHC and FW control system performance would be maintained at TPO uprate conditions. Additionally, confirmation testing will be performed during power ascension as stated in SE Section 3.10.4.

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION TSAR Section 5.2.4 - Leak Detection System Section 5.2.4 of the TSAR stated the effect of the TPO on leak detection system margin. For operation at TPO uprate conditions, the nominal reactor vessel steam dome pressure and temperature would be unchanged, and the FW temperature would increase by only a small amount (approximately 2°F). The majority of leak detection system setpoints consider the maximum vessel temperature in the steam dome for flow through the system, and are, therefore, unaffected by the TPO uprate. The FW temperature increase results in a negligible increase in normal temperatures within the steam tunnel and RWCU areas. Therefore, adequate margins to actuation of the leak detection systems on high area temperatures are maintained for operation at TPO conditions.

The NRC staffs review of the changes to the safety-related BOP system setpoints is provided in SE Section 3.5.3 (TSAR Section 5.3.16).

In summary, based on the above discussion and the NRC staffs review of the licensees LAR and TSAR report, the NRC staff finds that the licensee provided sufficient justifications for the proposed modifications to the BOP system. Therefore, the NRC staff concludes that the licensee has followed the practice defined in TLTR and has met the regulatory requirements of 10 CFR Part 50, Appendix K.

3.5.3 TSAR Section 5.3 - Technical Specification Instrument Setpoints The NRC staffs review of the TS changes for this LAR is contained in SE Section 3.11.

TSAR Section 5.3.1 - High Pressure Scram The licensee stated that because there is no increase in nominal reactor operating pressure with the TPO uprate, the scram analytic limit on reactor high pressure is unchanged. The NRC staff determines this conclusion is acceptable since the licensee has followed the practice defined in TLTR and, thus, has met the regulatory requirements of 10 CFR Part 50, Appendix K.

TSAR Section 5.3.2 - Hydraulic Pressure Scram The licensee stated that the analytic limit for the turbine hydraulic pressure (low oil pressure trip) that initiates the turbine-generator trip scram at high power remains the same as for CLTP. No modifications are being made to the turbine hydraulic control systems for TPO; actuation of these safety functions remains unchanged for this TPO.

The NRC staff determines that no modifications are required since the licensee has followed the practice defined in TLTR, thus, the licensee has met the regulatory requirements of 10 CFR Part 50, Appendix K.

TSAR Section 5.3.3 - High Pressure Recirculation Pump Trip The TPO uprate does not change any equipment or performance characteristics because the operating conditions such as operating pressure, SRV setpoints, and maximum rod line, do not change. Thus, the high pressure recirculation pump trip logic (i.e., ATWS-RPT) will not be modified and the NRC staff concludes this protective measure will continue to meet the requirements of 10 CFR 50.62.

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION TSAR Section 5.3.4 - Safety Relief Valve The licensee stated that there is no increase in reactor operating dome pressure and the SRV analytic limits are not changed.

The NRC staff determines that no modifications are required since the licensee has followed the practice defined in TLTR, thus, the licensee has met the regulatory requirements of 10 CFR Part 50, Appendix K.

TSAR Section 5.3.5 - Main Steam Line High Flow Isolation The licensee stated in TSAR Section 5.3.5 that the TS AV of this function is expressed in terms of differential pressure (psid). The licensee stated that the MS flow increased approximately 2 percent, but the MS line high flow isolation AL in terms of pressure differential does not change for the TPO uprate. This is because the AL in terms of percent rated steam flow is decreased as a result of the higher absolute flow at TPO. Thus, the licensee stated no new instrumentation or changes are needed for this function. The licensee also stated that sufficient margin exists to the trip setpoint to allow for normal plant testing of the MSIVs. The licensee stated that the basis is consistent with TLTR Section F.4.2.5.

The NRC staff reviewed the TSAR Section 5.3.5 and the TLTR Section F.4.2.5 and determines that the licensees basis for no changes to the MS line high flow isolation instrumentation or setpoint is consistent with the TLTR since the MS line differential pressure is not changed for the TPO. Thus, the NRC staff concludes that the MS line high flow isolation will continue to perform its function at the TPO uprate conditions.

TSAR Section 5.3.6 - Fixed APRM Scram The licensee stated in TSAR Section 5.3.6 that the fixed APRM scram ALs in percent RTP do not change for the TPO uprate, and that the generic guidelines in TLTR Section F.4.2.2 are applicable. The limiting transient relying on the fixed APRM trip is the MS line isolation valve closure with indirect scram causing possible vessel overpressurization. This event was analyzed assuming 102 percent CLTP and is reanalyzed on a cycle-specific basis.

The NRC staff reviewed the TSAR Section 5.3.6 and the TLTR Section F.4.2.2 and determines that the licensees justification for no change to the fixed APRM scram is consistent with the TLTR. Further, the TLTR states that licensees will reanalyze the limiting transient relying on the fixed APRM trip during the first TPO reload to ensure the revised setpoint provides adequate protection. Therefore, the NRC staff concludes that the fixed APRM scram setpoint will continue to meet its function for the TPO uprate conditions.

TSAR Section 5.3.7 - APRM Simulated Thermal Power - Upscale Flow Biased Scram The proposed changes to the flow-referenced APRM allowable values and setpoints are summarized in Table 5-1 in TSAR Section 5.3. The licensee stated that the flow-referenced APRM AVs and setpoints for both two loop operation and single loop operation are unchanged in units of absolute core thermal power versus recirculation drive flow. Because the allowable values and setpoints are expressed in percent of RTP, they decrease in proportion to the power uprate, or CLTP/TPO. The proposed changes to the allowable values and nominal trip

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION setpoints for the Simulated Thermal Power - Upscale functions are based on the NRC approved approach described in TLTR Section F.4.2.1.

In Table 2.1 of Enclosure 1 of the LAR, Item 14, the licensee summarized the change of the single loop operation simulated thermal power upscale rod block nominal trip setpoint in percent of RTP from 0.57*(Wd-10.6) + 54.0 to 0.56*(Wd-10.8) + 53.1 where Wd is the percentage recirculation drive flow, where 100 percent drive flow is required at 100 percent core power and flow. The licensee provided the calculation details of the change of the recirculation offset in this formula from 10.6 to 10.8 in its supplement dated December 19, 2017 (Proprietary information. Not publicly available.).

The NRC staff determines that the licensees proposed changes to the flow-referenced APRM AVs and setpoints are acceptable since the licensee has followed the practice defined in TLTR thus, the licensee continues to meet the regulatory requirements of 10 CFR Part 50, Appendix K for the TPO uprate.

TSAR Section 5.3.8 - Rod Worth Minimizer Low Power Setpoint The licensee stated that the generic guidelines in TLTR Section F.4.2.9 are applicable to the HCGS TPO. The generic guidelines state that the low power setpoint at which rod patterns are enforced is usually expressed in terms of percent of rated power. The current value of this setpoint will be maintained in terms of the absolute power, and its value relative to the licensed power will be reduced.

The HCGS TPO rod worth minimizer lower power setpoint analytic limit has been scaled in terms of percent power to maintain the value in absolute power. The licensee summarized in the LAR Enclosure 1 Table 2.1 OL and TS Changes, Item 11, that the limiting condition for operation (LCO) 3.1.4.1 applicability is revised from 8.6 percent to 8.5 percent RTP as the minimum allowable low power setpoint. The licensee also listed in the TSAR Table 5-1 that the rod worth minimizer low power setpoint analytic limit (RWM LPSP - AL) in percent rated thermal power is changed from 8.576 to 8.441. The licensee confirmed in its supplement dated December 19, 2017 that these two values refer to the same parameter (i.e., RWM LPSP - AL).

In addition, the licensee clarified that the change in the RWM LPSP - AL can be calculated two ways:

(1)

Based on the published TPO increase of 1.6 percent RTP in the TSAR:

(8.576 / 1.016) = 8.441 (2)

Based on the ratio of the change in RTP from 3,840 MWt to 3,902 MWt:

(8.576) / (3,902/3,840) = 8.440 Both calculation results are rounded up for conservatism.

The larger of these two calculated analytic limits (i.e., 8.441 percent RTP) was chosen for the TPO for conservatism. Thus, the rod worth minimizer rod withdrawal blocks enforcement will be based on a conservative change in the AL for TPO.

The NRC staff reviewed the LAR including supplement and determines the proposed changes are acceptable since the licensee followed the practice defined in TLTR, thus will continue to meet the regulatory requirements of 10 CFR Part 50, Appendix K.

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION TSAR Section 5.3.9 - Rod Block Monitor As stated in UFSAR Section 7.7.2.1, the rod block monitor (RBM) prohibits erroneous withdrawal of a control rod during operation at high power levels. Although the RBM does not perform a safety-related function, in the interest of plant economics and availability it is designed to meet certain salient design principles of a safety system. In TSAR Section 5.3.9, the licensee stated that, ((

))

The NRC staff reviewed the licensees submittal and determines the licensee conclusion is acceptable regarding the RBM trip setpoints.

TSAR Section 5.3.10 - Flow-Biased Rod Block Monitor The licensee stated that HCGS does not have a flow-biased RBM system, thus, the NRC staff concludes that this section does not apply.

TSAR Section 5.3.11 - Main Steam Line High Radiation Isolation The licensee stated that the MSL high radiation isolation setpoint remains unchanged and this approach is consistent with the NRC approved guidelines defined in the TLTR Section F.4.2.8.

The NRC staff finds that the licensees conclusion is acceptable and the licensee has followed the practice defined in TLTR and has met the regulatory requirements of 10 CFR Part 50, Appendix K.

TSAR Section 5.3.12 - Low Steam Line Pressure MSIVC (RUN Mode)

The licensee stated that the AL for this function to initiate the main steamline isolation valve closure (MSIVC) on low steam line pressure when the reactor is in the RUN mode remains unchanged and this approach is consistent with the NRC approved guidelines defined in the TLTR Section F.4.2.7.

The NRC staff finds that the licensees conclusion is acceptable and the licensee has followed the practice defined in TLTR and has met the regulatory requirements of 10 CFR Part 50, Appendix K.

TSAR Section 5.3.13 - Reactor Water Level Instruments The licensee stated that the use of current ALs of the reactor water level instruments maintains acceptable safety system performance per the NRC approved guidelines defined in the TLTR Section F.4.2.10. The low reactor water level TS setpoints for scram and automatic depressurization system (ADS)/emergency core cooling system (ECCS) are not changed for the TPO uprate. The high water level ALs for trip of the main turbine and the FW pumps are not changed for the TPO uprate.

The NRC staff finds that the licensees conclusion is acceptable and the licensee has followed the practice defined and justified in the TLTR and has met the regulatory requirements of 10 CFR Part 50, Appendix K.

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION TSAR Section 5.3.14 - Main Steam Line Tunnel High Temperature Isolations The licensee stated that the high steam tunnel temperature AL remains unchanged for the TPO uprate per the discussion presented in Section 5.2. The NRC staff determined that the licensees conclusion is acceptable and the licensee has followed the practice defined in TLTR and has met the regulatory requirements of 10 CFR Part 50, Appendix K.

TSAR Section 5.3.15 - Low Condenser Vacuum (TSAR Section 5.3.15)

The licensee stated in TSAR Section 5.3.15 that the proposed TPO would slightly increase the amount of heat discharge to the main condenser, resulting in a slight increase in condenser backpressure. The licensee stated that the change is insignificant and would not impact any trip signals associated with a low condenser vacuum.

The NRC staff determines that the licensees conclusion is acceptable since the changes in the operating conditions of the main condenser are minimal (i.e., less than 0.15 inch of mercury absolute).

TSAR Section 5.3.16 - Turbine Stop Valve (TSV) Closure Scram, Turbine Closure Valve (TCV)

Fast Closure Scram, and End-of-Cycle Recirculation Pump Trip (EOC-RPT) Bypasses The TSV closure scram, TCV fast closure scram, and EOC-RPT bypass functions are used to disable scrams and recirculation pump trips at low power levels. The basis for these bypasses is to allow operational flexibility so that a scram may be avoided during turbine generator trips at low power. When the reactor pressure is sufficiently low, the TSV closure scram, TCV fast closure scram, and EOC-RPT signals are bypassed. The bypass setpoint for these scrams signals is expressed in percent of RTP. The scram is avoided by transferring steam to the turbine bypass system during a turbine generator trip at low power.

The licensee stated that the high pressure turbine is being modified for the TPO uprate to maintain adequate flow margin on the TCVs. This modification changes the turbine first stage power/pressure relationship. Because the turbine modifications were not assumed in the TLTR, the basis for following the TLTR approach was re-evaluated.

In the EPU analysis, the licensee reduced the AL on which the scram and EOC-RPT bypass setpoint was based from 30 to 25.7 percent of the RTP. Rescaling the 25.7 percent for the TPO results in an AL of 25.2 percent of the RTP. Another reduction by 1.2 percent conservatively reduces the AL to the pre-TPO level of 24 percent. The NRC staff determines that the AL of 24 percent at the TPO uprate power is acceptable because the TSV closure scram, TCV fast closure scram and EOC-RPT would be conservatively enforced at a lower value of the RTP.

Thus, the NRC staff concludes that the TSV closure scram, TCV fast closure scram, and EOC-RPT bypass functions will continue to perform their function at the TPO uprate conditions.

3.5.4 Thermal Power Measurement Uncertainty Background - LEFM Technology and Measurement Nuclear power plants are licensed to operate at a specified core thermal power. Appendix K, ECCS Evaluation Models, to 10 CFR Part 50 requires ECCS analyses to assume that the

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION reactor has been operating continuously at a power level at least 1.02 times the licensed thermal power level (i.e., 102 percent RTP) to allow for instrumentation uncertainties.

Alternatively, Appendix K allows such analyses to assume a value lower than 102 percent RTP (but not less than the licensed thermal power level), 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 power level assumed in the ECCS analysis and the licensed power level by using more accurate instrumentation to calculate the reactor thermal power.

Cameron LEFM 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.

Multiple diagonal acoustic paths are used in the LEFM, 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. The velocity profile correction factor is derived from calibration testing of the LEFM in a plant-specific piping model at a calibration laboratory.

The licensee states that the LEFM system will replace the currently installed CE Nuclear Power Cross Flow Ultrasonic Flow Meter and resistance temperature detector indication. The currently installed FW flow venturis will be used if the LEFM system is not functional. The licensee further states:

The LEFM system consists of a single measurement spool piece meter to be installed in the 30-inch common feedwater header, two transmitter signal processing units and two redundant central processing units. The measurement spool piece contains 16 ultrasonic, multi-path, transit time transducers grouped into the two planes of eight transducers each, two 4-wire RTDs, and two pressure transmitters.

The LEFM system is used to provide FW flow input for the plant thermal heat balance calculation. The Cameron LEFM flow meters have two operating modes (Normal and Maintenance) and a Fail mode.

Normal Mode: measures the average flow of two independent LEFM subsystems, where each subsystem consists of four acoustic paths that are summed into the eight paths that comprise the LEFM system. The normal mode is displayed when the FW flow, temperature and header pressure signals are normal and operating within design limits. Power level uncertainty in this condition is 0.34 percent.

Maintenance Mode: the state when any LEFM system has only one of two LEFM subsystems fully operational, which results in flow computation based on the one fully operational LEFM. An alert alarm indicates a loss of system redundancy, and the power level uncertainty in this condition is 0.66 percent. Power is to be reduced to less than or equal to 3889 MWt (CLTP) within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> if LEFM functionality cannot be restored to the normal mode.

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION Fail Mode: indicates a loss of function and causes a fail alarm. Power is to be reduced to 3840 MWt within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> if LEFM functionality cannot be restored to either normal or maintenance modes.

The LEFM system performs continuous self-checking of the transducer signals and the calculation results. The testing verifies that the digital circuits are operating correctly and within the specified accuracy. This allows the system to identify failure conditions that will cause the LEFM to switch modes. Validated LEFM data including calculated results, status, and signal process information is sent to the plant computer at regular intervals.

The following sections provide the NRC staffs review of the licensees justification regarding the thermal power measurement uncertainty for the proposed power uprate for HCGS.

5 - Leading Edge Flowmeter (LEFM) Installation Location Drawings LAR Enclosure 15 provides the system isometric drawing from the reactor feed pump to the drywell and shows the proposed installed location of the LEFM, however, the licensee did not provide a reference to a standard to which the LEFM will be installed.

By letter dated December 22, 2017, the licensee supplemented the LAR providing a discussion of the installation recommendations given in ASME Performance Test Code (PTC) 19.5 Flow Measurement, Section 10-9.1, Acoustic Path Length and Angle, 10-9.4, Secondary Flow and Distorted Velocity Profiles, and 10-9.5, Integration. In its response, the licensee stated that the LEFM is designed to meet the recommendations of the ASME PTC 19.5-2004.

The staff reviewed the licensee response and determined it is acceptable because of the following: (1) as recommended in ASME PTC Section 10-9.1, the licensee considered the effect of temperature and pressure changes, and thermal expansion on the as-built dimensions of the LEFM and included in the uncertainty analysis, (2) as recommended in Section 10-9.4, the LEFM body used by HCGS has a crossed-plane configuration required for a high accuracy measurement, (3) as recommended in Section 10-9.5, the flow profile changes are addressed through calibration and continuous monitoring of hydraulic parameters in the LEFM electronics, (4) the LEFM electronics continuously monitor several hydraulic parameters, including the flatness of the velocity profile, and provide alarms to indicate a change in flow profile that falls outside the bounding value for the meter factor used in the site-specific uncertainty analysis, (5) possible dimensional changes are addressed through plant maintenance procedures, (6) the licensee will perform periodic monitoring of the pipe wall thickness to confirm that meter dimensions remain within the bounding values of the site-specific uncertainty analysis, and (7) the LEFM is calibrated at a National Institute of Standards and Technology (NIST)-traceable laboratory in a hydraulic model representative of the plant piping in which it will be installed.

Feedwater Flow Measurement Technique and Power Measurement Uncertainty Items A through C Items A, B, and C in Section I of Attachment 1 to RIS 2002-03 guide licensees in identifying the approved topical reports, providing references to the NRCs approval of the measurement technique, and discussing the plant-specific implementation of the guidelines in the topical report and the NRC staffs approval of the FW flow measurement technique, respectively. In its

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION LAR, the licensee identified topical reports ER-80P and ER-157P, Revision 8, as applicable to the Cameron International LEFM CheckPlus' system. The licensee also referenced NRC SEs dated March 8, 1999, for ER-80P, and August 16, 2010, for ER-157P. In its discussion of item C, the licensee stated that the LEFM CheckPlus' system will be installed according to the requirements specified in ER-80P and ER-157P. Based on its review of the licensees submittals as stated above, the NRC staff finds that the licensee has sufficiently addressed the plant-specific implementation of the Cameron LEFM CheckPlus' system using proper topical report guidelines. Therefore, the licensees description of the FW flow measurement technique and implementation of the power uprate using this technique follows the guidance in Items A through C of Section I of Attachment 1 to RlS 2002-03 and, therefore, has met the regulatory requirements of 10 CFR Part 50, Appendix K.

Item D Item D in Section I of Attachment 1 to RIS 2002-03 guides licensees in addressing four criteria provided in the NRC staffs SEs for ER-80P and ER-157P when implementing the FW flow measurement uncertainty technique. The staffs SEs both include four plant-specific criteria to be addressed by a licensee when using the topical reports for power uprate. The licensees submittals address each of the four criteria in its LAR Enclosure I Section 3.2.4 Criterion 1 through Criterion 4.

The NRC staff reviewed the licensee submittal including the discussion of the four criteria to be addressed in ER-80P and ER-157P. Based on its review, the NRC staff concludes that the licensee adequately addressed Criteria 1, 3 and 4. As described in the topical reports, Criterion 2 applies only to licensees that have already installed the LEFM. The licensee states in its LAR that Criterion 2 is not applicable to HCGS as the LEFM is not currently installed at HCGS; therefore, the NRC staff concludes this criterion is not applicable.

Item E Item E in Section I of Attachment 1 to RIS 2002-03 guides licensees in 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 the LEFM and core thermal power measurement uncertainty and methodology in its LAR Enclosure I Section 3.2.3. In the LAR,, Cameron Document ER-1123P, Bounding Uncertainty Analysis for Thermal Power Determination at Hope Creek Unit 1 Nuclear Generating Station Using the LEFM CheckPlus' system Revision 2 (Proprietary), the licensee provided an analysis of the LEFM CheckPlus' system uncertainty contributions when operating in the normal mode and maintenance mode to the overall calculated thermal power uncertainty. The NRC staff reviewed these reports and determined that 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 licensees approach used in the setpoint methodology involved statistically combining inputs to determine the overall uncertainty. Channel statistical allowances were calculated for the instrument channels, dependent parameters were arithmetically combined to form statistically independent groups, and then these were combined using the SRSS approach to determine the overall uncertainty. This methodology is consistent with the vendors

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION determination of the uncertainty of the Cameron LEFM CheckPlus' system, as described in the referenced topical reports, and is consistent with the guidelines in RG 1.105, Revision 3.

As a result, the NRC staff determined 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 NRC staff concludes that the licensee has adequately addressed the guidance in Item E of Section I of to RIS 2002-03 and, therefore, has met the regulatory requirements of 10 CFR Part 50, Appendix K.

Item F Item F in Section I of Attachment 1 to RIS 2002-03 guides licensees in providing information to address the specified five aspects of the calibration and maintenance procedures related to all instruments that affect the power calorimetric:

i.

maintaining calibration ii.

controlling software and hardware configuration iii.

performing corrective actions iv.

reporting deficiencies to the manufacturer

v.

receiving and addressing manufacturer deficiency reports In the LAR, Enclosure 1, Section 3.2.4 Disposition of NRC Criteria for Use of LEFM Topical Reports, the licensee provided its responses to these aspects in Response to Criterion 1, in, Section 3.2.5 Deficiencies and Corrective Actions, and in Enclosure 5 Summary of Regulatory Commitments. The NRC staff reviewed the licensees responses to each of the five aspects of the calibration and maintenance procedures listed and determined that the licensee has adequately responded to the criteria. Therefore, the NRC staff concludes that the licensee has followed the guidance in Items G and H of Section I of Attachment 1 to RIS 2002-03 and; therefore, has met the regulatory requirements of 10 CFR Part 50, Appendix K.

Items G and H Items G and H in Section I of Attachment 1 to RIS 2002-03 guide licensees to provide a proposed TS allowed outage time (AOT) for the instrument and to propose actions to reduce power if the AOT is exceeded.

The licensee proposed a 72-hour AOT for operation at any power level above the current licensed power of 3840 MWt with the LEFM not fully functional. The licensee justified the 72-hour AOT in its LAR in Enclosure 1, Section 3.2.4, Criterion 1. In addition, the licensee provided a regulatory commitment in the LAR, Enclosure 5, Item 1, which provides procedural guidance to the operators regarding the required actions when the LEFM system is not in the normal mode. The NRC staff reviewed the LAR including the Cameron engineering reports and the licensees commitment, Item 1, and determines that the licensee has provided adequate responses to items G and H.

Based on the above, the NRC staff concludes that the licensee provided sufficient justifications for the proposed AOT and the proposed power reduction actions if the AOT is exceeded.

Therefore, the NRC staff concludes that the licensee has followed the guidance in Items G and

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION H of Section I of Attachment 1 to RIS 2002-03 and, therefore, has met the regulatory requirements of 10 CFR Part 50, Appendix K.

Technical Specifications, Protection System and Emergency System SettingsSection VIII of Attachment 1 to RIS 2002-03 guides licensees in providing information to address the changes to the plants TSs, protection system settings, and emergency system settings needed to support the power uprate.

Items A through C Items A, B, and C in Section VIII Changes to technical specifications, protection system settings, and emergency system settings of Attachment 1 to RIS 2002-03 guide licensees in providing a description of the change, identification of analyses affected by or supporting the change, and the justification for the change for any analyses that support or are affected by the change. The NRC staffs evaluation of the identified instrumentation for the new power level is based on the ALs documented by the licensee in the submitted application.

The licensee summarized the proposed changes to the RFOL and TS in its LAR Enclosure 1 Section 2.1 OL and TS Changes with the associated marked-up pages included in. The licensee described the changes to the instrumentation and control systems in TSAR Enclosure 6, Section 5.3.16 TSV Closure Scram, TCV Fast Closure Scram, and EOC-RPT Bypasses. The licensee stated in its TSAR Section 10.9, that the HCGS emergency operating procedures (EOPs) action thresholds are plant specific and will be addressed using standard procedure updating processes. The licensee claims that the TPO uprate will have a negligible effect or no effect on the operator action thresholds and on the EOPs in general.

The NRC staff reviewed the licensees responses to Items A, B, and C of RIS 2002-03 and determines that the licensee has adequately addressed these items. The NRC staff concludes that the licensee provided sufficient justifications for the proposed TS changes, that the licensee has followed the guidance in RIS 2002-03, and has, therefore, met the regulatory requirements of 10 CFR Part 50, Appendix K.

3.6 TSAR Section 6.0 - Electrical Power and Auxiliary Systems The electrical equipment design information is provided in LAR, Enclosure 1, Section 3.0 Description and Evaluation of the Proposed Change, and Sections 6, 9 and 10 of the TSAR.

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

On-Site Alternating Current (AC) Power Systems Off-Site Power Block Equipment (Generator, Transformers/Switchyard Equipment, and Isolated-Phase Bus Duct)

Direct Current (DC) Power System Emergency Diesel Generators (EDGs)

Grid Stability Station Blackout (SBO)

Equipment Qualification (EQ) Program

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION The following provides the NRC staffs technical review of the topics in Section 6.0 of the TSAR.

3.6.1 TSAR Section 6.1 - Alternating Current (AC) Power According to the UFSAR Section 8.3, Onsite Power Systems, the onsite AC power systems include a Class 1E system and a non-Class 1E system. The non-Class 1E system supplies AC power to non-Class 1E loads. The Class 1E power system supplies all Class 1E loads, which are needed for safe and orderly shutdown of the reactor, maintaining the plant in a safe shutdown condition, and mitigating the consequences of an accident. In addition to Class 1E loads, the Class 1E system supplies power, through isolation devices, to a limited number of non-Class 1E loads that are important to the integrity of the power generating equipment.

Isolation between Class 1E power supply buses and the non-Class 1E loads is achieved by tripping the Class 1E breaker under LOCA conditions. The on-site Class 1E AC power system distributes power at 4.16 kilo-Volts (kV), 480 V, and 208/120 V.

The Class 1E power system is divided into four independent channels. Each channel supplies power to loads in its own load group. Each Class 1E 4.16 kV bus is provided with connections to the two off-site power sources. One of these sources is designated as the normal source and the other as the alternate source for the bus. In addition to these two connections to the off-site power, each of the 4.16-kV Class 1E buses is connected to its dedicated standby diesel generator (SDG). These SDGs serve as the standby electric power source for their respective channels in case both the normal and alternate power supplies to a bus are lost.

The on-site non-Class 1E AC system distributes power at 7.2 kV, 4.16 kV, 480 V, and 208/120 V. The 7.2-kV system supplies power to large auxiliary loads, such as the motor generator sets for the reactor recirculation pumps, condensate pumps, station air compressors and water chillers. The 4.16-kV system feeds large auxiliary motor loads and single-ended and double-ended unit substations. A tie circuit breaker is provided for each of the double-ended unit substations. Interlocks are provided so that the tie breaker can be closed only if one of the infeed breakers of the double-ended unit substation is open. The 480 V unit substations feed 480 V motor control centers (MCC), motors of 100 to 250 horsepower rating, and 480 V power panels. The MCCs supply power to motors of up to 75 horsepower rating, battery chargers, 480/277 V power distribution panels, and 480 and 208/120 V power distribution panels.

Uninterruptible power supply (UPS) panels of 120 V AC supply the security system, public address system, process computer, and balance of plant (BOP) computer. The 120 V AC UPS are provided to the select control cabinets considered critical for stable power operation associated with FW heaters, service air, safety auxiliary cooling, reactor recirculation system motor generator sets and off-gas systems. The distribution panels feed miscellaneous loads such as lighting, space heaters, and unit heaters.

TSAR Section 6.1.1 - Offsite Power (Power Block Equipment)

In Section 6.1.1 Off-Site Power of the TSAR, the licensee described the off-site power block equipment design and its review of the impact of the TPO uprate on the equipment below.

Main Generator: The generator is a direct-driven 3-phase 60 Hz, 25,000 V, 1,800 rpm, hydrogen inner-cooled, synchronous generator rated for 1,373 megavolt amps (MVA) at a 0.94 power factor (PF), with a 0.50 short circuit ratio at a nominal hydrogen pressure of 75 psig.

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION The licensee further stated that the main generator will be operated within the existing generating capability curve for TPO uprate. The gross generator megawatts electric (MWe) output is on or within the existing generator reactive capability curve.

Main Transformers: The 1,400 MVA main power transformer system consists of three single-phase, 466.7 MVA 24 - GND Y / 288.7 kilovolt (kV), forced oil and air (FOA), 65°C rise, 60 Hz, oil-filled type, outdoor transformers.

The licensee further stated that the main transformers and the associated switchyard components (rated for maximum generator output) are adequate for the TPO uprate-related transformer output. The items with the least margin are the disconnect switches that have 25.6 percent margin.

Isolated Phase Bus Duct: The isolated phase bus duct consists of a main bus and a delta bus. The isolated phase bus continuous current rating is based on a 105°C operating temperature (65°C rise above a 40°C ambient temperature) with forced air cooling for the main bus and the delta bus. The main bus is rated at 34,000 A with a momentary fault current rating of 468,000 A. The delta bus is rated at 19,500 A with a momentary fault current rating of 468,000 A. The voltage rating of the system is 25,000 V. The forced cooling is handled by an air handling unit with a design heat transfer capacity of 682,000 Btu/hr.

The licensee stated that the isolated phase bus duct is adequate for both rated voltage and low voltage current output. The isolated phase bus duct cooling system capacity is adequate for the expected heat rejection loads during the TPO uprate operation. Therefore, the isolated phase bus duct cooling system is adequate to support the TPO uprate.

In Section 6.1.1, Enclosure 8, of the LAR, the licensee stated that the operation of HCGS at the TPO level will not require modifications to the generator, bus duct, main transformer and the transmission components (disconnect switches, tubular bus and transmission lead) leading to the 500 kV switchyard to support operation at the nameplate output capacity.

The NRC staff reviewed the LAR and UFSAR and found that the main generator, main transformers and associated switchyard components, and isolated phase bus duct remains adequately sized for the application conditions because these components were evaluated at EPU conditions and found to be adequate. Based on this, the NRC finds that the licensee adequately addressed the impact of the MUR power uprate conditions on the power block equipment, and that the power block equipment will have adequate capacity in accordance with GDC 17 and within its design to support implementation of the MUR power uprate.

TSAR Section 6.1.1 - Offsite Power (Grid Studies)

In Section 3.4.5 Grid Stability Studies of LAR Enclosure 1, the licensee stated:

Grid stability studies were performed for HCGS operation at a bounding electrical power output of 1320 MWe. These results bound operation at the proposed MUR power level of 3902 MWt.

The PJM [regional transmission organization] studies were performed using generator operating curves defined in the Artificial Island Operating Guide

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION (AIOG) A-5-500-EEE-1686. These curves are not modified for operation at MUR power levels. Since HCGS will continue to operate within the existing generator curves, the existing PJM studies are bounding.

Grid stability is a function of the overall grid configuration with all the lines and equipment connected, and the balance of the generation compared to the grid loading. The HCGS contribution to grid stability is determined by the generator electrical output and the turbine, generator and main transformer characteristics which are all fixed by the equipment design.

Hope Creek is operated in close proximity with the PSEG Nuclear Salem Units 1 and 2 generating stations. HCGS has been analyzed for stability for the following transients, provided the station is operated per the AIOG:

Loss of the HCGS Generator Loss of the most critical generating unit on the grid Loss of the most critical transmission line Electrical component ratings and design parameters are kept up to date in the AIOG to assure system stability. Sufficient margin exists for operation at 3902 MWt since all the equipment will remain within its nameplate rating. HCGS has determined that the MUR power uprate to 3,902 MWt will have no significant effect on grid stability or reliability and no modifications to the transmission system are required.

In its supplement dated December 14, 2017, the licensee further stated that the existing grid stability study determined operating curves for the HCGS generator under multiple postulated grid conditions. This study provided operating curves with a maximum MWe output of 1,320 MWe under normal conditions. Existing operating procedures direct control room operators to maintain electrical output below the operating curves. The relationship between core thermal power and generator MWe output is dependent on multiple variables, primarily circulating water inlet temperature. At the current licensed thermal power (CLTP) of 3840 MWt, generator output historically varied between approximately 1220 MWe and 1306 MWe. Under TPO power uprate conditions, approximately 20 additional MWe are expected. If plant conditions allow generator output to approach 1,320 MWe, operators will decrease reactor power to maintain the generator output below the allowed operating curves.

The NRC staff reviewed the LAR and the supplement dated December 14, 2017, and determines that the licensee adequately addressed the impact of the TPO uprate conditions on grid stability because the plant operator will maintain the generator output below the allowed operating curves specified in the grid stability study. Based on the review above, the NRC staff determines that the TPO uprate would not adversely impact the stability of the transmission/grid system.

TSAR Section 6.1.2 - On-Site Power Section 6.1.2 On-Site Power of the TSAR stated that the on-site distribution system loads were reviewed under normal and emergency operating scenarios. In both cases, the loads were computed based primarily on equipment nameplate data or brake horsepower (BHP).

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION TSAR Section 6.1.3 - Emergency Diesel Generator According to Section 8.1 Introduction of the UFSAR, in the event of loss of offsite power, four independent SDGs provide the standby power for Class 1E loads and selected non-Class 1E loads important to the power generating equipment integrity.

In Section 6.1.3 Emergency Diesel Generator of the TSAR, the licensee stated that there are no modifications associated with the TPO uprate that would increase the electrical loads associated with the engineered safeguard and selected non-safeguard systems or alter the diesel generator subsystems. Therefore, the performance of the EDG and the 4kV emergency system is not affected by the TPO uprate.

The NRC staff reviewed the LAR and the UFSAR and found that there is no need for changes to the EDGs associated with the TPO uprate since there are no load changes. Therefore, the NRC staff determines that the MUR power uprate conditions will not impact the EDGs and that GDC 17 requirements will continue to be met.

3.6.2 TSAR Section 6.2 - Direct Current (DC) Power According to Section 8.3 of the UFSAR, the DC power systems consist of the non-Class 1E and Class 1E systems. The non-Class 1E DC system distributes power at +/-24, 125, and 250 V.

Each of these systems is supplied from independent batteries and battery chargers. The power for the battery chargers are supplied from non-Class 1E, SDG-backed, MCCs. The non-Class 1E MCCs are supplied from Class 1E unit substations, however, the unit substation feeder breakers to these MCCs are automatically tripped on a LOCA. The non-Class 1E DC systems supply power to non-Class 1E loads such as turbine-generator emergency seal oil pump, emergency bearing lube oil pumps, control for non-Class 1E switchgear, and neutron monitoring system.

The Class 1E DC system distributes power at 125 V and 250 V. The Class 1E 125 V DC system is divided into four independent channel systems. Power for each of these systems is supplied from batteries and battery chargers of the corresponding load group channel. There are two 250 V DC systems. One is associated with the high pressure coolant injection (HPCI) system and the other with the RCIC System.

In Section 6.2, DC Power, of the TSAR, the licensee stated that the DC loading requirements documented in the UFSAR and station load calculations were reviewed, and no reactor power-dependent loads were identified. The DC power distribution system provides control and motive power for various systems and components. These loads were used as inputs for the computation of load, voltage drop, and short circuit current values. Operation at the TPO RTP-level does not increase any loads or revise control logic. Therefore, there are no changes to the load, voltage drop, or short circuit current values.

The licensee further stated that the changes to the auxiliary power system as a result of the TPO uprate are small increases in the horsepower of the condensate pump and the RRC pump motors. The DC system does not power the affected pumps; therefore, the DC system is not affected by the increase in motor duty. There are no changes to the DC system loading resulting from TPO other than loads associated with the LEFM system. The licensee further stated that the effect of the DC load change imposed by the LEFM modification has been evaluated using the methodology documented in the existing electrical design analysis

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION calculations and has been found to be within the current acceptance criteria.

The NRC staff reviewed the LAR and the UFSAR and determines that the DC load changes as a result of the MUR power uprate conditions are small and within the existing DC power system capacity. Therefore, the staff concludes that the licensee addressed the impact of the MUR power uprate conditions on the DC power system, and that the DC power system has adequate capacity to operate the plant equipment in accordance with GDC 17 and within its design to support implementation of the MUR power uprate.

3.6.3 TSAR Section 6.3 - Fuel Pool The description of the spent fuel pool (SFP) cooling system provided in TSAR Section 6.3 indicated that the licensee maintains the heat load within the capability of the fuel pool cooling (FPC) system through cycle-specific calculations to verify the SFP heat load would be maintained within the previously analyzed value. The TPO uprate does not affect the heat removal capability of the FPC system used for normal refueling or the FPC system supplemented with the residual heat removal system (RHR) fuel pool cooling assist mode for full-core discharges. The TSAR stated that the TPO heat load would remain within the design basis heat load for the FPC system supplemented with RHR assist mode. Thus, the decay heat load resulting from operation at TPO uprate conditions is acceptable and can be maintained within the capacity of the FPC system for normal discharges and the FPC system supplemented by the RHR SFP cooling assist mode for full core discharges.

3.6.4 TSAR Section 6.4 - Water Systems The HCGS cooling water systems include station service water system (SSWS), safety auxiliaries cooling system (SACS), reactor auxiliaries cooling system (RACS), turbine auxiliaries cooling system (TACS), and circulating water (CW). Of these systems, the SSWS and SACS are classified as safety-related. The SSWS provides a reliable supply of service water to the SACS during normal operation, normal shutdown, and following design-basis accidents. The SACS cools the following major equipment:

RHR heat exchangers RHR pump seal and motor coolers RHR room coolers Diesel generator coolers Diesel generator room cooling HPCI, RCIC, and CS pump room coolers Filtration recirculation and ventilation system coolers Control room and equipment chillers SFP heat exchangers The ultimate heat sink (UHS) for HCGS is the Delaware River. The SSWS provides water from the UHS to the SACS heat exchangers, which support safety-related equipment cooling by the SACS. The ability of the UHS, SSWS, and SACS to provide adequate cooling water for its safety functions had been confirmed through previous analyses based on 102 percent of CLTP, which remains bounding for TPO operation. Therefore, the UHS, SSWS, and SACS remain acceptable for TPO operation.

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION The non-safety-related cooling systems could experience an increase in water temperature due to power-related increases in heat rejection to the RACS, TACS, and CW system. The RACS is cooled by the SSWS during normal operation, normal shutdown, and during a loss of offsite power. The heat loads served by the RACS are minimally affected by operation at TPO and remain within the design heat load for the system. The TACS loops are served from the SACS heat exchangers and isolate from the SACS following design basis events or indications of a loss of system integrity. The TACS would experience a heat load increase from TPO operation due to an increase in main generator losses rejected to the stator water and hydrogen coolers and other loads. The TSAR noted that the TACS has flow margin such that increased TACS flow would maintain the TACS system temperatures. The CW system removes the heat rejected to the main condenser, and the licensee confirmed that the CW system would remain adequate for TPO operation. Therefore, the plant-specific analyses described in TSAR Section 6.4 indicates that the increase in heat loads resulting from TPO operation is within the available margin for the RACS, TACS, and CW systems.

3.6.5 TSAR Section 6.5 - Standby Liquid Control System The SLC system is an independent backup system for the control rod drive system but not intended for prompt reactor shutdown. The SLC system is capable of orderly and safe shutdown of the reactor from a full power condition, and maintaining it subcritical until the cold shutdown condition is achieved, without control rod movement. The system pumps a sodium pentaborate solution into the reactor vessel to achieve a sub-critical condition.

The generic evaluation of the SLC system is given in TLTR Section 5.6.5 and Appendix L.3.

The TPO uprate does not affect shutdown or injection capability of the SLC system. The shutdown margin is the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming all full-length rod cluster assemblies (shutdown and control) are fully inserted, except for a single rod cluster assembly assumed to have the highest reactivity worth that is assumed to be fully withdrawn.

Because of its dependency on reload, the shutdown margin and the required reactor boron concentration are confirmed for each reload core. The system relief valve margin is adequate for the TPO uprate because the system was confirmed to have a minimum relief valve margin of greater than or equal to 141 psi prior to the TPO uprate. This margin was established to be 141 psi for a power level of 3,952 MWt, and therefore, greater than or equal to 141 psi will be adequate for the TPO uprate to ensure that the relief valves do not lift during system injection.

In the event of an ATWS, injection of the sodium pentaborate solution can be initiated manually by the operator or automatically by the redundant reactivity control system.

The NRC staff determines that the licensees generic evaluation presented in TLTR Section 5.6.5 (SLC system) and Appendix L.3 (ATWS Evaluation) is applicable to the HCGS TPO uprate. Therefore, the NRC staff concludes that there is reasonable assurance that the SLCS is capable for ATWS mitigation and the SLCS will continue to meet the requirements of GDC 27 under the TPO uprate conditions.

3.6.6 TSAR Section 6.6 - Power-Dependent Heating, Ventilation, and Air Conditioning Section 6.6 of the TSAR presented an evaluation of heating, ventilation and air conditioning (HVAC) systems that are potentially affected by the TPO uprate, which consist of heating, cooling supply, exhaust, and recirculation units in the turbine building, reactor building, steam

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION tunnel and primary containment (drywell). The evaluation indicated that the TPO uprate would result in a minor increase in the main steam tunnel heat load caused by the slightly higher FW temperature. Other areas experience a negligible increase or no change in heat load.

Therefore, the power-dependent HVAC systems are adequate to support the TPO uprate.

3.6.7 TSAR Section 6.7 - Fire Protection Section 9.5.1 Fire Protection Program of the UFSAR describes the approved fire protection program at HCGS and explain how it complies with the requirements of 10 CFR 50.48, Fire protection, and the guidelines of Branch Technical Position (BTP) Chemical Engineering Branch 9.5-1 (ADAMS Accession No. ML070660454) and Appendix A to BTP Auxiliary and Power Conversion Systems Branch (APCSB) 9.5-1 (ADAMS Accession No. ML15322A269).

The UFSAR is cited in the HCGS RFOL license condition 2.C.7.

The NRC staff reviewed TSAR Section 6.7, Fire Protection, the licensees commitment to 10 CFR 50.48, Fire protection (i.e., the approved fire protection program), and the impact of the LAR on the results of the safe-shutdown fire analysis as noted in RIS 2002-03,, Sections II and III. The review focused on the effects of the LAR on the post-fire safe-shutdown capability and increase in decay heat generation following plant trips.

In Section 6.7 of the TSAR, the licensee stated that the operation of the plant at the TPO level does not affect the fire suppression or detection systems. There is no change in the physical plant configuration and the potential for minor changes to combustible loading as a result of the TPO uprate will be addressed by controlled design change procedures. In a supplement to the LAR dated November 1, 2017 (ADAMS Accession No. ML17305B270), the licensee provided all changes to the combustible loading, even those considered minor, and stated the impact of these changes on the plants compliance with the fire protection program. The licensee stated that PSEG may make changes to the approved fire protection program without prior approval of the Commission only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire. The licensee further stated that, combustible load changes are evaluated for impact on the ability to achieve and maintain safe shutdown in the event of a fire under the approved fire protection programs administrative controls, and those changes that do not have an adverse effect are implemented under license condition 2.C.7.

The licensee indicated that these administrative controls comply with the license condition and the MUR power uprate does not alter any aspects of the approved fire protection programs administrative controls.

In addition, the licensee indicated that the MUR power uprate modification adds 27 pounds of combustible material to Room 3449 in the Aux Radwaste building, which increases the loading of the room from 36,210 to 36,367 British thermal units per square feet. This increase in loading is less than a 0.5 percent increase and this room does not contain safe shutdown equipment or cables. This increase in combustible loading is made in accordance with the approved fire protection program and has been determined not to have an adverse impact on the ability to achieve and maintain safe shutdown in the event of a fire.

The NRC staff concludes that the change in combustible loading is acceptable because the increase is small and in a room that contains no safe shutdown equipment or cables such that it has no adverse impact on the fire protection system or program and does not affect safe-shutdown of the plant.

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION In TSAR Section 6.7, the licensee stated that the operator manual actions that are used for compliance with 10 CFR 50, Appendix R, were reviewed. No new operator actions have been identified in areas where environmental conditions, such as heat, would challenge the operator.

The licensee indicated that because this uprate is to be performed at a constant P-T, the normal temperature environments are not affected by TPO. Therefore, the operator manual actions required to mitigate the consequences of a fire are not affected. In the LAR the licensee conducted a review of the fire protection program related to administrative controls, fire barriers, fire protection responsibilities of plant personnel, and resources necessary for systems required to achieve and maintain safe-shutdown. The review focused on the effect of the TPO uprate and how it would affect these areas. The licensee concluded that the TPO uprate will not have an adverse effect on fire protection administrative controls, fire barriers, fire protection responsibilities of plant personnel, and resources necessary for systems required to achieve and maintain safe-shutdown.

The NRC staff reviewed the licensees statements in the LAR related to the impact of the MUR power uprate on the plant safe-shutdown and impacts due to increase in decay heat. For the MUR power uprate, the licensee reviewed its systems to achieve and maintain the plant in cold shutdown condition. The Appendix R analysis demonstrated that the plant can reach cold shutdown with significant margin to the 72-hours requirements in 10 CFR 50, Appendix R, Sections III.G.1.b and III.L. The TPO and the additional decay heat removal would not affect the ability to reach and maintain cold shutdown within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> without any time critical repair.

The NRC staff concludes that HCGS will continue to meet the 72-hour requirements in both 10 CFR Part 50, Appendix R, Sections III.G.1.b and III.L at TPO uprate conditions. Further, the NRC staff concludes that the changes proposed in the LAR will not adversely impact post-fire safe-shutdown capability.

In TSAR Section 6.7.1, 10 CFR 50 Appendix R Fire Event, the licensee stated that the TLTR Section L.4 presents a generic evaluation of Appendix R events for an increase of 1.5 percent of CLTP. The results of the Appendix R evaluation demonstrate that the fuel cladding integrity, reactor vessel integrity and containment integrity are maintained.

The NRC staff requested PSEG to provide additional information on whether it credits aspects of its fire protection system for other activities for the increased power operating conditions and total decay heat (e.g., by using the fire water pumps and water supply as backup cooling or inventory for non-primary reactor systems). Further, the NRC staff requested that the licensee discuss how any non-fire suppression use of fire protection water impacts the ability to meet the fire protection system design demands per NUREG-1048, Section 9.5.1.5, Fire Detection and Suppression.

In a supplement dated November 1, 2017, the licensee stated that, Section 9.5.1.2.3.1 of the UFSAR addresses fire water storage. The UFSAR states:

Fire protection water supply is from two, 350,000 gallon nominal capacity, fire water storage tanks located north of the plant. Each tank feeds the [Fire Protection System] and the demineralized water and boiler makeup systems. Of the 350,000 gallons of storage capacity for each tank, 328,000 gallons is dedicated to the Fire Protection Water System, and the remaining amount is available for the demineralized water system. The demineralized and boiler makeup water systems are fed through an external tap physically located above the fire water level of 328,000 gallons.

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION In addition, the licensee stated that no additional non-fire protection uses of fire water are credited in the design or licensing basis events. Fire water may be utilized in beyond design basis events or severe accident management for an alternate source of water for reactor pressure vessel (RPV) makeup, service water, emergency diesel generator cooling, condensate storage tank makeup, SFP makeup, safety auxiliaries cooling system makeup, hotwell makeup, or containment flooding. Procedural guidance is provided for makeup to the fire water ring header from either the cooling tower basin or the Delaware River during these scenarios.

Further, the licensee indicated that the MUR power uprate does not affect demineralized or boiler water makeup demand, fire water demand, or the amount of fire water storage reserved for fire suppression use. No additional non-fire suppression uses for fire protection water are added as part of the MUR power up rate. Use of fire water for beyond design basis events or severe accident management is not affected by the MUR uprate.

The NRC staff reviewed the licensees response and concludes that it satisfactorily addresses the request for information because the licensee identified several other features of the fire protection system used for other activities. The licensee indicated that the fire water is utilized to backup the water supply makeup to RPV, service water, emergency diesel generator cooling, condensate storage tank makeup, SFP makeup, safety auxiliaries cooling system makeup, hotwell makeup, or containment flooding. The licensee indicated that these scenarios are beyond-design-basis events crediting the fire protection system and are unaffected by the MUR power uprate. The staff finds the licensee information acceptable because the licensees analysis concluded that non-fire protection aspects of the fire protection system for other than fire protection activities would not be impacted nor would they impact fire protection system design demands by the proposed MUR power uprate.

The NRC staff has reviewed the information provided in the LAR regarding the fire suppression, detection systems, physical plant configuration, resulting combustible loading, and operator manual actions for compliance with 10 CFR 50, Appendix R. As stated above, the LAR involves no changes to the fire protection program that may adversely impact the post-fire safe-shutdown capability in accordance with 10 CFR 50, Appendix R.

The licensee has not made significant changes to the plant configuration as a result of the MUR power uprate. The licensee has shown no adverse effects would be created by the MUR power uprate on the fire detection or suppression systems, fire protection administrative controls, fire barriers, fire protection responsibilities of plant personnel, and resources necessary for systems required to achieve and maintain safe-shutdown.

Based on its review, the NRC staff has concluded that the LAR will not have an adverse impact on the fire protection program or post-fire safe-shutdown capability and; therefore, the NRC staff finds the LAR acceptable with respect to fire protection.

3.6.8 TSAR Section 6.8 - Systems Not Affected by TPO Uprate Based on previous NRC reviews, all systems that are significantly affected by TPO have been addressed in the TSAR.

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION 3.7 TSAR Section 7.0 - Power Conversion Systems The licensee provided an evaluation of TPO uprate effects on non-safety related power conversion systems in TSAR Section 7.0, including the turbine generator, turbine steam bypass system, and the FW and condensate system. In the SE under Section 3.5.2 it addressed the control systems associated with these mechanical systems. The turbine generator is important to safety with respect to turbine missile generation probability, but that analysis is not affected by the TPO uprate because the turbine speed remains constant and the turbine operating conditions do not significantly change. Turbine bypass capability is considered in the analyses for generator load rejection and turbine trip transients and is unchanged by the TPO uprate.

Furthermore, the limiting transient condition assumes the failure of the turbine bypass. Also, the effect of the TPO uprate on the turbine bypass capacity measured as a fraction of total steam flow is insignificant, so the turbine bypass capacity remains acceptable. The FW and condensate systems have adequate margin to accommodate the minor flow increase associated with the TPO uprate. Therefore, the power conversion systems are acceptable for operation at TPO uprate conditions.

The NRC concludes that based on the licensees evaluation provided in TSAR Section 7.0, there is reasonable assurance that the TPO uprate will not impact the ability of the power conversion systems from performing their intended functions.

3.8 TSAR Section 8.0 - Radwaste and Radiation Sources The following provides the NRC staffs technical review of the topics in Section 8.0 of the TSAR.

3.8.1 TSAR Section 8.1 - Liquid and Solid Waste Management Section 8.0 of the TSAR addressed the liquid and solid radioactive waste systems, which collect, process, and store radioactive waste for reuse, discharge of liquids, and shipment. The primary source of liquid and solid radioactive waste is from the condensate filter/demineralizers (CFD). The 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 increase in proportion to TPO uprate caused by the increase in power and flow through the CFD. The other systems contributing to liquid and solid waste experience no significant changes, and the total volume of process waste was not expected to increase an appreciable amount as a result of operation at TPO conditions.

These factors have no effect on plant safety and have minimal effect on the radioactive waste system, thus, the requirements of 10 CFR Part 20 and 10 CFR Part 50, Appendix I will continue to be met. Therefore, the NRC staff concludes that the liquid and solid radioactive waste handling capabilities are satisfactory for the TPO uprate conditions.

3.8.2 TSAR Section 8.2 - Gaseous Waste Management The gaseous radioactive waste management systems include the offgas system and the various building ventilation systems, which collect, control, and process gaseous radioactive waste.

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, and, thus, does not cause any adverse effect on the offgas system. The non-condensable radioactive gases from the main condenser will increase from the TPO uprate, but will be managed to meet the requirements of 10 CFR Part 20 and 10 CFR

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION Part 50, Appendix I. Therefore, the TPO uprate does not significantly affect the offgas system design or operation. Based on the above considerations, the NRC staff concludes that the gaseous radioactive waste handling capabilities are satisfactory for TPO.

3.8.3 TSAR Section 8.3 - Radiation Sources in the Reactor Core

=

Background===

On October 3, 2001, the NRC issued Amendment No. 134 (ADAMS Accession No. ML012600176), approving the implementation of the alternative source term in accordance with 10 CFR 50.67 following the guidance provided in applicable sections of RG 1.183. The NRC staff also considered relevant information in the HCGS UFSAR and TSs during its review.

For the EPU evaluation in 2008, the licensee performed a LOCA analyses at 102 percent of the EPU power level of 3,840 MWt, which equates to 3,917 MWt. Consistent with NRC regulations and regulatory guidance, the licensee determined the design basis analyses met the applicable accident dose acceptance criteria. Therefore, the current licensing basis dose consequence analyses remains bounding at the proposed MUR uprated power level of 3,902 MWt with a margin that is within the assumed uncertainty (0.374 percent) associated with advanced flow measurement techniques using the Cameron International CheckPlus' LEFM system.

NRC Staff Review of Radiation Sources in the Reactor Core and Coolant In TSAR Section 8.3 and 8.4, the licensee reviewed the radiological effects for the TPO uprate using CLB methodologies to verify that expected coolant concentrations at the uprated power level will be bounded by the CLB values. The NRC staff reviewed the various radioactive source terms associated with the MUR license amendment request (LAR) to ensure the adequacy of the sources of radioactivity used by the licensee as input to calculations to verify that the radioactive waste management systems have adequate capacity for the treatment of radioactive liquid and gaseous wastes.

TLTR, Appendix H, describes the methodology and assumptions for the evaluation of radiological effects for the TPO uprate. The NRC staffs review included the parameters used to determine: (1) the concentration of each radionuclide in the reactor coolant, (2) the fraction of fission product activity released to the reactor coolant, (3) concentrations of all radionuclides other than fission products in the reactor coolant, (4) leakage rates and associated fluid activity of all potentially radioactive water and steam systems, and (5) potential sources of radioactive materials in effluents that are not considered in the plants UFSAR.

The licensee provided its evaluation of the radionuclides in the reactor core. The core isotopic inventory is a function of the core power level. The reactor coolant isotopic activity concentration is a function of the core power level, the migration of radionuclides from the fuel, radioactive decay and the removal of radioactive material by coolant purification systems.

Radiation sources in the reactor coolant include activation products, activated corrosion products and fission products. Under the MUR conditions there is a necessary increase in the steam flow. An increase in steam flow will lead to an increase in the activation rate of metallic corrosion products resulting in an increase of corrosion products present in the reactor coolant.

This increase in steam flow tends to balance the increase in the activation products in the reactor coolant, resulting in no significant increase in the coolant concentrations.

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION The concentration of noble gases and other volatile fission products in the MSL will not change.

The increased production rate of these materials in the reactor core is offset by the corresponding increase in steam flow; therefore, the concentration of these materials in the steam line remains constant. Although the power uprate will result in an increase in the rate these materials are introduced into the main condenser and off-gas systems, these expected increases continue to be within the design margins of the off-gas system.

During reactor operation, some stable isotopes in the coolant passing through the core become radioactive (activated) as a result of nuclear reactions. For example, the non-radioactive isotope oxygen-16 is activated to become radioactive nitrogen-16 (N-16) by a neutron-proton reaction as it passes through the neutron-rich core at power. For the short lived activities, the most significant is N-16, the decreased transit (and decay) time in the main steam line and the increased mass flow of the steam results in a larger increase in these activities in the major turbine building components.

Another source of activity in the reactor coolant is from the activation of metallic corrosion products contained in the coolant as it passes through the reactor core. Under the MUR conditions, an increase in steam flow will lead to an increase in the activation rate of metallic corrosion products resulting in an increase of corrosion products present in the reactor coolant.

Under the MUR power uprate conditions, the activation rate in the reactor region increases with power, and the filter efficiency of the condensate demineralizers may decrease. The net result is an increase in the activated corrosion product production. However, total activated corrosion product activity levels in the reactor water remain less than the design basis activated corrosion product activity. Therefore, no change is required in the design basis activated corrosion product concentrations for the MUR power uprate.

3.8.4 TSAR Section 8.4 - Radiation Sources in the Reactor Coolant The NRC staffs review of TSAR Section 8.4 is combined with the review of TSAR Section 8.3 and is found in SE Section 3.8.3.

3.8.5 TSAR Section 8.5 - Radiation Levels The licensee reviewed the projected radiation exposure at normal operation radiation levels for the power uprate. Normal operation radiation levels increase slightly for the MUR power uprate.

During normal and post-accident conditions, radiation levels in most areas of the plant increase by no more than the percentage increase in power level. Previous analyses of post-accident radiation levels performed at the EPU power level uprate bound the effects of the MUR power uprate on the plant and the habitability of the on-site emergency response facilities. Due to the design of the shielding and containment surrounding the reactor vessel, and since the reactor vessel is inaccessible to plant personnel during operation, an increase in the radiation sources in the reactor core over the CLTP, will have no significant effect on occupational worker personnel doses during power operations. Similarly, the radiation shielding provided in the BOP (i.e., around radioactive waste systems, MSL, the main turbine, etc.) is conservatively sized such that the increased source terms discussed are not expected to significantly increase the dose rates in the normally occupied areas of the plant. The licensee has determined that dose rate increases due to MUR power uprate will remain within acceptable zone designations with the current shielding designs and is acceptable to the NRC staff.

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION The HCGS site was designed with sufficient margin for higher-than-expected radiation sources.

In addition, occupational exposure is controlled by the plant radiation protection program and is maintained within limits required by regulations. Thus, the increase in radiation levels does not affect radiation zoning or shielding in the various areas of the plant because it is offset by conservatism in the design, source terms, and analytical techniques. Therefore, no change is required in the design basis radiation protection design features for the MUR uprate. The current as low as reasonably achievable program practices at HCGS (i.e., work planning, source term minimization, etc.), coupled with existing radiation exposure procedural controls, will be able to compensate for the anticipated increases in dose rates associated with this MUR.

Therefore, the increased radiation sources resulting from this proposed MUR, as stated above, will not adversely impact the licensees ability to maintain doses resulting from plant operation with the applicable limits in 10 CFR 20 and as low as reasonably achievable.

3.8.6 TSAR Section 8.6 - Normal Operation Off-Site Doses There are two factors, associated with this power uprate that may impact public and offsite radiation exposures during plant operations. These are the possible increase in gaseous and liquid effluents released from the site, and the possible increase in offsite radiation exposure from radioactive plant components and solid wastes stored onsite, either directly or from atmospheric scatter (known as skyshine).

Skyshine is caused by the radioactive decay of N-16 in the steam bearing components within the turbine building. The gamma radiation emitted skyward interacts with air molecules and subsequently scattered back down to the ground where it can expose members of the public.

Since there is significantly less shielding above the steam bearing components in the turbine building than on the sides of these components, skyshine from N-16 gamma radiation can be a significant contributor to dose rates outside plant buildings. In addition, the practice of injecting hydrogen into the reactor coolant to reduce stress-corrosion cracking significantly increases the fraction of N-16 in the reactor water that is released into the steam during power operations.

The licensee reviewed the TS limits implementing the guidelines of 10 CFR 50, Appendix I. The review found the normal radiological effluent doses shows that at CLTP, the annual doses are a small fraction of the doses allowed by TS limits. The MUR power level uprate does not involve significant increases in the offsite dose from noble gases, airborne particulates, iodine, tritium, or liquid effluents. In addition, radiation from skyshine is not a significant exposure pathway.

Present offsite radiation levels are a negligible portion of background radiation. Therefore, the normal offsite doses are not significantly affected by operation at the MUR and remain below the limits of 10 CFR 20 and 10 CFR 50, Appendix I and is therefore considered acceptable.

3.9 TSAR Section 9.0 - Reactor Safety Performance Evaluations The following provides the NRC staffs technical review of the topics in Section 9.0 of the TSAR.

3.9.1 TSAR Section 9.1 - Anticipated Operational Occurrences Anticipated operational occurrences (AOOs) are transients initiated by a malfunction, a single failure of equipment, or a personnel error, and are expected to occur one or more times in the life of a plant.

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION As stated in HCGS UFSAR Chapter 15, AOOs are evaluated in three major categories:

1. Fuel thermal margin transient events that are due to:

a. decreases in reactor coolant temperature,

b. decreases in reactor coolant system flow rate,

c. reactivity and power distribution anomalies, and

d. increases in reactor coolant inventory.

2. Limiting transient overpressure event due to increase in reactor pressure.

3. Limiting loss of water level transient event is due to decrease in the reactor coolant inventory.

A limiting event within the above categories is an event or an accident that has the potential to affect the fuel safety and operating limits. The plant responses to the most limiting transients that affect the fuel thermal limits are analyzed by the licensee during each reload cycle and corresponding changes in the MCPR are added to the SLMCPR to establish the OLMCPR.

Regarding the fuel thermal margin transient events, the licensee referred to the generic evaluation of the AOOs in TLTR Appendix E, which ((

)) The GE methodologies GEMINI and GENESIS listed in TLTR Table E-1 used for the transient analysis have been approved by NRC (Reference 18).

In TLTR Table E-2, the licensee provided the effect of a 1.5 percent power uprate on the OLMCPR ((

)). The licensee stated that the OLMCPR changes for a 1.7 percent uprate, which bounds the HCGS 1.6 percent TPO uprate, may be slightly larger than shown in the TLTR Table E-2 but is still expected to be within the normal cycle-to-cycle variation.

The NRC staff reviewed the licensees analysis and determined that the evaluation of the fuel thermal margin transient events is acceptable based on the generic evaluation in TLTR Appendix E. The analysis after the reload core is configured will provide precise results. At that time, the cycle-specific SLMCPR will also be available, so that the required OLMCPR can be established. This process ensures that adequate fuel thermal margin will be maintained for TPO uprate operation.

The limiting transient overpressure events are analyzed in the analysis of record (HCGS EPU uprate analysis) at 102 percent CLTP and therefore bounds the TPO uprate.

The limiting loss of water level transient (i.e. loss of feedwater event) analyses are performed at 102 percent CLTP (HCGS EPU uprate analysis) and therefore bounds the TPO uprate.

Based on the above, the NRC staff determines that the evaluations performed at 102 percent CLTP (current analyses of record) show that for the TPO uprate conditions the requirements will continue to be met for: (1) GDC 10, because the acceptable fuel design limits are not affected by the AOOs; (2) GDC 15, because the RCPB design conditions are not affected by the AOOs; and (3) GDC 20, because the protection system will be automatically initiated to ensure that the

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION specified acceptable fuel design limits are not exceeded during AOOs.

TSAR Section 9.1.1 - Alternate Shutdown Cooling Evaluation The licensee stated that HCGS UFSAR Section 15.2.9.3 provides a qualitative evaluation of the ASDC mode of decay heat removal using only safety-related equipment. An alternate path is used to accomplish the shutdown cooling function using the RHR and SRV systems by transferring heat through the RHR heat exchangers. The heat removal capacity of using one or two RHR heat exchangers is determined to be acceptable under the TPO uprate as evaluated in SE Section 3.3.10. The NRC staff determines that the ASDC mode for decay and RHR is acceptable under the TPO uprate conditions because it only uses an alternate path of heat removal without changes to the RHR and SRV systems. Therefore, the NRC staff concludes the ASDC mode of the RHR system meets the requirements of GDC 38 because, in conjunction with the SRVs, the alternate path for decay and residual heat removal is not modified and will continue to perform its intended function under the TPO uprate conditions.

3.9.2 TSAR Section 9.2 - Design Basis Accidents The NRC staff reviewed the regulatory and technical analyses performed by the licensee as provided in the TSAR in support of its proposed MUR power level uprate license amendment, as they relate to the radiological consequences of design basis analyses (DBA) analyses. The NRC staff reviewed the impact of the proposed 1.6 percent MUR power uprate on DBA radiological consequence analyses, as documented in Chapter 15 of the UFSAR. The radiological consequences of a DBA are increased with a larger quantity of radioactivity released to the environment. This quantity is a function of the fission products released from the core as well as the transport mechanisms from the core to the release point. The radiological releases at the MUR uprate power are generally expected to increase in proportion to the core inventory increase, which is approximately in proportion to the power increase.

In its application, the licensee indicated that DBA events have been evaluated and analyzed to show that NRC regulations are met. DBA events have been previously analyzed at 102 percent of CLTP of 3,840 MWt, equating to 3,917 MWt, and bound this MUR power uprate of 3,902 MWt with a margin that is within the assumed uncertainty (0.374 percent) associated with the advanced flow measurement techniques using the Cameron LEFM system. The MS line break accident outside containment was evaluated using a 4 micro-Curie per gram (Ci/g) dose equivalent I-131 limit on reactor coolant activity. The limit on reactor coolant activity is unchanged for the MUR uprate condition. The NRC staff confirmed that the CLB dose consequence analyses remains bounding at the proposed MUR uprated power level of 3,902 MWt with a margin that is within the assumed uncertainty associated with advanced flow measurement techniques, including use of the Cameron International CheckPlus' LEFM system credited by the licensee.

The LEFM system has a continuously operating online self-diagnostic processes to verify that the digital circuits are operating correctly and within the design basis uncertainty limits. These processes can identify failure conditions that will cause the LEFM to switch from the normal (i.e.,

check plus) mode to the maintenance (i.e., check) mode or to the fail mode. See SE Section 3.5.4 for further discussion of the LEFM operation. The control room operators are provided a visual alarm when the LEFM system shifts from check plus mode (normal mode) to the check mode (maintenance mode). The visual alarm is displayed on the operator overview display screen.

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION When the LEFM system shifts from the normal mode to the maintenance mode, the visual alarm indicates that there has been a loss of LEFM system redundancy and uncertainty increases.

The LEFM systems fail mode indicates a loss of function that causes the LEFM system to operate outside the specified accuracy envelope.

The NRC staff confirmed that the licensee has accounted for the potential for an increase in measurement uncertainty should the LEFM system experience operational limitations.

The NRC staff reviewed the power level uncertainty associated with the operational modes of the LEFM as stated in SE Section 3.5.4 and the proposed TS changes for these limitations.

The calculated power level uncertainty associated with the LEFM flow measuring system in this condition is 0.66 percent. The plant can operate indefinitely at 3,889 MWt with only one LEFM subsystem operational. Power will be reduced to less than or equal to 3,889 MWt (CLTP) within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> if LEFM functionality cannot be restored to the Normal mode. A LEFM System Fail alarm indicates a loss of function. Reactor power will be reduced to less than or equal to 3,840 MWt (CLTP) within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> if LEFM functionality cannot be restored to either the NORMAL or MAINTENANCE mode. If the plant experiences a power decrease below 3,840 MWt (98.4 percent of rated thermal power) with the LEFM in the FAIL mode during the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> allowed outage time, the maximum permitted power level will be 3,840 MWt until the LEFM is restored to either NORMAL or MAINTENANCE mode operation. These required actions will ensure that the CLB dose consequence analyses remain bounding.

The NRC staff verified that the existing HCGS UFSAR Chapter 15 radiological analyses and release assumptions bound the conditions for the proposed TPO power uprate considering the higher accuracy of the proposed FW flow measurement instrumentation.

3.9.3 TSAR Section 9.3 - Special Events TSAR Section 9.3.1 - Anticipated Transient Without Scram The ATWS is an AOO with a failure of the reactor protection system to initiate a reactor scram to terminate the event. The requirements for ATWS are specified in 10 CFR 50.62. For BWRs, 10 CFR 50.62 requires, in part, that:

1) Each BWR have an alternate rod injection system that is designed to perform its function in a reliable manner and be independent (from the existing reactor trip system) from sensor output to the final actuation device.

2) Each BWR have an SLC system with the capability of injecting into the reactor vessel a borated water solution with reactivity control at least equivalent to the control obtained by injecting 86 gallons per minute (gpm) of a 13 weight percent sodium pentaborate decahydrate solution at the natural boron-10 isotope abundance into a 251 inch inside diameter reactor vessel.

3) Each BWR have equipment to trip the reactor coolant recirculation pumps automatically under conditions indicative of an ATWS.

The current ATWS analysis was performed at a power level of 3,952 MWt which bounds the TPO power. The analysis results and the acceptance criteria are given below in SE Table 4.

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION consistent with SHEX results. Therefore, the confirmatory calculations with SHEX (benchmarking with current licensing basis assumptions - pre-uprate) for plant specific modeling are not required for extended power uprates for Mark I and Mark III containment designs...

It is therefore concluded that the benchmarking analysis is not required, and therefore, use of the SHEX code is considered acceptable for the ATWS containment analysis.

Based on the above, the NRC staff determines that under the TPO uprate condition, the ATWS analysis continues to meet the requirements of (1) GDC 20, because the protection systems continue to be available for mitigation of an ATWS event; (2)GDC 38, because the RHR system continues to support decay and RHR from the suppression pool, and (3) 10 CFR 50.62, because the TPO uprate does not impact HCGSs ability to satisfy the BWR requirements (1),

(2), and (3) listed above.

TSAR Section 9.3.2 - Station Blackout For continued compliance with 10 CFR 50.63, Loss of all alternating current power, the licensee evaluated the following items:

(a) The adequacy of the condensate and reactor coolant inventory (b) The capacity of the Class 1E batteries (c) The station blackout (SBO) compressed nitrogen requirements (d) The ability to maintain containment integrity (e) The effect of loss of ventilation on rooms that contain equipment essential for plant response to an SBO event Regarding items (a) and (d), the licensee stated:

HCGS currently has margins of 21,768 gallons to the available condensate storage inventory and 7.4°F to the containment peak temperature limit. ((

)) Therefore, no HCGS-specific SBO analysis is performed for the TPO uprate.

The NRC staff reviewed the licensees justification that the generic SBO requirements (a) and (d) are applicable, and determines that the condensate storage inventory is acceptable because there is sufficient margin. Additionally, the NRC staff determines that containment integrity will be maintained because, the EPU containment analysis for the SBO event was performed assuming reactor thermal power 3,952 MWt which bounds the TPO uprate power.

In a supplement dated December 22, 2017, the licensee provided evaluations for items (b), (c),

and (e) under the TPO uprate conditions. In the December 22, 2017, response, the licensee stated:

The capacity of the Class 1E batteries, SBO compressed nitrogen requirements, and the evaluation of a loss of ventilation were not changed from the Current Licensed Thermal Power (CLTP) bases.

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION The NRC staff finds the licensees response acceptable for the SBO compressed nitrogen requirement, and the effect of loss of ventilation on rooms that contain equipment essential for plant response to an SBO event because these evaluations do not change from the CLTP bases.

In Section 9.3.2 Station Blackout of the TSAR, the licensee stated, in part, that the TLTR, Appendix L, Section L.5 Evaluation of Station Blackout, provides a generic evaluation of a potential loss of all AC power supplies based on previous plant response and coping capability analyses for typical power uprate projects. The previous power uprate evaluations have been performed according to the applicable bases for the plant.

The licensee further stated that applicable operator actions have previously been assumed to be consistent with the plant emergency procedure guidelines. These are the currently accepted procedures for each plant and SBO analysis. For the TPO uprate, there is no significant change in the time available for the operator to perform these assumed actions. The licensee also stated that HCGS currently has sufficient margins of condensate storage inventory and the containment peak temperature limit; and therefore, no further HCGS-specific SBO analysis is required to be performed for the TPO uprate.

In its supplement dated December 22, 2017, the licensee further stated that the Class 1E batteries are not changed from the CLTP bases.

The NRC staff reviewed the LAR and the letter dated December 22, 2017, and found that the TPO uprate will have no impact on the capacity of the Class 1E batteries, and minimal impact on SBO coping duration. Therefore, the staff finds that HCGS will continue to meet the requirements of 10 CFR 50.63 under the proposed MUR power uprate conditions.

Based on the above evaluation, the SBO analysis continues to meet the requirements of:

(a) GDC 20, because the required systems are available and initiated to ensure that the fuel design limit is not exceeded during the event; (b) GDC 38, because the peak suppression pool temperature during the event is bounded by the peak temperature during a LOCA and therefore the RHR system pumps will have adequate NPSH once it is available for containment cooling, and (c) 10 CFR 50.63, since the plant will continue to be able to recover from a SBO.

3.10 TSAR Section 10.0 - Other Evaluations The following provides the NRC staffs technical review of the topics in Section 10.0 of the TSAR.

3.10.1 TSAR Section 10.1 - High Energy Line Break Section 10.1 of the TSAR provided an evaluation of the effect of TPO operation on analyzed high energy line breaks (HELBs). The evaluation noted that operating temperatures and pressures either remain unchanged or change slightly for high energy systems, which results in no significant change in HELB mass and energy releases. The TSAR included summaries of the results for HELBs in the following systems: main steam, FW, ECCS, RCIC, RWCU, control rod drive, and auxiliary steam. The licensee determined that the existing HELB analyses for reactor-connected systems bound TPO uprate conditions because vessel dome pressure and other portions of the RCS remain at current operating pressure or lower. In addition, the locations of postulated breaks do not change because the locations are a function of the piping

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION configuration, which is unchanged for the TPO uprate. Therefore, the consequences of postulated HELBs remain acceptable for TPO uprate conditions.

3.10.2 TSAR Section 10.2 - Moderate Energy Line Break Section 10.2 of the TSAR addresses moderate energy line break (MELB) analyses. The plant-specific evaluation determined that none of the plant flooding zones contained a potential MELB location that would be affected by TPO uprate conditions. The evaluation noted that moderate energy systems that could be sources of flooding or spray potentially affecting safe-shutdown equipment are either not power-dependent or negligibly affected by the TPO uprate.

Therefore, the TPO uprate does not affect the existing evaluations of breaks in moderate energy lines.

3.10.3 TSAR Section 10.3 - Environmental Qualification In TSAR Section 10.3 Environmental Qualification, the licensee stated that the TPO increase in power level increases the radiation levels experienced by equipment during normal operation and accident conditions. However, the licensee explained that because the TPO uprate does not increase the nominal vessel dome pressure, there is only a small effect on pressure and temperature conditions experienced by equipment during normal operation and accident conditions. Therefore, the licensee stated that the resulting environmental conditions are bounded by the existing environmental parameters specified for use in the EQ program.

The licensee also stated that the environmental conditions for safety-related electrical equipment were reviewed to ensure that the existing qualification for the normal and accident conditions expected in the areas where the devices are located remain adequate. The licensee concluded that no change is needed for the TPO uprate.

In its supplement dated December 14, 2017, the licensee further stated that the TPO uprate does not affect environmental parameters for EQ equipment. In TSAR under Section 6.6 Power-Dependent Heating, Ventilation, and Air Conditioning identified that the increased heat loads in the drywell are within the capacity of the drywell cooling system. Normal area temperatures are based on the design specification of the drywell cooling system and any actual increases in local temperatures are bounded by the analyzed area temperature. The DBA containment temperature and pressure were determined for EPU conditions using a core thermal power of 4,031 MWt (1.02 x 3,952 MWt), with the exception of peak bulk suppression pool temperature, which was determined at 3917 MWt (1.02 x 3840 MWt). The results are documented in Table 4-1 of Attachment 4 of the HCGS EPU LAR (ADAMS Accession No. ML062690086). These results bound operation at TPO conditions. The EQ testing is performed using bounding pressure and temperature profiles with peak values of 62 psig and 340°F, which bound the accident pressure and temperature conditions for TPO. The DBA containment humidity is conservatively assumed to be 100 percent for 100 days. The EQ equipment is tested to 100 percent humidity. The current normal radiation dose is based on a core thermal power of 3,952 MWt and a plant life of 60 years. Any actual increases in normal dose rates are bounded by analyzed dose rates.

In its supplement dated December 14, 2017, the licensee stated that the DBA containment radiation conditions were determined for EPU using a core thermal power of 3917 MWt (1.02 x 3840 MWt). This value bounds the TPO uprate power level, including heat balance uncertainty.

In conclusion, the licensee stated that the existing analyses bound operation at TPO conditions

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION and all EQ electrical equipment remains qualified for TPO uprate conditions under normal and accident conditions.

The NRC staff reviewed the LAR and the letter dated December 14, 2017, and concludes that the TPO conditions are bound by the existing electrical equipment EQ and the equipment remains qualified for the TPO conditions. Therefore, the NRC staff determines that implementation of the proposed MUR power uprate will have no adverse impact on the HCGS EQ Program or its ability to continue to meet the requirements of 10 CFR 50.49.

3.10.4 TSAR Section 10.4 - Testing As stated in TLTR Section 5.11.9, the guidelines for the TPO uprate power ascension testing plan are provided in TLTR, Appendix L, Section L.2. In the HCGS TSAR, Section 10.4, the licensee stated that the power ascension test plan is based on the guidelines in the TLTR. The key elements of the planned testing include the following:

Routine measurements of reactor and system pressures, flows, and vibration of selected major rotating equipment will be taken near 95 percent CLTP, 100 percent CLTP, and at full TPO thermal power conditions.

Demonstration of acceptable fuel thermal margin will be performed prior to and during power ascension at each of the steady state heat balance conditions at the above power levels.

Performance of the pressure and feedwater/level control system will be recorded during power ascension. These performance checks will utilize the methods and criteria described in the original startup testing of these systems to demonstrate acceptable operational capability.

The turbine pressure controller setpoint will be readjusted in order that the reactor dome pressure at the TPO power is the same as at CLTP.

The NRC staff reviewed the proposed testing and determines that it is consistent with the generic guidelines in TLTR, Appendix L, Section L.2.

The licensee also stated that large transient testing is not necessary since the increase in power for the TPO uprate is small and that testing during initial plant startup demonstrated the adequacy of the safety and protection systems for such large transients. The licensee also stated that operational occurrences have shown that the plant response is bounded by the safety analyses for these events.

The NRC staff determines that the licensees justification for not performing large transient tests is consistent with the generic justification provided in TLTR, Appendix L, Section L.2, therefore, the NRC staff concludes that the HCGS testing plan for the TPO uprate is acceptable.

3.10.5 TSAR Section 10.5 - Operator Training and Human Factors Operator Training The licensee stated in LAR Enclosure 1, Section 3.4.6, Operator Training, Human Factors, and Procedures, that the operator responses to plant transients and accidents are unaffected by the

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION proposed power uprate. Furthermore, there are no new manual operator actions being created and existing manual actions will not be automated. Therefore, there will be no training related to new operator actions or timing.

There is a new alarm that will be received in the control room with a 72-hour AOT that operators must diagnose if the LEFM system or a portion of the system becomes non-functional. The licensee states in Enclosure 1 of the submittal, Section 3.2.4, Disposition of NRC Criteria for Use of LEFM Topical Reports, that a control room alarm response procedure will be developed to provide guidance to the operators for initial alarm diagnosis. Cameron documentation will be used to develop the specific procedures for operators and maintenance response actions.

Based on the statements above, the staff concludes that the proposed changes to the operator training program do not adversely affect defense-in-depth or safety margins. The staff finds that the statements provided by the licensee are in conformance with Section VII, Item 2.D of to RIS 2002-003.

Operator Actions The licensee stated in LAR Enclosure 1, Section 3.4.6, that no new manual operator actions were created and no existing manual actions were automated as a result of the proposed amendment. Additionally, the time available for operators to respond is not affected (i.e., no reduction in time). Operator response to plant transients and accidents remains unaffected by the proposed power uprate change.

The licensee reviewed the following safety analyses for potential impact for operator actions:

(1) containment system performance, (2) fire protection, and (3) special events, specifically, ATWS and SBO. The operator actions for these analyses remain unaffected.

In the LAR Enclosure 1, Section 3.2.4, Disposition of NRC Criteria for Use of LEFM Topical Reports, the licensee described its plans to make a software change to the plant process computer that will include a change to the alarms. The alarm will alert the operators to changes in the LEFM instrumentation status but only for sustained loss of data between the LEFM and the plant computer; the LEFM system will be considered non-functional should that happen.

When the alarm is received in the control room, a 72-hour AOT begins. The NRC staffs review of the 72-hour AOT is provided in SE Section 3.5.4. In addition to the 72-hour AOT, the licensee stated that should the LEFM become non-functional, the feedwater flow input to the core thermal power calculation would then be provided by the existing feedwater flow venturis, and temperature input would be provided by the existing resistance temperature detector. The control room alarm response procedures that provide guidance to the operators for initial alarm diagnosis will be developed prior to raising power above the CLTP.

Based on the statements above, the NRC staff concludes that the proposed MUR power uprate will not adversely impact the licensee-identified operator actions or their response times.

Additionally, the NRC staff concludes that the statements provided by the licensee are in conformance with Section VII, Item 1 of Attachment 1 to RIS 2002-03.

3.10.6 TSAR Section 10.6 - Plant Life In TSAR Section 10.6, the licensee evaluated the effect of the proposed TPO uprate on age-related degradation of plant equipment. The licensee stated two degradation mechanisms that may be influenced by the TPO uprate: 1) irradiation assisted stress corrosion cracking and

2) FAC. The licensee stated that the equipment sensitivity to these changes is small and that existing programs are able to monitor the effects of any degradation. These programs include

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION environmental qualification (stated in SE Section 3.10.3), FAC (stated in SE Section 3.3.5), and in-service inspection. The licensee stated that these programs do not change for the TPO uprate. Further, the licensee concluded that no changes to maintenance, inspection, testing or surveillance procedures are required because there are no significant changes in the operating conditions.

The NRC staff reviewed the RFOL, and the HCGS UFSAR Appendix A License Renewal Final Safety Analysis Report Supplement, which contains the specific programs and activities for managing the effects of aging and the evaluation of time-limited aging analyses for operation beyond the original 40-year operating license.

The NRC staff concludes that, with respect to age-related degradation, no additional maintenance, inspection, testing or surveillance procedures are needed because the changes in operating conditions for the TPO uprate are small and HCGS has aging management programs in place to detect age-related degradation. Thus, there is reasonable assurance that the plant equipment will continue to perform its safety function for the TPO uprate.

3.10.7 TSAR Section 10.7 - NRC and Industry Communications In Section 10.7 of Enclosure 1 of the LAR, the licensee states that NRC and industry communications are generically addressed in the TLTR, Section 10.8. The NRC staff reviewed the TLTR, Appendix B, Section B.4 Review of NRC and Industry Generic Communications, which concludes that a plant-specific review of generic NRC and industry communications is not needed for a TPO uprate. Therefore, the NRC staff concludes that the LAR is acceptable with respect to this topic since the disposition is consistent with the TLTR.

3.10.8 TSAR Section 10.8 - Plant Procedures and Programs Changes to Control Room Controls, Displays, and Alarms The licensee stated in Enclosure 1 of the LAR, Section 3.2.1 LEFM Feedwater Flow and Temperature Measurement, that there will be software changes to the plant computer to support the interface with the Cameron LEFM Check Plus system for operation above the CLTP limit of 3840 MWt. The change is an alarm provided by the plant computer when there is a change in the LEFM system status. This alarm occurs when there is a sustained loss of data between the LEFM and the plant computer. In such a case, the core thermal power calculations automatically revert to the calibrated venturi output. Additionally, the licensee stated that if the interface between the LEFM system and the plant computer has failed, the LEFM will be considered non-functional and the appropriate procedural actions will be applied.

The licensee stated in Enclosure 1, Section 3.4.3, Plant Modifications, of the LAR that the software modification will be made per the requirements of 10 CFR 50.59, Changes, Tests, and Experiments, and will be implemented prior to, or concurrently with, the proposed power uprate implementation. The LEFM system software and the plant computer software configuration will be maintained using HCGS procedures, which include verification and validation of changes to software configuration per the licensee statement in Section 3.2.4, Disposition of NRC Criteria for Use of LEFM Topical Reports, in Enclosure 1 of the submittal.

With regard to the control room displays, the minor changes to the power to flow map and the flow-referenced setpoints will be communicated through normal operator training.

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION Based on the statements above, the NRC staff concludes that the proposed changes to the control room controls, displays, and alarms do not adversely affect defense-in-depth or safety margins. The staff finds that the statements provided by the licensee are in conformance with Section VII, Item 2.B of Attachment 1 to RIS 2002-003.

Control Room Plant Reference Simulator The licensee stated in the LAR Enclosure 1, Section 3.4.6, Operator Training, Human Factors, and Procedures, that the simulator will be modified for the uprated conditions. Validation will be performed in accordance with established HCGS plant simulator certification testing procedures. The simulator modifications are considered to be minor as there are no new manual operator actions being proposed or automated, and no reduction in time for the required operator actions.

Based on the statements above, the NRC staff concludes that the proposed changes to the control room plant Reference simulator do not adversely affect defense-in-depth or safety margins. The staff finds that the statements provided by the licensee are in conformance with Section VII, Item 2.C of Attachment 1 to RIS 2002-003.

Plant Modifications to Support Power Uprate In the LAR Enclosure 1, the licensee described modifications to support the proposed amendment. The NRC staff determined that there are no physical modifications that will affect operator actions or training. However, there is a software change to the plant computer to support the LEFM system for operation above the CLTP limit of 3,840 MWt, as discussed in Section 3.5.4 of this evaluation.

The staff finds that the statements provided by the licensee are in conformance with Section VII, Item 3 of Attachment 1 to RIS 2002-003.

Temporary Operation Above Licensed Full Power Level The licensee refers to the following statement in section, 3.2.6, Power Reactor Monitoring of, as the cross-reference to RIS 2002-03, Attachment 1,Section VII.4:

Plant procedures provide requirements for monitoring and controlling reactor power in compliance with the TS.

On December 14, 2017, the NRC staff requested additional information from the licensee to summarize the revisions to be made to plant procedures associated with reduction of the magnitude of the allowed deviation from the licensed power level (ADAMS Accession No. ML17348A624).

The licensee responded on December 19, 2017 (ADAMS Accession No. ML17353A778) with the following:

The reduction in uncertainty for measurement of licensed thermal power level does not impact how Hope Creek controls operation of the plant at the allowed licensed thermal power limit. Procedures HC.OP-IO.ZZ-0006, Power Changes During Operation, and OP-HC-300-2020, Review of Reactor Core Performance

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION Information, implement controls for operation at the licensed thermal power level.

Hope Creek monitors a 1-hour average of core thermal power and takes action to ensure that he 1-hour average does not exceed the Licensed Power Limit. In addition to monitoring the 1-hour average, Hope Creek also monitors a 5-minute average of core thermal power. If the 5-minute average exceeds a value of 3848 MWt (current licensed power limit plus 8 MWt) then prompt action is taken to reduce the 5 minute average below this value. In addition the procedures require that core thermal power level over a completed shift does not exceed the Licensed Power Limit. These controls are consistent with Regulatory Information Summary (RIS) 2007-21, Rev 1, Adherence to Licensed Power Limits, and the NEI Position Statement - Guidance to Licensees on Complying with the Licensed Power Limit (ADAMS No. ML081750537). The NEI Position Statement was endorsed by NRC Safety Evaluation issued October 8, 2008 (ADAMS No. ML082690105). The above procedures will be revised during implementation of the license amendment request to reflect the new licensed thermal power limit.

The staff concludes that the statements provided above by HCGS are in conformance with Section VII, Item 4 of Attachment 1 to RIS 2002-003.

3.10.9 TSAR Section 10.9 - Emergency Operating Procedures The licensee stated in the LAR Enclosure 1 that necessary operating procedure revisions, including emergency operating procedures (EOPs) and abnormal operating procedures, will be completed prior to implementation of the proposed changes.

Specific to EOPs, the licensee stated in TSAR Section 10.9, Emergency Operating Procedures, that the thermal power optimization (TPO) uprate will have a negligible effect or no effect on the operator action thresholds and on the EOPs in general.

Based on the statements above, the NRC staff determines that the proposed changes to the EOPs and abnormal operating procedures do not adversely affect defense-in-depth or safety margins. Additionally, the NRC staff concludes that the statements provided by the licensee are in conformance with Section VII, Item 2.A of Attachment 1 to RIS 2002-003.

3.10.10 TSAR Section 10.10 - Individual Plant Examination In Section 10.10 of Enclosure 1 of the LAR, the licensee described that HCGS maintains and updates a station probabilistic risk assessment model and that this model is integrated with station operations and decision-making. The licensee stated that the model and analysis will not be updated for the TPO uprate because the change in risk is insignificant. The NRC staff reviewed TLTR Section 5.11.11 Probabilistic Safety Assessment, which discusses that the change in risk due to a TPO uprate is insignificant. Thus, the TLTR concluded that neither the plant-specific individual plant examination nor the probabilistic risk assessment need be updated to support the TPO uprate. The NRC staff concludes that the LAR is acceptable with respect to this topic since the disposition is consistent with the TLTR.

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION 3.11 License and Technical Specification Changes The following provides the NRC staff evaluation of the proposed RFOL and TS changes associated with the LAR. The proposed changes to the RFOL and TSs are stated in Section 2.0 of Attachment 1 to the application and a markup of the changes is shown in to the licensees application.

3.11.1 RFOL Paragraph 2.C.(1) - Maximum Power Level The licensee proposed the following changes to the renewed facility operating license:

The value of RTP for Hope Creek Renewed Facility Operating License Number NPF-57, Section 2.C.(1), Maximum Power Level, is revised from 3840 MWt to 3,902 MWt.

The NRC staff reviewed the proposed change to RFOL paragraph 2.C.(1), and confirmed that the change reflects the approximate 1.6 percent increase in thermal power level and is consistent with the licensees supporting safety analyses. Therefore, the NRC staff concludes that the proposed change to RFOL paragraph 2.C.(1) is acceptable.

3.11.2 RFOL Paragraph 2.C.(11) - Partial Feedwater Heating The licensee proposed to change the value of RTP FW temperature in RFOL paragraph 2.C.(11) Partial Feedwater Heating from 329.6°F to 331.5°F. The staff reviewed the proposed change and found it acceptable since the change is consistent with the licensees supporting safety analyses. Therefore, the NRC staff concludes that the proposed change to paragraph 2.C.(1) is acceptable.

3.11.3 TS 1.35 - Definitions The licensee proposed to change the TS 1.35 definition for Rated Thermal Power from the CLTP level of 3840 MWt to the proposed TPO power level of 3902 MWt. This change reflects the proposed approximate 1.6 percent increase in thermal power level and is consistent with 10 CFR 50.36 and the licensees supporting safety analyses. Therefore, the NRC staff concludes that the proposed change to TS 1.35 is acceptable.

3.11.4 TS 2.2 - Limiting Safety System Settings Flow-Biased Scram - Allowable Value Changes HCGS TS Table 2.2.1-1, Function 2.b, and its associated note ** contain requirements for the average power range monitor (APRM) Simulated Thermal Power - Upscale function. This function, also known as the APRM flow-biased scram, currently has the following allowable valuates (AVs):

Two Loop Operation:

0.57w + 61 percent RTP Single Loop Operation:

0.57(w - 9 percent) + 61 percent RTP

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION As stated in TS Table 2.2.1-1, w is recirculation loop drive flow.

The licensee proposed the following changes to the flow-biased scram allowable values:

Two Loop Operation:

0.56w + 60 percent RTP Single Loop Operation:

0.56(w - 9 percent) + 60 percent RTP Flow-Biased Scram - Trip Setpoint Value Changes Also in TS Table 2.2.1-1, Function 2.b, the APRM flow-biased scram currently has the following trip setpoint values:

Two Loop Operation:

0.57w + 59 percent RTP Single Loop Operation:

0.57(w - 10.6 percent) + 59 percent RTP The licensee proposed the following changes to the flow-biased scram trip setpoint values:

Two Loop Operation:

0.56w + 58 percent RTP Single Loop Operation:

0.56(w - 10.8 percent) + 58 percent RTP As stated in TSAR Section 5.3.7, the APRM AVs for both two-loop and single-loop operation are unchanged in units of absolute core thermal power versus recirculation drive flow; however, because the setpoints are expressed in percent of RTP, they decrease in proportion to the power uprate. This is consistent with NRC-approved GE BWR uprates in TR NEDC-32424P-A Generic Guidelines for General Electric Boiling Water Reactor Extended Power Uprate dated September 1996 (Proprietary information. Not publicly available). The staff reviewed the changes and determined they are acceptable because they are consistent with the 10 CFR 50.36(b) and the licensees supporting safety analyses. Therefore, the NRC staff concludes that the proposed change to TS 2.2 is acceptable.

3.11.5 TS Limiting Condition for Operation (LCO) 3.1.3.1 Control Rod Operability, TS 3.1.4.1 LCO Rod Worth Minimizer, and TS 3.10.2 LCO Rod Worth Minimizer TS LCO 3.1.3.1, Applicability, LCO Required Actions c. and d., Reference note ***** that states the action is not applicable when greater than 8.6 percent RTP. The licensee proposed to change the note to greater than 8.6 percent RTP. Additionally, TS LCO 3.1.4.1 and TS LCO 3.10.2, Rod Worth Minimizer proposed to change the applicability from 8.6 percent RTP to 8.5 percent RTP.

As stated in TSAR Section 5.3.8, the RWM low power setpoint is used to enforce the rod patterns established for the control rod drop accident at low power levels. This power setpoint has been scaled in terms of percent power to maintain the value in absolute power, thus, it is proposed to be 8.5 percent RTP.

The NRC staff reviewed the change and determines it is acceptable because it is consistent with the generic evaluation in TLTR Section F.4.2.9, it continues to meet 10 CFR 50.36(c)(2) since it addresses the lowest functional capability of the control rods, and is consistent with the

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION licensees supporting safety analysis. Therefore, the NRC staff concludes that the proposed changes are acceptable for the TPO uprate.

3.11.6 Changes to TS 3.3.6 Control Rod Block Instrumentation Control Rod Block Instrumentation - Allowable Values HCGS TS Table 3.3.6-2, Function 2.a, contains requirements for the APRM Simulated Thermal Power - Upscale function. This function currently has the following flow biased AVs:

Two Loop Operation:

0.57w + 56 percent RTP Single Loop Operation:

0.57(w - 9 percent) + 56 percent RTP As stated in Table 3.3.6-2 note *, the rod block function is varied as a function of recirculation loop flow (w).

The licensee proposed the following changes in Table 3.3.6-2 to the flow-biased AVs:

Two Loop Operation:

0.56w + 55.1 percent RTP Single Loop Operation:

0.56(w - 9 percent) + 55.1 percent RTP Control Rod Block Instrumentation - Trip Setpoint Value Changes Also, TS Table 3.3.6-2, Function 2.a, contains the following trip setpoint values for the APRM Simulated Thermal Power - Upscale function:

Two Loop Operation:

0.57w + 54 percent RTP Single Loop Operation:

0.57(w - 10.6 percent) + 54 percent RTP The licensee proposed the following changes to the flow-biased scram trip setpoint values:

Two Loop Operation:

0.56w + 53.1 percent RTP Single Loop Operation:

0.56(w - 10.8 percent) + 53.1 percent RTP As stated in TSAR Section 2.1, the proposed changes to the Simulated Thermal Power -

Upscale functions are based on the approach in TLTR Section F.4.2.1, Flow Reference APRM Trip and Alarm Setpoints. The licensee stated that the absolute power is unchanged versus recirculation drive flow and decreases in proportion to the power uprate.

The staff reviewed the changes and determines they are acceptable because they are consistent with the TLTR methodology in Section F.4.2.1. Therefore, the NRC staff concludes that the proposed change to TS 2.2 is acceptable.

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION 3.11.7 Changes to TS 3.4.1.1 LCO and TS 4.4.1.1.1.a. SR for the Recirculation System The TS 3.4.1.1 LCO action statement a.1.b), and TS SR 4.4.1.1.1.a SR currently require the licensee to reduce thermal power to less than or equal to 60.86 percent RTP. The licensee proposed changes to both of these requirements to reduce thermal power to less than or equal to 59.86 percent RTP.

The staff reviewed the change and found it acceptable because the values in TS are expressed in percent of RTP, the licensee rescaled the thermal power in proportion to the power increase.

This is also consistent with the methodology described in the TLTR Section 5.2; therefore, this change is acceptable for the TPO uprate.

3.12 Technical Evaluation Conclusion Based on the considerations discussed in SE Sections 3.2 through 3.11, the NRC staff concludes that the proposed TPO uprate is acceptable.

STATE CONSULTATION In accordance with the Commissions regulations, the Pennsylvania State official was notified of the proposed issuance of the amendments. On March 9, 2018, the State official responded that he had no comments.

ENVIRONMENTAL CONSIDERATION The amendments change a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and changes SRs.

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 (82 FR 46098). 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.

CONCLUSION The Commission has concluded, based on the considerations stated 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 Commissions 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:

J. Hughey N. Iqbal S. Jones D. Ki

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION M. Li K. Nguyen B. Parks A. Sallman G. Singh T. Sweat C. Sydnor A. Young L. Regner Date: April 24, 2018 Appendices:

A. List of Acronyms

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION

-A1-APPENDIX A LIST OF ACRONYMS ACRONYM DEFINITION AC Alternating Current ADAMS Agencywide Documents Access and Management System ADS Automatic Depressurization System AIOG Artificial Island Operating Guide AL Analytical Limit AOO Anticipated Operational Occurrences APRM Average Power Range Monitor ART Adjusted Reference Temperature ASDC Alternate Shutdown Cooling ASME American Society of Mechanical Engineers ASME Code ASME Boiler and Vessel Pressure Code ATWS Anticipated Transient Without Scram BOP Balance of Plant BSP Backup Stability Protection BTP Branch Technical Position BWR Boiling-Water Reactor BWRVIP BWR Vessel and Internals Project BWROG Boiling-Water Reactors Owners Group CAP Containment Accident Pressure CCF Common-Cause Failure CDA Confirmation Density Algorithm CDF Core Damage Frequency CFR Code of Federal Regulations CLB Current Licensing Basis CLTP Current Licensed Thermal Power CS Core Spray CSS Containment Spray Cooling Cu Copper CW Circulating Water DBA Design Basis Accident DC Direct Current DSS-CD Detect and Suppress Solution - Confirmation Density ECCS Emergency Core Cooling System EDG Emergency Diesel Generator EFPY Effective Full Power Years EHC Electro-hydraulic Control EMA Equivalent Margins Analysis EOC End-of-Cycle EOPs Emergency Operating Procedures EPRI Electric Power Research Institute EPU Extended Power Uprate

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-A2-OFFICIAL USE ONLY - PROPRIETARY INFORMATION ACRONYM DEFINITION EQ Environmental Qualification

°F Degrees Fahrenheit FAC Flow-Accelerated Corrosion FIV Flow-Induced Vibration ft-lbs Foot-Pounds FW Feedwater GDC General Design Criterion/Criteria GE General Electric GEH GE-Hitachi Nuclear Energy GL Generic Letter GNF Global Nuclear Fuel Gpm gallons per minute HCGS Hope Creek Generating Station HELB High Energy Line Break HPCI High Pressure Coolant Injection HRA Human Reliability Analysis HSBW Hot Shutdown Boron Weight HVAC Heating, Ventilating, and Air Conditioning HVCS Hardened Containment Vent Systems IRM Intermediate Range Monitor ISI Inservice Inspection ISP Integrated Surveillance Program kWt Kilowatts Thermal LCO Limiting Condition for Operation LAR License Amendment Request LEFM Leading Edge Flow Meter LHGR Linear Heat Generation Rate LOCA Loss-of-Coolant Accident LPCI Low Pressure Coolant Injection LPRM Local Power Range Monitor LR License Renewal LTR Licensing Topical Report MAPLHGR Maximum Average Planar Linear Heat Generation Rate MCPR Minimum Critical Power Ratio MELB Moderate Energy Line Break MELLLA Maximum Extended Load Line Limit Analysis MOV Motor-Operated Valve MS Main Steam MSIV Main Steam Isolation Valve MSIVC Main Steam Isolation Valve Closure MSL Main Steam Line MSO Multiple Spurious Operation MUR Measurement Uncertainty Recapture

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-A3-OFFICIAL USE ONLY - PROPRIETARY INFORMATION ACRONYM DEFINITION MWe Megawatts Electric MWt Megawatts Thermal N-16 Nitrogen-16 NEI Nuclear Energy Institute NPSH Net Positive Suction Head NPSHa Net Positive Suction Head Available NPSHr Net Positive Suction Head Required NRC U.S. Nuclear Regulatory Commission NSSS Nuclear Steam Supply System OLMCPR Operating Limit Minimum Critical Power Ratio OLTP Original Licensed Thermal Power OPRM Oscillation Power Range Monitor P-T Pressure-Temperature PCS Pressure Control System PCT Peak Cladding Temperature PFM Probabilistic Fracture Mechanics PRA Probabilistic Risk Analysis PSEG PSEG Nuclear LLC PTLR Pressure-Temperature Limits Report RAMA Radiation Analysis Modeling Application RBM Rod Block Monitor RCIC Reactor Core Isolation Cooling RCPB Reactor Coolant Pressure Boundary RCS Reactor Coolant System Rem Roentgen Equivalent Man RFOL Renewed Facility Operating License RG Regulatory Guide RHR Residual Heat Removal RIPD Reactor Internal Pressure Difference RIS Regulatory Issue Summary RPT Recirculation Pump Trip RPV Reactor Pressure Vessel RRS Reactor Recirculation System RS Review Standard RSD Replacement Steam Dryer RTNDT Reference Nil-Ductility Temperature RTP Rated Thermal Power RWCU Reactor Water Cleanup RWM Rod Worth Minimizer SACS Safety Auxiliaries Cooling System SAFDL Specified Acceptable Fuel Design Limit SAR Safety Analysis Report SBO Station Blackout

OFFICIAL USE ONLY - PROPRIETARY INFORMATION

-A4-OFFICIAL USE ONLY - PROPRIETARY INFORMATION ACRONYM DEFINITION SC Safety Communication SDC Shutdown Cooling SE Safety Evaluation SLC Standby Liquid Control SLCS Standby Liquid Control System SLMCPR Safety Limit Minimum Critical Power Ratio SFP Spent Fuel Pool SPC Suppression Pool Cooling SR Surveillance Requirement SRP Standard Review Plan SRV Safety Relief Valve SSCs Structures, Systems, and Components SSWS Station Service Water System STP Stimulated Thermal Power TACS Turbine Auxiliaries Cooling System TIP Traversing Incore Probes TLMT Thermal Limits Monitoring Threshold TLTR Thermal Power Optimization Licensing Topical Report TPO Thermal Power Optimization TS Technical Specification TSAR Thermal Power Optimization Safety Analysis Report UFM Ultrasonic Flow Meter UFSAR Updated Final Safety Analysis Report UHS Ultimate Heat Sink USE Upper Shelf Energy

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SUBJECT:

HOPE CREEK GENERATING STATION - ISSUANCE OF AMENDMENT RE: MEASUREMENT UNCERTAINTY RECAPTURE POWER UPRATE (CAC NO. MF9930; EPID L-2017-LLS-0002) DATED APRIL 24, 2018.

DISTRIBUTION:

PUBLIC PM File Copy RidsNrrDssStsb Resource RidsRgn1 MailCenter Resource RidsNrrDssSnpb Resource RidsNrrDeEicb Resource RidsNrrDeEseb Resource RidsNrrDmlrMvib Resource RidsNrrDraArcb Resource RidsNrrPMHopeCreek Resource SJones, SCPB TSweat, STSB KNguyen, EEOB AYoung, MVIB DKi, APHB ADAMS A ccess1on N ML18096A542 o.:

OFFICE NRR/DORL/LPL4/PM NAME LRegner DATE 4/5/2018 OFFICE NRR/DSS/SNPB/BC*

NAME SAnderson DATE 3/28/2018 OFFICE NRR/DE/EICB/BC*

NAME MWaters DATE 3/26/2018 OFFICE NRR/DMLR/MCCB/BC*

NAME SBloom DATE 4/2/2018 OFFICE NRR/DRA/APHB/BC*

NAME CJFong DATE 4/2/2018 OFFICE NRR/DORL/LPL 1/BC NAME Joanna DATE 4/23/2018 RidsACRS_MailCTR Resource RidsNrrLAIBetts Resource RidsNrrDorllpl1 Resource RidsNrrDssSrxb Resource RidsNrrDssScpb Resource RidsNrrDeEeob Resource RidsNrrDmlrMccb Resource RidsNrrDraAplb Resource ASallman, SRXB AChereskin, MCCB BParks, SNPB Mli, EICB CBasavaraju, EMIB Nlqbal, APLB EDickson, SRCB

• b *yema1

,y memo NRR/DORL/LPL 1 /LA N RR/DSS/SRXB/BC*

I Betts JWhitman 4/13/2018 4/3/2018 NRR/DSS/SCPB/BC.._

NRR/DSS/STSB/BC*

RDennig VCusumano 3/5/2018 3/27/2018 NRR/DE/EEOB/BC*

NRR/DE/ESEB/BC*

JQuichocho SBailey 3/20/2018 2/26/2018 NRR/DMLR/MVIB/BC.._

NRR/DRA/APLB/BC.._

SRuffin JRobinson (A) 2/21/2018 11/14/2018 NRR/DRA/ARCB/BC.._

OGC-NLO*

KHsueh JGillespie 2/16/2018 4/19/2018 NRR/DORL NRR/DORL/LPL4/PM Tlnverso LRegner 4/23/2018 4/24/2018 OFFICIAL RECORD COPY OFFICIAL USE ONLY PROPRIETARY INFORMATION