Text

License Amendment Request for Main Steam Safety Valve Lift Setting Tolerance Change
	ML15362A169
Person / Time
Site:	Harris
Issue date:	12/17/2015
From:	Waldrep B; Duke Energy Progress
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
Shared Package
ML15362A168	List: HNP-15-038, License Amendment Request for Main Steam Safety Valve Lift Setting Tolerance Change;
References
	HNP-15-038
	Download: ML15362A169 (101)
	v • d • e

Benjamin C. Waldrep Vice President Harris Nuclear Plant 5413 Shearon Harris Rd New Hill NC 27562-9300 919-362-2502 PROPRIETARY INFORMATION - WITHHOLD UNDER 10 CFR 2.390 UPON REMOVAL OF ATTACHMENT C THIS LETTER IS UNCONTROLLED 10 CFR 50.90 December 17, 2015 Serial: HNP-15-038 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Shearon Harris Nuclear Power Plant, Unit 1 Docket No. 50-400/Renewed License No. NPF-63

Subject:

License Amendment Request for Main Steam Safety Valve Lift Setting Tolerance Change Technical Specifications (TS) Sections:

2.2.1, Limiting Safety System Settings Reactor Trip System Instrumentation Trip Setpoints 3.4.3, Pressurizer 3.7.1.1, Turbine Cycle Safety Valves Ladies and Gentlemen:

In accordance with the provisions of 10 CFR 50.90, Duke Energy Progress, Inc. (Duke Energy),

hereby requests a revision to the Technical Specifications (TS) for Shearon Harris Nuclear Power Plant, Unit 1 (HNP). The proposed license amendment revises the as-found lift setting tolerance for main steam line code safety valves (MSSVs) in TS 3.7.1.1, Table 3.7-2, from +/- 1%

to +/- 3%. To support the MSSV setpoint tolerance change, changes are required to TS 2.2.1, Table 2.2-1. Specifically, the reactor trip system instrumentation trip setpoint for pressurizer water level-high percentage of the instrument span is reduced from 92% to 87%. Further, the allowable value of the instrument span is requested to be reduced from 93.5% to 88.5%. A change to reduce the maximum pressurizer water level limiting condition of operation from less than or equal to 92% of indicated span to less than or equal to 75% of indicated span, which requires a change to TS 3.4.3, is also proposed with this change.

Many of the analyses supporting the proposed license amendment were performed using analytical methods previously reviewed and approved by the NRC for use at HNP and are included in HNP TS 6.9.1.6, Core Operating Limits Report. Duke Energy has performed a new analysis for the overpressure evaluation of the Final Safety Analysis Report (FSAR), Section 15.2.3 turbine trip event, which is described in Attachment B. The new turbine trip analysis is based on analytical methods previously reviewed and approved by the NRC for use at other Duke Energy facilities, but not for HNP. Therefore, Duke Energy also requests approval of the PROPRIETARY INFORMATION - WITHHOLD UNDER 10 CFR 2.390 UPON REMOVAL OF ATTACHMENT C THIS LETTER IS UNCONTROLLED

PROPRIETARY INFORMATION - WITHHOLD UNDER 10 CFR 2.390 UPON REMOVAL OF ATTACHMENT C THIS LETTER IS UNCONTROLLED U.S. Nuclear Regulatory Commission Page2 HNP-15-038 new FSAR Section 15.2.3 turbine trip analysis presented in Attachment B, which uses analytical methods described in Attachment C.

An evaluation of the proposed changes included in this proposed license amendment is provided in Enclosure 1. The proposed TS changes are provided in Enclosure 2. The revised TS changes are provided in Enclosure 3. The proposed TS Bases changes are provided in . Attachment A provides a section of HNP-l/INST-1010, which pertains to the proposed pressurizer water level high trip set point change described in Enclosure 1.

Attachment B provides the HNP turbine trip analysis to address the revised safety valve tolerances described in Enclosure 1. Attachment C provides the HNP turbine trip methodology qualification developed to support the proposed license amendment.

Information provided in Attachment C is proprietary to Duke Energy. Duke Energy requests that the NRC withhold this information in accordance with 10 CFR 2.390 as trade secrets and commercial or financial information. An affidavit is included (Enclosure 5) attesting to the proprietary nature of the information. A non-proprietary version of Attachment C is included in Attachment D.

Approval of the proposed amendment is requested by June 17, 2016. The amendment shall be implemented within 90 days following approval.

In accordance with 10 CFR 50.91, Duke Energy is notifying the State of North Carolina of this license amendment request by transmitting a copy of this letter to the designated State Official.

This document contains no new regulatory commitments.

Please refer any questions regarding this submittal to John Caves, HNP Regulatory Affairs Manager, at (919) 362-2406.

I declare under penalty of perjury that the foregoing is true and correct. Executed on December {=?- , 2015.

Sincerely, PROPRIETARY INFORMATION - WITHHOLD UNDER 10 CFR 2.390 UPON REMOVAL OF ATTACHMENT C THIS LETTER IS UNCONTROLLED

PROPRIETARY INFORMATION - WITHHOLD UNDER 10 CFR 2.390 UPON REMOVAL OF ATTACHMENT C THIS LETTER IS UNCONTROLLED U.S. Nuclear Regulatory Commission Page 2 HNP-15-038 new FSAR Section 15.2.3 turbine trip analysis presented in Attachment B, which uses analytical methods described in Attachment C.

An evaluation of the proposed changes included in this proposed license amendment is provided in Enclosure 1. The proposed TS changes are provided in Enclosure 2. The revised TS changes are provided in Enclosure 3. The proposed TS Bases changes are provided in . Attachment A provides a section of HNP-I/INST-1010, which pertains to the proposed pressurizer water level high trip set point change described in Enclosure 1.

Attachment B provides the HNP turbine trip analysis to address the revised safety valve tolerances described in Enclosure 1. Attachment C provides the HNP turbine trip methodology qualification developed to support the proposed license amendment.

Information provided in Attachment C is proprietary to Duke Energy. Duke Energy requests that the NRC withhold this information in accordance with 10 CFR 2.390 as trade secrets and commercial or financial information. An affidavit is included (Enclosure 5) attesting to the proprietary nature of the information. A non-proprietary version of Attachment C is included in Attachment D.

Approval of the proposed amendment is requested by June 17, 2016. The amendment shall be implemented within 90 days following approval.

In accordance with 10 CFR 50.91, Duke Energy is notifying the State of North Carolina of this license amendment request by transmitting a copy of this letter to the designated State Official.

This document contains no new regulatory commitments.

Please refer any questions regarding this submittal to John Caves, HNP Regulatory Affairs Manager, at (919) 362-2406.

I declare under penalty of perjury that the foregoing is true and correct. Executed on December , 2015.

Sincerely, Benjamin C. Waldrep PROPRIETARY INFORMATION - WITHHOLD UNDER 10 CFR 2.390 UPON REMOVAL OF ATTACHMENT C THIS LETTER IS UNCONTROLLED

PROPRIETARY INFORMATION - WITHHOLD UNDER 10 CFR 2.390 UPON REMOVAL OF ATTACHMENT C THIS LETTER IS UNCONTROLLED U.S. Nuclear Regulatory Commission Page 3 HNP-15-038

Enclosures:

1. Evaluation of the Proposed Change

2. Proposed Technical Specification Changes

3. Revised Technical Specification Changes

4. Proposed Technical Specification Bases Changes

5. Affidavit Attachments:

A. HNP-I/INST-1010, Evaluation of RTS/ESFAS Tech Spec Related Setpoints, Allowable Values, and Uncertainties, Table 3.8 B. Harris Turbine Trip Analysis to Address Revised Safety Valve Tolerances C. Harris Turbine Trip Methodology Qualification (Proprietary)

D. Harris Turbine Trip Methodology Qualification (Redacted) cc: Mr. J. D. Austin, NRC Sr. Resident Inspector, HNP Mr. W. L. Cox, III, Section Chief, N.C. DHSR Ms. M. Barillas, NRC Project Manager, HNP NRC Regional Administrator, Region II PROPRIETARY INFORMATION - WITHHOLD UNDER 10 CFR 2.390 UPON REMOVAL OF ATTACHMENT C THIS LETTER IS UNCONTROLLED

U.S. Nuclear Regulatory Commission Serial HNP-15-038 SERIAL HNP-15-038 ENCLOSURE 1 EVALUATION OF PROPOSED CHANGE SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 DOCKET NO. 50-400 RENEWED LICENSE NUMBER NPF-63

U.S. Nuclear Regulatory Commission Serial HNP-15-038 Evaluation of the Proposed Change 1.0

SUMMARY

DESCRIPTION 2.0 DETAILED DESCRIPTION

3.0 TECHNICAL EVALUATION

3.1 System Description

3.2 Description of the Changes 3.3 Impact to Transient and Accident Analyses 3.4 Conclusions

4. REGULATORY EVALUATION 4.1 Applicable Regulatory Requirements 4.2 Precedent 4.3 No Significant Hazards Consideration Determination 4.4 Conclusions

5. ENVIRONMENTAL CONSIDERATION

6. REFERENCES

U.S. Nuclear Regulatory Commission Page 1 of 20 Serial HNP-15-038

1.

SUMMARY

DESCRIPTION In accordance with the provisions of 10 CFR 50.90, Duke Energy Progress, Inc. (Duke Energy), is proposing a change to Shearon Harris Nuclear Power Plant, Unit 1 (HNP) Technical Specifications (TS) to change the as-found lift setting tolerance for the main steam line code safety valves (MSSVs) from +/- 1% to

+/- 3%, which requires a change to TS 3.7.1.1, Table 3.7-2, Steam Line Safety Valves Per Loop. The change will provide additional operational flexibility and will preserve the capabilities of the MSSVs to perform their safety function. To support the MSSV setpoint tolerance change, changes are required to TS 2.2.1, Table 2.2-1, Reactor Trip System Instrumentation Trip Setpoints. Specifically, the instrumentation trip setpoint for pressurizer water level-high percentage of the instrument span is reduced from 92% to 87%. Further, the allowable value of the instrument span is requested to be reduced from 93.5% to 88.5%. A change to reduce the maximum pressurizer water level limiting condition of operation (LCO) from less than or equal to 92% of indicated span to less than or equal to 75% of indicated span, which requires a change to TS 3.4.3 is also proposed with this change. Duke Energy also requests approval of a new Final Safety Analysis Report (FSAR), Section 15.2.3 turbine trip analysis, which has been performed to assess the impact of the requested changes on the transient and accident analyses included in the HNP FSAR.

2. DETAILED DESCRIPTION The proposed change would revise the following HNP TS:

TS 3.7.1.1 LCO requires that all main steam line code safety valves associated with each steam generator shall be operable with lift settings in modes 1, 2, and 3 as specified in Table 3.7-2. Duke Energy requests a revision to the MSSV as-found lift setpoint tolerance from +/- 1% to +/- 3% described in TS Table 3.7-2.

TS 2.2.1 requires that reactor trip system instrumentation and interlock setpoints to be set consistent with the trip setpoint values shown in TS Table 2.2-1 in the modes listed in TS Table 3.3-1.

Duke Energy requests a revision to the nominal reactor trip setpoint on pressurizer water level high functional unit from 92% of the instrument span to 87% of the instrument span (Item 11 in TS Table 2.2-

1) and to apply the existing notes 7 and 8 to the pressurizer water level high functional unit per Technical Specifications Task Force Traveler (TSTF)-493, "Clarify Application of Setpoint Methodology for LSSS Functions, Revision 4 (Agency-wide Documents Access and Management System (ADAMS)

Accession No. ML100060064). The existing notes 7 and 8 correspond to TSTF-493, option A, notes 1 and 2, respectively and are shown in the markup of TS Table 2.2-1 provided in Enclosure 2. Duke Energy also requests a revision to the allowable value for Item 11 in TS Table 2.2-1 from 93.5% of the instrument span to 88.5% of the instrument span.

The current TS Table 2.2-1 Bases description accounts for notes 7 and 8 that require verifying both the trip set point setting as-found and as-left values during surveillance testing. The LAR proposes to add Notes 7 and 8 to the pressurizer water level high functional unit and therefore the TS Bases changes are to add this functional unit to the existing functional units that are associated with notes 7 and 8. The proposed TS Bases changes are shown in Enclosure 4.

TS 3.4.3 LCO requires that the pressurizer be operable with a water level of less than or equal to 92% of indicated span, and at least two groups of pressurizer heaters each having a capacity of at least 125 kilowatts (kW) in modes 1, 2, and 3. Due to the initial condition assumptions utilized in the new FSAR, Section 15.2.3 turbine trip analysis in Attachment B, the TS 3.4.3 LCO for pressurizer level must be reduced from 92 to 75% of indicated span. Therefore, Duke Energy requests a revision to the LCO from "less than or equal to 92% of indicated span" to "less than or equal to 75% of indicated span."

U.S. Nuclear Regulatory Commission Page 2 of 20 Serial HNP-15-038 The current TS 3.4.3 Bases description for pressurizer operability is not up to date with the Improved Technical Specifications equivalent text contained in NUREG-1431, Standard Technical Specifications Westinghouse Plants, Revision 4, Volume 2, Bases, dated April 2012 (ADAMS Accession No. ML12100A228). The text requires modification so that the value for pressurizer level (TS 3.4.3) is tied to the analysis rather than the general statement that currently exists. The proposed language creates a tie between the analysis for FSAR Section 15.2.3 and the sensitivity completed to demonstrate the adequacy of pressurizer safety valve (PSV) sizing (both of these items are presented in Attachment B). The recommended update to TS Bases Section 3.4.3 is provided in Enclosure 4.

The change to the maximum pressurizer water level limit in TS 3.4.3 is consistent with the initial level assumed in the FSAR, Section 15.2.3 turbine trip overpressure analysis presented in this submittal. The initial pressurizer level assumed for all other FSAR events will continue to be established in accordance with the applicable analysis methodology and are not associated with this LCO.

In combination with a main steam line code safety valve tolerance of +/- 3% and a high pressurizer level reactor trip setpoint of 87%, which are proposed in this submittal, a pressurizer water level limit of 75% of indicated span provides acceptable operational flexibility while minimizing the consequences of the FSAR, Section 15.2.3 turbine trip overpressure analysis presented in this LAR. The proposed TS pages illustrating the proposed changes are provided in Enclosures 2 and 3.

Many of the analyses supporting the LAR were performed using analytical methods previously reviewed and approved by the NRC for use at HNP. To support an assessment of the impact of the proposed changes to the transient and accident analyses included in the HNP FSAR, however, Duke Energy performed a new analysis for the FSAR 15.2.3 turbine trip overpressure evaluation, which is summarized in Attachment B.

The applied methodology is described in Attachment C and is based on analytical methods previously reviewed and approved by the NRC for use at other Duke Energy facilities (McGuire Nuclear Station (MNS) and Catawba Nuclear Station (CNS)). Therefore, Duke Energy requests approval of the new FSAR, Section 15.2.3 turbine trip analysis presented in Attachment B.

The HNP FSAR, Section 15.2.3 turbine trip analysis is reanalyzed to evaluate changes to the primary and secondary system safety valve tolerances. Two cases are analyzed for this event: one challenging the primary overpressurization criterion and one challenging the secondary system overpressurization criterion.

In addition, a sensitivity case is performed to confirm the requirements of NUREG-0800, Standard Review Plan for the Review of Safety Analysis Report for Nuclear Power Plants (Reference 12), Chapter 5.2.2 -

Overpressure Protection, continue to be satisfied. An evaluation of the Departure from Nucleate Boiling (DNB) analysis is also performed.

The evaluations performed to assess the impact to the HNP licensing analyses from the requested change and the results are summarized below:

The existing overtemperature delta-T (OTT) and overpressure delta-T (OPT) trip equations are confirmed to remain effective with modified MSSV setpoint tolerances.

The existing Core Safety Limit Lines (CSLL) as displayed in TS Figure 2.1-1 are determined to remain effective.

The FSAR, Section 15.2.3 turbine trip overpressure evaluation is reanalyzed as described in Attachments B and C. The peak pressures for the Reactor Coolant System (RCS) and Main Steam (MS) systems continue to result from the turbine trip event. Therefore, all limiting event determinations previously described in the HNP FSAR remain valid.

DNB statepoints for turbine trip are determined to be insensitive to the change in the MSSV setpoint tolerance. Duke will continue to rely on AREVA to calculate the Minimum Departure from Nucleate Boiling Ratio (MDNBR) result for FSAR Section 15.2.3. AREVA methods require that the MDNBR be recalculated or dispositioned based on cycle-to-cycle variations in the limiting axial power distribution.

U.S. Nuclear Regulatory Commission Page 3 of 20 Serial HNP-15-038

Other non-Loss of Coolant Accident (LOCA) events are discussed in Table 2 in Section 3.3 of this Enclosure, and were found to be either 1) bounded by the current analysis of record (AOR), 2) bounded by the turbine trip event regarding overpressure, or 3) the Fuel Centerline Melt (FCM) and/or MDNBR limits are not affected.

The results of Large Break LOCA (FSAR 15.6.5.2) are unaffected by the change to MSSV tolerances and the associated reduction in credited auxiliary feedwater flow (AFW) flow.

The results of Small Break LOCA (FSAR 15.6.5.3) were evaluated. The impact of the MSSV setpoint tolerance and AFW flow rate changes were calculated to be +32°F on peak cladding temperature (PCT),

as shown in Table 3 of this Enclosure.

The results of Anticipated Transients Without Scram (ATWS) were not impacted by the changes.

The change to the MSSV tolerances do not result in an increase in the radiological doses for any design basis accident.

The Main Steam Line Break (MSLB) and LOCA mass and energy release data and downstream containment transients are not impacted by the requested change.

In Reference 13, the NRC endorsed a licensee commitment for HNP to perform future replacement safety analyses using the thermal-hydraulic analysis methodology EMF-2310 (Reference 9). Duke Energy will continue to maintain this commitment; however, upon approval of the requested license amendment, the FSAR, Section 15.2.3 turbine trip overpressure analysis will be performed using the thermal-hydraulic analysis methodology discussed in Attachments B and C.

3.0 TECHNICAL EVALUATION

3.1 System

Description:

The main steam system (MSS) conveys steam produced in the three steam generators to the main turbine.

The MSS also supplies steam to the moistureseparator reheaters, the auxiliary feed pump turbine, main turbine shaft gland seals, and auxiliary steam system. The MSS piping from the steam generators up to the main steam isolation valves (MSIVs) are designed and fabricated to the requirements of American Society of Mechanical Engineers (ASME) Boiler and Pressure (B&PV) Vessel Code,Section III, Class 2. The safetyrelated portion of the MSS is designed for sustained loads (pressure, dead weight), thermal expansion, occasional loads, and jet impingement from outside the system. Loads imposed by operational transient conditions, including the effects of steam hammer, are also considered in the design. Each MS line from the steam generator is provided with five spring loaded type MSSVs that meet the requirements of ASME B&PV Code,Section III, Class 2 and Seismic Category I. The safety valves are designed to attain full lift at a pressure no greater than 3% above their set pressure, while maintaining the steam generator below the maximum allowable of 10% above the steam generator design pressure. Table 1 shown below provides summary of parameters for MSSVs, which are described in the HNP design specification.

U.S. Nuclear Regulatory Commission Page 4 of 20 Serial HNP-15-038 Table 1: Main Steam Line Code Safety Valve Parameters Valve Number Set Design Design Accumulation Steam Generator Pressure Pressure Temperature

(+/- 1%)

A B C 1MS43 1MS44 1MS45 1170 psig 1185 psig 600 °F 3%

1MS46 1MS47 1MS48 1185 psig 1185 psig 600 °F 3%

1MS49 1MS50 1MS51 1200 psig 1185 psig 600 °F 3%

1MS52 1MS53 1MS54 1215 psig 1185 psig 600 °F 3%

1MS55 1MS56 1MS57 1230 psig 1185 psig 600 °F 3%

HNP FSAR, Section 5.2.2 discusses the requirements for overpressure protection. Per this section of the FSAR, RCS overpressure protection during normal plant operation is accomplished by the utilization of PSVs along with the reactor protection system and associated equipment. Combinations of these systems provide compliance with the overpressure requirements of the ASME B&PV Code,Section III, paragraph NB7300 and NC7300 for pressurized water reactor systems. Additionally, from ASME B&PV Code,Section III, Article NC-7411: The total rated relieving capacity of the pressure relief devices intended for overpressure protection of the system whose components are within the scope of this Subsection shall be sufficient to prevent a rise in pressure of more than 10% above system design pressure at design temperature within the protected boundary of the system under any pressure transients anticipated to arise.

Overpressure protection for the shell side of the steam generators and the main steam line up to the main steam isolation valves is provided by the 15 steam generator safety valves (MSSVs), 5 on each main steam line. The steam generator safety valve capacity is based on providing enough relief to remove 105 percent of the rated Nuclear Steam Supply System (NSSS) steam flow. This must be done by limiting main steam system pressure to less than 110 percent of the steam generator shell side design pressure.

The reactor protection system provides an automatic reactor trip function to the reactor trip breakers to protect against unsafe and improper reactor operation during steady state and transient power operation and to provide initiating signals to mitigate the consequences of faulted conditions. The system uses input signals including neutron flux, RCS temperature, RCS Flow, pressurizer pressure, pressurizer level, steam generator level, reactor coolant pump under-voltage and under-frequency, turbine trip signals, and safety injection to provide a reactor trip signal.

The pressurizer water level - high trip is designed to prevent rapid thermal expansion of the reactor coolant from filling the pressurizer. This trip function ensures a reactor trip is actuated prior to the pressurizer becoming water solid. Satisfaction of this criterion is demonstrated in the safety analyses performed in Attachment B, in which a key criterion was to prevent water from reaching the pressurizer relief valves. Two-out-of-three logic is used for this trip. Isolated outputs from the pressurizer level protection channels are used as inputs to the pressurizer level control system. A level control failure could fill or empty the pressurizer at a slow rate; however, a level control failure would not actuate the safety valves because the RCS High Pressure reactor trip setpoint is set below the safety valves set pressure. With the slow rate of

U.S. Nuclear Regulatory Commission Page 5 of 20 Serial HNP-15-038 charging available, overshoot in pressure before the reactor trip occurs is much less than the difference between the reactor trip and safety valves set pressure. Therefore, a pressurizer level control system failure does not require reactor protection system actuation. For a pressurizer level control system failure which tends to empty the pressurizer, a signal of low level from either of two independent level control channels isolates letdown, thus preventing further loss of coolant. In addition, ample time and alarms exist for operator action.

3.2 Description of the Changes The proposed change to the HNP TS is an increase in the as-found lift setting tolerance for the main steam line code safety valves from +/- 1% to +/- 3%, which requires a change to TS 3.7.1.1, Table 3.7-2, Steam Line Safety Valves Per Loop. To support the MSSV setpoint tolerance change, changes are required to TS 2.2.1, Table 2.2-1, Reactor Trip System Instrumentation Trip Setpoints. Specifically, the instrumentation trip setpoint for pressurizer water level-high percentage of the instrument span is reduced from 92% to 87%. Further, the allowable value of the instrument span is requested to be reduced from 93.5% to 88.5%. A change to reduce the maximum pressurizer water level LCO from less than or equal to 92% of indicated span to less than or equal to 75% of indicated span is also proposed with this change, which requires a change to TS 3.4.3.

To support an assessment of the impact of the proposed changes to the transient and accident analyses included in the HNP FSAR, Duke Energy performed a new analysis for the overpressure evaluation of FSAR 15.2.3 using the methodology described in Attachments B and C. The methodology is based on Duke Energy Topical Reports DPC-NE-3000-PA and DPC-NE-3002-A (References 1 and 2), which have been previously reviewed and approved by the NRC for McGuire Nuclear Station (MNS) and Catawba Nuclear Station (CNS). Therefore, Duke Energy also requests approval of the new FSAR 15.2.3 turbine trip analysis presented in Attachment B, which uses analytical methods described in Attachment C.

The FSAR 15.2.3 overpressure analysis summarized in Attachment B shows that the resulting peak pressure in the main steam system is within the limit of 1320 pounds per square inch absolute (psia) with the proposed changes, which is 110% of peak design pressure for main steam system of 1200 psia (or 1185 psi gauge). Therefore, the requirements of ASME B&PV Code,Section III, paragraph NC7300 are met. Additionally, it may be concluded that due to the proposed changes in the MSSV as-found set pressure tolerance from +/- 1% to +/- 3%, there will be no adverse impact to the current pipe stress analysis and the main steam system piping will remain code qualified.

Based on results from the NUREG-0800, Standard Review Plan (SRP) Chapter 5.2.2 Section II.3.B.iii sensitivity case performed in Attachment B, the reactor trip system (RTS) pressurizer water level - high trip setpoint will be reduced from 92% to 87% indicated span, as discussed above. Due to the initial condition assumptions utilized for this sensitivity case in Attachment B, it is required that the TS 3.4.3 LCO for pressurizer level be reduced from 92% to 75% of indicated span for mode 1. The same limit of 75% of indicated span will be applied to modes 2 and 3 also. The level control function will remain as-is, normally controlled between 25% and 60% of its indicated span as Tavg varies from 557°F to 588.8°F.

Since the proposed LAR impacts a reactor trip system function, TSTF-493 needs to be addressed and satisfied for this change. The procedure used at HNP for engineering instrument setpoints describes the method for satisfying the TSTF-493 criteria.

HNP-I/INST-1010, Evaluation of RTS/ESFAS Tech Spec Related Setpoints, Allowable Values, and Uncertainties, Table 3-8 describes the methodology used for the proposed pressurizer water level - high trip setpoint change and the calculation results. This section of the calculation is provided in Attachment A of this submittal as it pertains to the requested change. To support the TSTF-493 requirements, Table 3-8 is expanded to include supplemental "as-found" and as-left" tolerance criteria. When surveillance test results exceed these tolerances, specific additional review actions are required on the part of the technicians,

U.S. Nuclear Regulatory Commission Page 6 of 20 Serial HNP-15-038 operations staff and engineering prior to and following returning the affected channels to service. The intent is to ensure that during testing the instruments and loop perform in accordance with "expected" capability rather than more simply within allowable values, which can include additional margin. These actions are described in an engineering procedure and are invoked by existing TS Table 2.2-1, Notes 7 and 8 (which are currently only applicable to the Power Range Neutron Flux RTS functions).

Notes 7 and 8 were added to TS Table 2.2-1 under License Amendment 139 approved on May 30, 2012 (per Reference 6) that require verifying both trip setpoint setting as-found and as-left values during surveillance testing. In accordance with 10 CFR 50.36, these functions are Limiting Safety System Settings.

The LAR proposes to add notes 7 and 8 to the functional unit pressurizer water level-high. The existing Notes 7 and 8 notes correspond to TSTF-493 option A, notes 1 and 2, respectively. HNP TS Bases for TS Table 2.2-1, which is provided in Enclosure 4, states, adding test requirements ensures that instruments will function as required to initiate protective systems or actuate mitigating systems at the point assumed in the applicable safety analysis. These notes address NRC staff concerns with TS Allowable Values.

Specifically, calculated Allowable Values may be non-conservative depending upon the evaluation of instrument performance history, and the as-left requirements of the calibration procedures could have an adverse effect on equipment operability. In addition, using Allowable Values as the limiting setting for assessing instrument channel operability may not be fully in compliance with the intent of 10 CFR 50.36, and the existing surveillance requirements would not provide adequate assurance that instruments will always actuate safety functions at the point assumed in the applicable safety analysis. In the HNP Technical Specifications, the term Trip Setpoint is analogous to Nominal Trip Setpoint (NTSP) in TSTF-493.

3.3 Impact to Transient and Accident Analyses Evaluations presented in this section support an increase in the MSSV setpoint tolerance from +/-1% to +/-3%.

A consequence of increased MSSV setpoint tolerance is a reduction of the credited auxiliary feedwater (AFW) flow in the safety analyses at the lowest lifting MSSV setpoint pressure plus tolerance. The change to AFW flow in certain accident and transient analyses is a reduction from 390 gallons per minute (gpm) to 374 gpm. This reduction has been considered in the evaluations presented herein. Many of these evaluations also support an increase in the PSV setpoint tolerance from +/-1% to +/-3%; however, this LAR does not request a change to the PSV setpoint tolerances in TS 3.4.2.1 and TS 3.4.2.2. Evaluations that consider both an MSSV and PSV setpoint tolerance change from +/-1% to +/-3% bound the requested change for only the MSSV setpoint tolerance increase. Finally, the requested change to the maximum pressurizer water level limit in TS 3.4.3 is consistent with the initial level assumed in the FSAR Section 15.2.3 turbine trip overpressure analysis presented in Attachment B. The initial pressurizer level assumed for all other FSAR events will continue to be established in accordance with the applicable analysis methodology and are not associated with this LCO or affected by the requested change to TS 3.4.3.

3.3.1 OTT, OPT and Core Safety Limit Lines The margin in the OTT and OPT reactor trip equations has been re-evaluated using the methodology in Reference 3. The evaluation determined that the equations continue to be effective with positive trip margin as a result of the change to MSSV setpoint tolerances. Evaluations of the Core Safety Limit Lines (CSLL) in TS Figure 2.1-1 using the Reference 3 methodology determined that the curves presented positive margin to MDNBR and bulk saturated conditions in the hot leg.

3.3.2 FSAR Chapter 15 Analysis Impact Summary The HNP FSAR, Chapter 15 events are listed below in Table 2, which summarizes the impact of the requested changes on individual FSAR analyses. The following acronyms are used within Table 2:

U.S. Nuclear Regulatory Commission Page 7 of 20 Serial HNP-15-038 American Nuclear Society (ANS), Thermal Hydraulic= (T/H), Anticipated Operational Occurrence (AOO),

and Postulated Accident (PA).

Table 2: Summary of FSAR Chapter 15 Event Dispositions FSAR ANS NRC-Approved Section Event Description Condition T/H Methodology Disposition 15.1.1 Feedwater System II (AOO) Reference 8 No impact to FSAR evaluation.

Malfunctions that Result in Event bounded by FSAR a Decrease in Feedwater 15.1.3.

Temperature 15.1.2 Feedwater System II (AOO) Reference 8 System transient bounded by Malfunctions that Result in AOR.

an Increase In Feedwater Flow 15.1.3 Excessive Increase in II (AOO) Reference 8 System transient bounded by Secondary Steam Flow AOR.

15.1.4 Inadvertent Opening of a II (AOO) Reference 8 At Power: Bounded by FSAR Steam Generator Relief or 15.1.3 Safety Valve After reactor trip: bounded by FSAR 15.1.5 15.1.5 Steam System Piping IV (PA) Reference 10 System transient bounded by Failure AOR.

15.2.1 Not applicable (BWR event) 15.2.2 Loss of External Electrical II (AOO) Reference 8 No impact to FSAR evaluation.

Load Event bounded by FSAR 15.2.3.

15.2.3 Turbine Trip II (AOO) See Transient re-analyzed by Duke Attachments Energy. Refer to Section 3.3.3.

B&C 15.2.4 Inadvertent Closure of II (AOO) Reference 8 No impact to FSAR evaluation.

Main Steam Isolation Event bounded by FSAR Valves 15.2.3.

15.2.5 Loss of Condenser II (AOO) Reference 8 No impact to FSAR evaluation.

Vacuum and Other Events Event bounded by FSAR Resulting in Turbine Trip 15.2.3.

U.S. Nuclear Regulatory Commission Page 8 of 20 Serial HNP-15-038 FSAR ANS NRC-Approved Event Description Disposition Section Condition T/H Methodology 15.2.6 Loss of Non-Emergency II (AOO) Reference 8 SG dryout bounded by FSAR Alternating Current (AC) 15.2.7 (see below).

Power to the Station Secondary side overpressure limit Auxiliaries is not challenged and is therefore bounded by FSAR 15.2.3.

DNB limits are bounded by FSAR 15.3.2 since FSAR 15.3.2 satisfies the ANS Condition II criteria.

15.2.7 Loss of Normal II (AOO) Reference 8 Evaluation verified adequate Feedwater Flow heat removal capacity conclusion in AOR.

Secondary side overpressure limit is not challenged and is therefore bounded by FSAR 15.2.3.

The DNB limits are bounded by FSAR 15.3.2 since FSAR 15.3.2 satisfies ANS Condition II criteria.

15.2.8 Feedwater System IV (PA) Reference 8 Overpressure limits are not Pipe Break challenged and are therefore bounded by FSAR 15.2.3.

15.3.1 Partial Loss of Forced II (AOO) Reference 8 No impact to FSAR evaluation.

Reactor Coolant Flow Bounded by FSAR15.3.2 since FSAR 15.3.2 satisfies ANS Condition II criteria.

15.3.2 Complete Loss of III (PA) Reference 8 System transient bounded by Forced Reactor AOR. Peak RCS pressure limit is Coolant Flow not challenged and is therefore bounded by FSAR 15.2.3.

U.S. Nuclear Regulatory Commission Page 9 of 20 Serial HNP-15-038 FSAR ANS NRC-Approved Event Description Disposition Section Condition T/H Methodology 15.3.3 Reactor Coolant Pump IV (PA) Reference 8 MDNBR bounded by AOR.

Shaft Seizure (Locked The peak RCS and secondary Rotor) pressures are not bounded by the AOR, but are bounded by the more limiting FSAR 15.2.3 Condition II event.

15.3.4 Reactor Coolant Pump IV (PA) Reference 8 No impact to FSAR Shaft Break evaluation. Event bounded by FSAR 15.3.3.

15.4.1 Uncontrolled Rod Cluster II (AOO) Reference 8 System transient bounded by Control Assembly Bank AOR.

Withdrawal from a Subcritical or Low Power Startup Condition 15.4.2 Uncontrolled Rod Cluster II (AOO) Reference 9 Secondary side overpressure Control Assembly Bank limit is not challenged and is Withdrawal at Power therefore bounded by FSAR 15.2.3. System transient for Linear Heat Rate (LHR) and DNB bounded by AOR.

15.4.3.1 Dropped Full Length Rod II (AOO) Reference 3 System transient bounded by Cluster Control Assembly AOR.

(RCCA) or RCCA Bank 15.4.3.2 Withdrawal of Single Full III (PA) Reference 9 Secondary side overpressure Length RCCA limit is not challenged and is therefore bounded by FSAR 15.2.3. System transient for LHR and DNB bounded by AOR.

15.4.3.3 Statically Misaligned II (AOO) Reference 8 No impact to FSAR RCCA or Bank evaluation.

15.4.4 Startup of an Inactive II (AOO) Reference 8 No impact to FSAR evaluation.

Reactor Coolant Pump at Bounded by FSAR 15.4.1.

an Incorrect Temperature

U.S. Nuclear Regulatory Commission Page 10 of 20 Serial HNP-15-038 FSAR ANS NRC-Approved Section Event Description Condition T/H Methodology Disposition 15.4.5 Not applicable (BWR event) 15.4.6 Chemical and Volume II (AOO) N/A No impact to FSAR evaluation.

Control System Bounded by FSAR 15.4.2.

Malfunction that Results in a Decrease in the Boron Concentration in the Reactor Coolant 15.4.7 Inadvertent Loading and III (PA) N/A This event does not involve an Operation of a Fuel NSSS transient.

Assembly in an Improper Position 15.4.8 Spectrum of Rod Cluster IV (PA) References 8 & System transient bounded by Control Assembly 11 AOR. Peak RCS pressure limit Ejection Accidents is not challenged and is therefore bounded by FSAR 15.2.3.

15.4.9 Not applicable (BWR event) 15.5.1 Inadvertent Operation of II (AOO) Reference 8 System transient bounded by the Emergency Core AOR.

Cooling System During Power Operation 15.5.2 Chemical and Volume II (AOO) Reference 8 No impact to FSAR Control System evaluation. Bounded by Malfunction that FSAR 15.4.6 and 15.5.1.

Increases Reactor Coolant Inventory 15.6.1 Inadvertent Opening of a II (AOO) Reference 8 System transient bounded by Pressurizer Safety or AOR for short term. Effects of Power Operated Relief the changes on the long term Valve response to the event were found to be negligible.

15.6.2 Break in Instrument Line II (AOO) N/A Does not involve an NSSS or Other Line from transient.

Reactor Coolant Pressure Boundary that Penetrate Containment

U.S. Nuclear Regulatory Commission Page 11 of 20 Serial HNP-15-038 Enclosure 1 FSAR ANS NRC-Approved Section Event Description Condition T/H Methodology Disposition 15.6.3 Steam Generator Tube IV (PA) Reference 7 No impact. Refer to Section 3.3.4 for Rupture additional information.

15.6.4 Not applicable (BWR event) 15.6.5 Small Break Loss of Coolant III (PA) Reference 5 See Section 3.3.5 Accidents 15.6.5 Large Break Loss of Coolant IV (PA) Reference 4 See Section 3.3.5 Accidents 15.7.1 Radioactive Waste Gas III (PA) N/A This event does not involve an NSSS System Leak or Failure transient.

15.7.2 Liquid Waste System Leak or III (PA) N/A This event does not involve an NSSS Failure transient.

15.7.3 Postulated Radioactive III (PA) N/A This event does not involve an NSSS Releases Due to Liquid Tank transient.

Failure 15.7.4 Design Basis Fuel Handling IV (PA) N/A This event does not involve an NSSS Accidents transient.

15.7.5 Spent Fuel Cask Drop III (PA) N/A This event does not involve an NSSS Accidents transient.

15.8 Anticipated Transients N/A N/A No impact. See Section 3.3.6 for more Without Scram information.

N/A Radiological Consequences N/A N/A No impact. See Section 3.3.7 for more information.

N/A Control Room Habitability N/A N/A No impact. See Section 3.3.7 for more information.

N/A Alternate Source Term N/A N/A No impact. See Section 3.3.7 for more information.

3.3.3 FSAR Section 15.2.3 Turbine Trip Overpressure Analysis The FSAR, Section 15.2.3 turbine trip overpressure analysis of the primary and secondary systems has been re-evaluated using a methodology based on References 1 and 2. Attachment B provides the analysis results and associated method, and Attachment C provides details on the method development and benchmarking.

The analysis considers all PSVs and MSSVs having a setpoint tolerance of +3%, a pressurizer water level -

High reactor trip setpoint of 95% (87% requested value for TS Table 2.2-1 plus 8% Allowance), and an initial

U.S. Nuclear Regulatory Commission Page 12 of 20 Serial HNP-15-038 Enclosure 1 pressurizer level of 75% plus uncertainty. While this LAR does not request a change to the PSV setpoint tolerance, increasing the PSV tolerance in the analysis to +3% is conservative because it delays PSV actuation and yields a higher peak pressure in the primary system overpressure analysis.

Results from the analysis show peak RCS pressure is within the limit of 2750 psia, and peak MS pressure is within the limit of 1320 psia. The new primary and secondary system overpressure results are similar to, or bound the values from the previous FSAR, Section 15.2.3 turbine trip overpressure analysis of record.

Therefore, the FSAR, Section 15.2.3 turbine trip overpressure results continue to bound the overpressurization results for other ANS Condition II, III, and IV events.

Attachment B also presents a sensitivity case to satisfy the requirements of NUREG-0800, Standard Review Plan (SRP) Chapter 5.2.2 Section II.3.B.iii (Reference 12). This case assumes a reactor trip on the second safety-grade trip from the reactor protection system. Results show that the primary system pressure remains below the 110% design pressure limit and demonstrate that the design and sizing of the PSVs meet the overpressure design criterion of SRP Chapter 5.2.2.

3.3.4 Steam Generator Tube Rupture The steam generator tube rupture (SGTR) assume steam generator power operated relief valves (SG PORVs) are available on all generators to control pressure below the lowest MSSV lift setpoint. For dose evaluations, a SG PORV is assumed stuck open on the affected generator. For SG overfill considerations, the method in Reference 7 is employed and AFW flow is maximized. Therefore, since the MSSVs do not operate in these analyses, and because AFW flow is maximized for SG overfill evaluations, the SGTR results are not adversely impacted by the requested change in MSSV setpoint tolerances and the associated reduction in AFW flow at the lowest lifting MSSV setpoint plus tolerance.

3.3.5 Loss of Coolant Accident (LOCA)

The FSAR 15.6.5 large break LOCA (LBLOCA) analysis of record methodology is described in Reference 4.

The LBLOCA event involves a rapid depressurization of the RCS below the pressure of the secondary system pressure. Therefore, the MSSVs are not challenged and the LBLOCA is not affected by the MSSV setpoint tolerance change.

Small Break LOCA (SBLOCA) transients can be affected by the requested MSSV tolerance change. The SBLOCA analysis of record is prepared using the methodology described in Reference 5.

In the SBLOCA transients, secondary pressure rises to the MSSV setpoint upon reactor/turbine trip and remains there until primary phase change at the break occurs with a commensurate increase in energy release from the primary system. Early in a SBLOCA event, an increase in the MSSV setpoint tolerance can affect the energy balance during the transient because it results in a secondary heat sink temperature change. The higher setpoints of the MSSVs cause less heat transfer from the primary system and higher primary pressure. This results in less high pressure safety injection flow into the system, an earlier core uncovery, and more extensive cladding heatup.

A sensitivity analysis that supports an MSSV setpoint tolerance of +/-3% was performed for a subset of SBLOCA cases from the AOR. The minimum AFW flow rate was evaluated at 374 gpm to reflect the minimum AFW flow rate at an increased steam generator pressure consistent with the increased MSSV tolerance. The Resultant SBLOCA PCT effective April 9, 2015, which was reported within Reference 14, is 1681°F. The estimated impact of this change on the SBLOCA analysis calculated peak cladding temperature is +32°F, with a new calculated PCT of 1713°F. This value is well within the acceptance criteria of 2200°F

U.S. Nuclear Regulatory Commission Page 13 of 20 Serial HNP-15-038 Enclosure 1 defined in 10 CFR 50.46(b)(1). Thus, the result is acceptable. The PCT changes and errors from the AOR are shown in Table 3 below.

Table 3: SBLOCA PCT Impact Analysis PCT (°F) PCT (°F) Year Comments ANP-3238 (AOR) Used at start of 1664 2013 Cycle 19 S-RELAP5 vapor AREVA CR 2012-8371 absorptivity

+17 2014 correlation correction Requested MSSV/AFW +32 2015 MSSV tolerance change change Total Delta +49 2015 Resulting PCT 1713 3.3.6 Anticipated Transients Without Scram The Anticipated Transient Without Scram (ATWS) analysis of record for HNP was created as part of the Measurement Uncertainty Recapture (MUR) project. The ATWS AOR is summarized in materials that accompanied the HNP MUR LAR, which was approved by the NRC in Reference 6. Since the analysis credits MSSV operation at the nominal pressure setpoints, there is no impact from changing the safety valve setpoint tolerance as requested. Additionally, since the MSSVs open at nominal setpoints in the analysis, there is no impact to the credited AFW flow.

3.3.7 Dose Analyses Many of the dose evaluations consider release from the steam generators through the MSSVs and the subsequent cool down using the steam generator power operated relief valves (SG PORVs).

Based on a review of the dose analyses, it is concluded that the offsite and onsite doses are not impacted by the change in the MSSV setpoint tolerance. The mass of steam released is a function of core decay heat and stored energy in the RCS. The change in AFW flow rate will have a negligible impact on the dose consequences. Therefore, the change to the MSSV setpoint tolerances do not result in an increase to the radiological doses for any design basis accident.

U.S. Nuclear Regulatory Commission Page 14 of 20 Serial HNP-15-038 3.3.8 Station Blackout (SBO)

The Station Blackout evaluation determines the total condensate storage tank (CST) consumption for the postulated SBO sequence. The calculation considers the core decay heat and sensible RCS heat that must be removed to reach the end state. The analysis uses an average enthalpy change for the various steam generator conditions considered, including initial RCS heat removal by steaming the SGs through the MSSVs. As such, the change in MSSV setpoint tolerance has a negligible effect on the total condensate requirements.

3.3.9 Containment Analysis of LOCA and MSLB Mass and Energy Release For MSLB mass and energy release analyses, the primary and secondary sides of the plant are depressurized during the associated peak containment pressure and temperature response. Therefore, there is no impact from the MSSV setpoint tolerance change. The MSLB evaluation produces the limiting short-term containment temperature profile.

For LBLOCA, the limiting containment pressure response is produced from the immediate pressure pulse following a double-ended hot leg break. The resulting peak containment pressure occurs well before AFW actuation; therefore, the limiting containment pressure response is not affected by the MSSV setpoint tolerance change, or the associated change in credited AFW flow at the lowest MSSV setpoint plus tolerance.

For the LOCA long-term containment response, the steam generators are a heat source for double-ended pump suction cold leg breaks where some break flow moves through the steam generators on the affected loop after exiting the core. For this condition, the steam generators lose heat to the primary side and there is no long-term challenge to the MSSVs and, consequently, the AFW flow rate used in the analysis remains applicable.

The limiting Environment Qualification (EQ) temperature envelope is a composite of the temperature profiles from LOCA and MSLB. Since there are no impacts to either the MSLB or the LOCA containment response, the EQ envelope is not affected.

3.3.10 Interfaces with Fuel Vendor to Incorporate New FSAR, Section 15.2.3 Turbine Trip Overpressure Analysis into Vendor Reload Methodology Incorporation of the new turbine trip overpressure analysis into the HNP licensing basis requires new interfaces between Duke Energy and AREVA to ensure the turbine trip analysis remains bounding for future cycles and to integrate the analysis into AREVAs NRC-approved reload methodology for HNP. These interfaces can be summarized in the following areas:

MDNBR Interface for FSAR, Section 15.2.3

Neutronic Physics Parameter Interface for FSAR, Section 15.2.3 While turbine trip is not a limiting transient for DNB (Cycle 20 has more than 15% DNB margin), Duke Energy has examined the impact of the MSSV setpoint tolerance change on the DNB response of the turbine trip event. As part of this examination, the core thermal-hydraulic conditions at the point of minimum DNBR were compared. A set of core thermal-hydraulic conditions for core heat flux, inlet fluid temperature, system pressure, and inlet flow at the point of MDNBR are defined as MDBNR statepoints. Very small changes were noted for these statepoints, and it is concluded that the existing statepoints calculated by AREVA continue to be applicable. Therefore, since the MDNBR statepoints are insensitive to the requested MSSV tolerance change, Duke Energy will continue to rely on the AREVA DNB analysis of record for

U.S. Nuclear Regulatory Commission Page 15 of 20 Serial HNP-15-038 FSAR, Section 15.2.3, as well as the cycle-to-cycle DNB assessments performed by AREVA per the existing HNP reload methodology.

Duke Energy will verify inputs used in the new FSAR, Section 15.2.3 turbine trip overpressure analysis continue to bound the plant configuration in future cycles, including the core physics parameters. Duke Energy will rely on the AREVA neutronic methods to determine cycle-specific physics parameters as part of the existing reload process. The methods used are those already licensed for HNP in TS Section 6.9.1.6.

Duke will confirm that the physics parameters assumed in the turbine trip overpressure analysis bound future cycle-specific values calculated by AREVA. This is acceptable because the thermal-hydraulic transient methods and neutronic methods are independent of each other, and the neutronic methods are credited to predict core performance during a cycles operation.

3.4 Conclusions Duke Energy is requesting a change to HNP TS Table 3.7-2 that supports increasing the as-found valve setting tolerance of the MSSVs from +/-1% to +/-3%. The increased valve setting tolerance for the MSSVs results in a reduction of the credited AFW flow in the applicable safety analyses at the lowest lifting MSSV setpoint pressure plus tolerance (reduced from 390 gpm to 374 gpm), which has been considered in the evaluations described below.

Duke Energy also requests a change to the pressurizer water level - high reactor trip setpoint in TS Table 2.2-1 and the maximum pressurizer water level LCO for TS 3.4.3. The requested changes are a pressurizer water level -high reactor trip setpoint of 87% and maximum pressurizer water level LCO of 75%. The change to the maximum pressurizer water level limit in TS 3.4.3 is consistent with the initial level assumed in the FSAR, Section 15.2.3, turbine trip overpressure analysis presented in this LAR. The initial pressurizer level assumed for all other FSAR events will continue to be established in accordance with the applicable analysis methodology and are not associated with the TS 3.4.3 LCO.

To support the LAR, Duke Energy performed a new analysis for the overpressure evaluation of FSAR, Section 15.2.3 turbine trip, using a methodology based on existing Duke Energy methods previously reviewed and approved by the NRC for MNS and CNS. The analysis considers MSSVs having a setpoint tolerance of +3%, a pressurizer water level - high reactor trip setpoint of 95% (87% requested value for TS Table 2.2-1 plus 8% Allowance), and an initial pressurizer level of 75% plus uncertainty. Results from the new turbine trip overpressure analysis show primary and secondary peak pressures remain below the acceptance criteria of 110% design pressure, and the results are similar to, or bound the values from the previous FSAR, Section 15.2.3 overpressure analysis of record. Therefore, the new turbine trip overpressure analysis continues to bound the overpressurization results for other ANS Condition II, III, and IV events. Additionally, a sensitivity case performed for the turbine trip event demonstrates that the design and sizing of the PSVs satisfy the overpressurization criteria of SRP Chapter 5.2.2. Interfaces between Duke Energy and AREVA will be established to ensure the turbine trip overpressure analysis remains bounding for future cycles.

The impact of the requested changes on the RTS has been evaluated and it has been determined that sufficient margin exists to demonstrate the system will continue to provide protection against and mitigation of accident and transient conditions, consistent with the underlying safety analyses.

Evaluation of the transient and accident analyses in the HNP FSAR and other supporting licensing analyses shows that, with exception of SBLOCA, all other events are

1. unaffected by the change,

2. the existing AOR continues to be bounding, or

U.S. Nuclear Regulatory Commission Page 16 of 20 Serial HNP-15-038

3. the event is bounded by the new turbine trip analysis for overpressure concerns.

Therefore, all limiting event determinations previously described in the HNP FSAR remain valid with the new FSAR, Section 15.2.3 turbine trip overpressure analysis. For SBLOCA, the impact of the MSSV setpoint tolerance and AFW flow rate changes are estimated to be +32°F on peak cladding temperature (PCT), with a new PCT estimate of 1713°F. With exception of FSAR, Section 15.2.3, turbine trip, the HNP licensing basis accident and transient analyses discussed herein, and all evaluations against the AORs for the requested change, are performed using existing TS 6.9.1.6 methodologies, where applicable, and no other new methodologies are being introduced to support the proposed TS change.

Finally, Duke Energy has made the determination that this amendment request involves a No Significant Hazards Consideration by applying the standards established by the NRC regulations in 10 CFR 50.92 in Section 4.0 of this Enclosure.

4. REGULATORY EVALUATION 4.1 Applicable Regulatory Requirements NUREG-0800, Standard Review Plan (SRP) Chapter 5.2.2 overpressure design criterion was evaluated for the proposed LAR and is addressed in the Attachment B turbine trip analysis. Results show primary system pressure remains below the 110% design pressure limit and demonstrate that the design and sizing of the PSVs meet the overpressure design criterion.

The safety valves are designed to attain full lift at a pressure no greater than 3% above their set pressure, while maintaining the steam generator below the maximum allowable of 10% above the steam generator design pressure. The resulting peak pressure in the main steam system due to the overpressure transient conditions must be within 110% of peak design pressure for the main steam system. The evaluation described in Section 3 concludes that the requirements of ASME B&PV Code,Section III, paragraph NC7300 are met with the proposed change of the as-found lift setting tolerance for main steam line code safety valves (MSSVs) from +/- 1% to +/- 3%.

10 CFR 50.36, Technical Specifications, paragraph (c)(1)(ii)(A) specifies, Where a limiting safety system setting is specified for a variable on which a safety limit has been placed, the setting must be so chosen that automatic protective action will correct the abnormal situation before a safety limit is exceeded.

4.2 Precedent HNP has previously implemented a setpoint-related TS change subject to TSTF-493 applicability. License Amendment 139 was approved on May 30, 2012 for changes to Table 2.2-1 functions 2, 3 and 4 (power range neutron flux) values in support of the Leading Edge Flow Meter Measurement Uncertainty Recapture Power Uprate (Reference 6). This amendment added the Table 2.2-1 footnotes 7 and 8, to which extended applicability is now proposed within this LAR. The associated TS Bases, Section 2.2.1 was also updated to describe the plant methodology for compliance with TSTF-493 requirements for applicable functions.

The following letters contain examples in which approval was granted to other pressurized water reactor (PWR) licensees for increasing the as-found MSSV lift setpoint tolerance allowed by TS: (1) By letter dated January 18, 1996 (ADAMS Accession Number ML012920591), the NRC issued an amendment to Millstone to increase the PSV and MSSV lift setpoint tolerance from +/-1 percent to +/-3 percent. (2) By letter dated June 8, 1995 (ADAMS Accession Number ML020840146), the NRC issued an amendment to Palisades to increase the PSV and MSSV lift setpoint tolerance from +/-1 percent to +/-3 percent.

U.S. Nuclear Regulatory Commission Page 17 of 20 Serial HNP-15-038 4.3 No Significant Hazards Consideration Determination Pursuant to 10 CFR 50.90, Duke Energy Progress, Inc. (Duke Energy) proposes a license amendment request (LAR) for the Technical Specifications (TS) for Harris Nuclear Plant, Unit 1 (HNP). The proposed change would revise the TS to relax an overly restrictive TS requirement by revising the main steam line code safety valves (MSSVs) as-found lift setting tolerance from +/- 1% to +/- 3%.

Duke Energy evaluated whether or not a significant hazards consideration (SHC) is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of Amendment," as discussed below.

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

The proposed TS changes allow for an increase in the as-found MSSV setpoint tolerance from +/-1%

to +/-3%. In addition, the proposed amendment request includes a conservative change to the reactor trip on high pressurizer level and makes TS 3.4.3 consistent with the initial pressurizer level used in the re-analysis of the HNP Final Safety Analysis Report (FSAR), Section 15.2.3 turbine trip overpressure event. The proposed changes do not alter the MSSV nominal lift setpoints. The proposed TS changes have been evaluated on a plant specific basis. The required plant specific analyses and evaluations included transient analysis of the turbine trip event (FSAR, Section 15.2.3), evaluation of the changes on the peak clad temperature from the Small Break the Loss of Coolant Accident (LOCA) event, and disposition of the changes on all other FSAR events. The revised analysis evaluations were based on the existing design pressure of the reactor coolant system (RCS) and the main steam (MS) system.

These analyses and evaluations demonstrate that there is adequate margin to the specified acceptable fuel design limits (SAFDL) and the design pressures of the RCS and the MS system.

The evaluations also demonstrate that the change will result in acceptable peak clad temperature (PCT) results for LOCA analyses. The change has no impact on the design pressure for the containment as peak containment pressure and temperature are obtained from postulated pipe breaks in the containment that do not challenge the MSSV lift setpoints. The MSSVs vent directly to open, ambient conditions and do not directly contribute to the temperature or pressure profile for any structure, system, or component.

There is a change in the flow rate credited for the auxiliary feedwater system (AFW) based on the higher MSSV opening tolerance. This change has been evaluated for each of the FSAR Chapter 15 events. The impact of the decrease in AFW flow is included in the PCT change for SBLOCA. The AFW flow effects for all other events have been determined to be acceptable.

As a result, the probability of a malfunction of the RCS and the main steam system are not increased and the consequences of such an accident remain acceptable. Therefore, the proposed TS changes do not significantly increase the probability or consequences of an accident previously evaluated.

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

The proposed TS changes allow for an increase in the as-found MSSV setpoint tolerance from +/-1%

to +/-3%. In addition, the proposed amendment request includes a conservative change to the reactor trip on high pressurizer level and makes TS 3.4.3 consistent with the initial pressurizer level

U.S. Nuclear Regulatory Commission Page 18 of 20 Serial HNP-15-038 used in the re-analysis of the FSAR, Section 15.2.3 turbine trip overpressure event.

Plant specific analyses and evaluations indicate that the plant response to any previously evaluated event will remain acceptable. All plant systems, structures, and components will continue to be capable of performing their required safety function as required by event analysis guidance.

The proposed TS changes do not alter the MSSV nominal lift setpoints. The operation and response of the affected equipment important to safety has been evaluated and found to be acceptable. All structures and components will continue to be operated within acceptable operating and/or design parameters. No system, structure, or component will be subjected to a condition that has not been evaluated and determined to be acceptable using the guidance required for specific event analysis.

Therefore, the proposed TS changes do not create the possibility of a new or different kind of accident from any previously evaluated.

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

The proposed TS changes allow for an increase in the as-found MSSV setpoint tolerance from +/-1%

to +/-3%. In addition, the proposed amendment request includes a conservative change to the reactor trip setpoint on high pressurizer level and makes TS 3.4.3 consistent with the initial pressurizer level used in the re-analysis of the FSAR Section, 15.2.3 turbine trip overpressure event.

The proposed TS changes do not alter the MSSV nominal lift setpoints. The operation and response of the affected equipment important to safety is unchanged. All systems, structures, and components will continue to be operated within acceptable operating and/or design parameters.

The calculated peak reactor vessel pressure and main steam system pressure for the turbine trip overpressure event remains within the acceptance criteria. A new analysis is submitted to support the change. The model used for the re-analyzed turbine trip event (FSAR, Section 15.2.3) is based on methodologies previously approved by the NRC for other licensees.

The consequences of the turbine trip event continue to be within the regulatory limit for the event, thus the margin of safety for overpressure remains unchanged. The impact on LOCA has been evaluated and the PCT change results in a PCT that is lower than the regulatory limit. Therefore, the margin to safety for cladding performance in this event is not reduced.

The margin of safety for the containment is unaffected by the proposed change. Therefore, the proposed TS changes do not involve a significant reduction in a margin of safety.

Based on the above, Duke Energy concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92, and, accordingly, a finding of "no significant hazards consideration" is justified.

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

U.S. Nuclear Regulatory Commission Page 19 of 20 Serial HNP-15-038

5. ENVIRONMENTAL CONSIDERATION Duke Energy has determined that the proposed amendment would change a requirement with respect to use of a facility component located within the restricted area, as defined in 10 CFR 20. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released onsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure.

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

U.S. Nuclear Regulatory Commission Page 20 of 20 Serial HNP-15-038

6. REFERENCES

1. Duke Energy Topical Report DPC-NE-3000-PA, Oconee Nuclear Station, McGuire Nuclear Station, Catawba Nuclear Station, Thermal-Hydraulic Transient Analysis Methodology, Revision 5a, October 2012.

2. Duke Energy Topical Report DPC-NE-3002-A, UFSAR Chapter 15 System Transient Methodology, Revision 4b, September 2010.

3. AREVA Report EMF-92-081 (P)(A), Statistical Setpoint/Transient Methodology for Westinghouse Type Reactors, Revision 1, Siemens Power Corporation, July 2000 (HNP TS 6.9.1.6.2.i).

4. AREVA Report ANP-3011 (P), Harris Nuclear Plant Unit 1 Realistic Large Break LOCA Analysis, Revision 1, as approved by NRC Safety Evaluation Report dated May 30, 2012 (HNP TS 6.9.1.6.2.f).

5. AREVA Report EMF-2328 (P)(A), PWR Small Break LOCA Evaluation Model, S- RELAP5 Based, Revision 0, Framatome ANP, May 2001, and Errata, January 2008 (HNP TS 6.9.1.6.2.m).

6. U.S. Nuclear Regulatory Commission Letter, Shearon Harris Nuclear Power Plant, Unit 1 - Issuance of Amendment Re: Measurement Uncertainty Recapture Power Uprate (TAC NO. ME6169), dated May 30, 2012 (ADAMS Accession No. ML11356A096)

7. Westinghouse Topical Report WCAP-10698-P-A, Steam Generator Tube Rupture Analysis Methodology to Determine the Margin to Steam Generator Overfill, August 1987.

8. AREVA Report ANF-89-151 (P)(A), ANF-RELAP Methodology for Pressurized Water Reactors:

Analysis of Non-LOCA Chapter 15 Events, Advanced Nuclear Fuels Corporation, May 1992 (HNP TS 6.9.1.6.2.b).

9. AREVA Report EMF-2310 (P)(A), SRP Chapter 15 Non-LOCA Methodology for Pressurized Water Reactors, Revision 1, Framatome ANP, May 2004 (HNP TS 6.9.1.6.2.n).

10. AREVA Report EMF-84-93 (P)(A), Steam Line Break Methodology for PWRs, Revision 1, Siemens Power Corporation, February 1999 (HNP TS 6.9.1.6.2.e).

11. AREVA Report XN-NF-78-44 (NP)(A), A Generic Analysis of the Control Rod Ejection Transient for Pressurized Water Reactors, Exxon Nuclear Company, October 1983 (HNP TS 6.9.1.6.2.g).

12. NUREG-0800, Standard Review Plan for the Review of Safety Analysis Report for Nuclear Power Plants, Section 5.2.2, Revision 3, March 2007 (ADAMS Accession No. ML070540076)

13. U.S. Nuclear Regulatory Commission Letter, Shearon Harris Nuclear Power Plant, Unit 1 - Issuance of Amendment to Allow the use of Thermal Hydraulic Analysis Code S- RELAP5 for Non-Loss-of-Coolant Accident Transients (TAC NO. ME1735), dated December 23, 2010 (ADAMS Accession No. ML102310361)

14. Duke Energy Letter, Annual Report of Changes Pursuant to 10 CFR 50.46, dated May 14, 2015 (ADAMS Accession No. ML15134A029)

U.S. Nuclear Regulatory Commission Serial HNP-15-038 SERIAL HNP-15-038 ENCLOSURE 2 PROPOSED TECHNICAL SPECIFICATION CHANGES SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 DOCKET NO. 50-400 RENEWED LICENSE NUMBER NPF-63

No changes to this page. Included for information only.

2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS REACTOR CORE 2.1.1 The combination of THERMAL POWER, pressurizer pressure, and the highest operating loop coolant temperature (Tavg) shall not exceed the limits shown in Figure 2.1-1 for 3-loop operation.

APPLICABILITY: MODES 1 and 2.

ACTION:

a. Whenever the point defined by the combination of the highest operating loop average temperature and THERMAL POWER has exceeded the appropriate pressurizer pressure line, be in HOT STANDBY within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , and comply with the requirements of Specification 6.7.1.

b. Operation with less than 3 loops is governed by Specification 3.4.1.1.

REACTOR COOLANT SYSTEM PRESSURE 2.1.2 The Reactor Coolant System pressure shall not exceed 2735 psig except during hydrostatic testing.

APPLICABILITY: MODES 1, 2, 3, 4, and 5.

ACTION:

MODES 1 and 2:

Whenever the Reactor Coolant System pressure has exceeded 2735 psig, be in HOT STANDBY with the Reactor Coolant System pressure within its limit within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , and comply with the requirements of Specification 6.7.1.

MODES 3, 4, and 5:

Whenever the Reactor Coolant System pressure has exceeded 2735 psig, reduce the Reactor Coolant System pressure to within its limit within 5 minutes, and comply with the requirements of Specification 6.7.1.

2.2 LIMITING SAFETY SYSTEM SETTINGS REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS 2.2.1 The Reactor Trip System Instrumentation and Interlock Setpoints shall be set consistent with the Trip Setpoint values shown in Table 2.2-1.

APPLICABILITY: As shown for each channel in Table 3.3-1.

SHEARON HARRIS - UNIT 1 2-1

No changes to this page. Included for information only.

FIGURE 2.1 1 REACTOR CORE SAFETY LIMITS - THREE LOOPS IN OPERATION WITH MEASURED RCS FLOW > [293,540 GPM X (1.0 + C1)]

SHEARON HARRIS - UNIT 1 2-2 Amendment No. 139

No changes to this page. Included for information only.

SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS APPLICABILITY (Continued)

ACTION:

a. With a Reactor Trip System Instrumentation or Interlock Setpoint less conservative than the value shown in the Trip Setpoint column but more conservative than the value shown in the Allowable Value column of Table 2.2-1, adjust the Setpoint consistent with the Trip Setpoint value.

b. With the Reactor Trip System Instrumentation or Interlock Setpoint less conservative than the value shown in the Allowable Values column of Table 2.2-1, either:

1. Adjust the Setpoint consistent with the Trip Setpoint value of Table 2.2-1 and determine within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months that Equation 2.2-1 was satisfied for the affected channel, or

2. Declare the channel inoperable and apply the applicable ACTION statement requirement of Specification 3.3.1 until the channel is restored to OPERABLE status with its Setpoint adjusted consistent with the Trip Setpoint value.

Equation 2.2-1 Z + R + S TA Where:

Z = The value from Column Z of Table 2.2-1 for the affected channel, R = The "as measured" value (in percent span) of rack error for the affected channel, S = Either the "as measured" value (in percent span) of the sensor error, or the value from Column S (Sensor Error) of Table 2.2-1 for the affected channel, and TA = The value from Column TA (Total Allowance) of Table 2.2-1 for the affected channel.

c. With a Reactor Trip System Instrumentation Channel or Interlock inoperable, take the appropriate ACTION shown in Table 3.3-1.

SHEARON HARRIS - UNIT 1 2-3 Amendment No. 107

TABLE 2.2-1 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS TOTAL SENSOR ERROR FUNCTIONAL UNIT ALLOWANCE (TA) Z (S) TRIP SETPOINT ALLOWABLE VALUE

1. Manual Reactor Trip N.A. N.A. N.A. N.A. N.A.

2. Power Range, Neutron Flux

a. High Setpoint 5.83 4.56 0 108% of RTP** 109.5% of RTP**

See NOTES 7, 8

b. Low Setpoint 7.83 4.56 0 25% of RTP** 26.8% of RTP**

See NOTES 7, 8

3. Power Range, Neutron Flux, 2.33 0.83 0 5% of RTP** with a 6.3% of RTP** with a time High Positive Rate time constant 2 constant 2 seconds seconds See NOTES 7, 8

4. Power Range, Neutron Flux, 2.33 0.83 0 5% of RTP** with a 6.3% of RTP** with a time High Negative Rate time constant 2 constant 2 seconds seconds See NOTES 7, 8

5. Intermediate Range, Neutron 17.0 8.41 0 25% of RTP** 30.9% of RTP**

Flux

6. Source Range, Neutron Flux 17.0 10.01 0 105 cps 1.4 x 105 cps

7. Overtemperature T 9.0 7.31 Note 5 See Note 1 See Note 2

8. Overpower T 4.0 2.32 1.3 See Note 3 See Note 4

9. Pressurizer Pressure-Low 5.0 1.52 1.5 1960 psig 1948 psig

10. Pressurizer Pressure-High 7.5 1.52 1.5 2385 psig 2397 psig

11. Pressurizer Water 8.0 3.42 1.75 92% of instrument 93.5% of instrument span Level-High span INSERT 88.5% of INSERT 87% of instrument span instrument span

- RTP = RATED THERMAL POWER See NOTES 7, 8 SHEARON HARRIS - UNIT 1 2-4 Amendment No. 139

No changes to this page. Included for information only.

TABLE 2.2-1 (continued)

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS TOTAL SENSOR ERROR FUNCTIONAL UNIT ALLOWANCE (TA) Z (S) TRIP SETPOINT ALLOWABLE VALUE

12. Reactor Coolant Flow-Low 4.58 1.98 0.6 90.5% of loop full 89.5% of loop full indicated flow indicated flow

13. Steam Generator Water 25.0 17.45 2.0 25.0% of narrow range 23.5% of narrow range Level Low-Low instrument span instrument span

14. Steam Generator Water 8.9 5.95 2.0 25.0% of narrow range 24.05% of narrow range Level - Low instrument span instrument spa Coincident With 20.0 3.01 Note 6 40% of full steam flow 43.1% of full steam flow Steam/Feedwater Flow at RTP** at RTP**

Mismatch

15. Undervoltage - Reactor 14.0 1.3 0.0 5148 volts 4920 volts Coolant Pumps

16. Underfrequency - Reactor 5.0 3.0 0.0 57.5 Hz 57.3 Hz Coolant Pumps

17. Turbine Trip

a. Low Fluid Oil Pressure N.A. N.A. N.A. 1000 psig 950 psig

b. Turbine Throttle Valve N.A. N.A. N.A. 1% open 1% open Closure

18. Safety Injection Input from N.A. N.A. N.A. N.A. N.A.

ESF

- RTP = RATED THERMAL POWER SHEARON HARRIS - UNIT 1 2-5 Amendment No. 126

No changes to this page. Included for information only.

TABLE 2.2-1 (continued)

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS TOTAL SENSOR ERROR FUNCTIONAL UNIT ALLOWANCE (TA) Z (S) TRIP SETPOINT ALLOWABLE VALUE

19. Reactor Trip System Interlocks

a. Intermediate Range N.A. N.A. N.A. 1 x 10-10 amp 6 x 10-11 amp Neutron Flux, P-6

b. Low Power Reactor Trips Block, P-7

1) P-10 input N.A. N.A. N.A. 10% of RTP** 12.1% of RTP**

2) P-13 input N.A. N.A. N.A. 10% RTP** Turbine 12.1% RTP** Turbine Inlet Pressure Inlet Pressure Equivalent Equivalent

c. Power Range Neutron N.A. N.A. N.A. 49% of RTP** 51.1% of RTP**

Flux, P-8

d. Power Range Neutron N.A. N.A. N.A. 10% of RTP** 7.9% of RTP**

Flux, P-10

e. Turbine Impulse N.A. N.A. N.A. 10% RTP** Turbine 12.1% RTP** Turbine Chamber Pressure, P-13 Inlet Pressure Inlet Pressure Equivalent Equivalent

20. Reactor Trip Breakers N.A. N.A. N.A. N.A. N.A.

21. Automatic Trip and Interlock N.A. N.A. N.A. N.A. N.A.

Logic

22. Reactor Trip Bypass N.A. N.A. N.A. N.A. N.A.

Breakers

- RTP = RATED THERMAL POWER SHEARON HARRIS - UNIT 1 2-6 Amendment No. 139

No changes to this page. Included for information only.

TABLE 2.2-1 (Continued)

TABLE NOTATIONS NOTE 1: OVERTEMPERATURE T

( ) 1 ( )

T 1+ T o + K (P ) f (I)

( ) 3 ( )

Where: T = Measured T by RTD Instrumentation;

= Lead-lag compensator on measured T; 1 2 = Time constants utilized in lead-lag compensator for T, 1 = 0 s, 2 = 0 s;

= Lag compensator on measured T; 3 = Time constants utilized in the lag compensator for T, 3 = 4 s; T o = Indicated T at RATED THERMAL POWER; K1 = 1.185; K2 = 0.0224/°F;

= The function generated by the lead-lag compensator for T avg dynamic compensation; 4 5 = Time constants utilized in the lead-lag compensator for T avg , 4 = 22 s, 5 = 4 s; SHEARON HARRIS - UNIT 1 2-7 Amendment No. 107

No changes to this page. Included for information only.

TABLE 2.2-1 (Continued)

TABLE NOTATIONS NOTE 1: (Continued)

T = Average temperature, °F;

= Lag compensator on measured T avg ;

6 = Time constant utilized in the measured T avg lag compensator, 6 = 0 s; T' = Reference T avg at RATED THERMAL POWER (588.8°F);

K3 = 0.0012/psig; P = Pressurizer pressure, psig; P = 2235 psig (Nominal RCS operating pressure);

S = Laplace transform operator, s-1; and f 1 (I) is a function of the indicated difference between top and bottom detectors of the power-range neutron ion chambers; with gains to be selected based on measured instrument response during plant startup tests such that:

(1) For q t - q b between -21.6% and +12.0%, f 1 (I) = 0, where q t and q b are percent RATED THERMAL POWER in the top and bottom halves of the core respectively, and q t + q b is total THERMAL POWER in percent of RATED THERMAL POWER; (2) For each percent that the magnitude of q t - q b exceeds -21.6%, the T Trip Setpoint shall be automatically reduced by 1.75% of its value at RATED THERMAL POWER; and (3) For each percent that the magnitude of q t - q b exceeds + 12.0%, the T Trip Setpoint shall be automatically reduced by 1.50% of its value at RATED THERMAL POWER.

NOTE 2: The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than 1.4% of T span for T input; 2.0% of T span for T avg input; 0.4% of T span for pressurizer pressure input; and 0.7% of T span for I input.

SHEARON HARRIS - UNIT 1 2-8 Amendment No. 107

No changes to this page. Included for information only.

TABLE 2.2-1 (Continued)

TABLE NOTATIONS NOTE 3: OVERPOWER T

( ) () ( ) () ()

T T o T f (I)

( ) ( ) ( ) ( ) ( )

Where: T = As defined in Note 1,

= As defined in Note 1, 1, 2 = As defined in Note 1,

= As defined in Note 1, 3 = As defined in Note 1, T o = As defined in Note 1, K4 = 1.12, K5 = 0.02/°F for increasing average temperature and 0 for decreasing average temperature,

= The function generated by the rate-lag compensator for T avg dynamic compensation, 7 = Time constants utilized in the rate-lag compensator for T avg , 7 = 13 s,

= As defined in Note 1, 6 = As defined in Note 1, SHEARON HARRIS - UNIT 1 2-9 Amendment No. 107

No changes to this page. Included for information only.

TABLE 2.2-1 (Continued)

TABLE NOTATIONS NOTE 3: (Continued)

K6 = 0.002/°F for T > T" and K 6 = 0 for T T",

T = As defined in Note 1, T" = Reference T avg at RATED THERMAL POWER (588.8°F),

S = As defined in Note 1, and f 2 (I) = 0 for all I.

NOTE 4: The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than 1.4% of T span for T input and 0.2% of T span for T avg input.

NOTE 5: The sensor error is: 1.3% of T span for T/T avg temperature measurements; and 1.0% of T span for pressurizer pressure measurements.

NOTE 6: The sensor error (in % span of Steam Flow) is: 1.1% for steam flow; 1.8% for feedwater flow; and 2.4% for steam pressure.

NOTE 7: If the as-found channel setpoint is outside its predefined as-found tolerance, the channel shall be evaluated to verify that it is functioning as required before returning the channel to service.

NOTE 8: The instrument channel setpoint shall be reset to a value that is within the as-left tolerance around the Trip Setpoint in Table 2.2-1 (Nominal Trip Setpoint (NTSP) at the completion of the surveillance; otherwise, the channel shall be declared inoperable. Setpoints more conservative than the NTSP are acceptable provided that the as-found and as-left tolerances apply to the actual setpoint implemented in the surveillance procedures (field setting) to confirm channel performance. The methodologies used to determine NTSPs and the as-found and the as-left tolerances are specified in EGR-NGGC-0153, Engineering Instrument Setpoints. The as-found and as-left tolerances are specified in PLP-106, Technical Specification Equipment List Program and Core Operating Limits Report.

SHEARON HARRIS - UNIT 1 2-10 Amendment No. 139

REACTOR COOLANT SYSTEM 3/4.4.3 PRESSURIZER INSERT 75%

LIMITING CONDITION FOR OPERATION 3.4.3 The pressurizer shall be OPERABLE with a water level of less than or equal to 92% of indicated span, and at least two groups of pressurizer heaters each having a capacity of at least 125 kW.

APPLICABILITY: MODES 1, 2, and 3.

ACTION:

a. With only one group of pressurizer heaters OPERABLE, restore at least two groups to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b. With the pressurizer otherwise inoperable, be in at least HOT STANDBY with the Reactor Trip System breakers open within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS 4.4.3.1 The pressurizer water level shall be determined to be within its limit at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

4.4.3.2 The capacity of each of the above required groups of pressurizer heaters shall be verified by energizing the heaters and measuring circuit power (kW) at least once per 18 months.

SHEARON HARRIS - UNIT 1 3/4 4-10 Amendment No. 109

No changes to this page. Included for information only.

3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE SAFETY VALVES LIMITING CONDITION FOR OPERATION 3.7.1.1 All main steam line Code safety valves associated with each steam generator shall be OPERABLE with lift settings as specified in Table 3.7-2.

APPLICABILITY: MODES 1, 2, and 3.

ACTION:

a. With one or more main steam line Code safety valves inoperable, operation may proceed provided, that within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , either the inoperable valve is restored to OPERABLE status or the Power Range Neutron Flux High Trip Setpoint is reduced per Table 3.7-1; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS 4.7.1.1 No additional requirements other than those required by the Inservice Testing Program.

SHEARON HARRIS - UNIT 1 3/4 7-1 Amendment No. 127

TABLE 3.7-2 STEAM LINE SAFETY VALVES PER LOOP VALVE NUMBER LIFT SETTING (+/- 1%)* ORIFICE SIZE (IN.2)

STEAM GENERATOR INSERT 3%

A B C 1MS-43 1MS-44 1MS-45 1170 psig 16.0 1MS-46 1MS-47 1MS-48 1185 psig 16.0 1MS-49 1MS-50 1MS-51 1200 psig 16.0 1MS-52 1MS-53 1MS-54 1215 psig 16.0 1MS-55 1MS-56 1MS-57 1230 psig 16.0

The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.

SHEARON HARRIS - UNIT 1 3/4 7-3

U.S. Nuclear Regulatory Commission Serial HNP-15-038 SERIAL HNP-15-038 ENCLOSURE 3 REVISED TECHNICAL SPECIFICATION CHANGES SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 DOCKET NO. 50-400 RENEWED LICENSE NUMBER NPF-63

TABLE 2.2-1 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS TOTAL ALLOWANCE SENSOR FUNCTIONAL UNIT (TA) Z ERROR (S) TRIP SETPOINT ALLOWABLE VALUE

1. Manual Reactor Trip N.A. N.A. N.A. N.A. N.A.

2. Power Range, Neutron Flux

a. High Setpoint 5.83 4.56 0 108% of RTP** 109.5% of RTP**

See NOTES 7, 8

b. Low Setpoint 7.83 4.56 0 25% of RTP** 26.8% of RTP**

See NOTES 7, 8

3. Power Range, Neutron Flux, 2.33 0.83 0 5% of RTP** with a 6.3% of RTP** with a time High Positive Rate time constant 2 constant 2 seconds seconds See NOTES 7, 8

4. Power Range, Neutron Flux, 2.33 0.83 0 5% of RTP** with a 6.3% of RTP** with a time High Negative Rate time constant 2 constant 2 seconds seconds See NOTES 7, 8

5. Intermediate Range, Neutron 17.0 8.41 0 25% of RTP** 30.9% of RTP**

Flux

6. Source Range, Neutron Flux 17.0 10.01 0 105 cps 1.4 x 105 cps

7. Overtemperature T 9.0 7.31 Note 5 See Note 1 See Note 2

8. Overpower T 4.0 2.32 1.3 See Note 3 See Note 4

9. Pressurizer Pressure-Low 5.0 1.52 1.5 1960 psig 1948 psig

10. Pressurizer Pressure-High 7.5 1.52 1.5 2385 psig 2397 psig

11. Pressurizer Water 8.0 3.42 1.75 87% of instrument 88.5% of instrument span Level-High span See NOTES 7, 8

- RTP = RATED THERMAL POWER SHEARON HARRIS - UNIT 1 2-4 Amendment No. ____

REACTOR COOLANT SYSTEM 3/4.4.3 PRESSURIZER LIMITING CONDITION FOR OPERATION 3.4.3 The pressurizer shall be OPERABLE with a water level of less than or equal to 75% of indicated span, and at least two groups of pressurizer heaters each having a capacity of at least 125 kW.

APPLICABILITY: MODES 1, 2, and 3.

ACTION:

a. With only one group of pressurizer heaters OPERABLE, restore at least two groups to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b. With the pressurizer otherwise inoperable, be in at least HOT STANDBY with the Reactor Trip System breakers open within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS 4.4.3.1 The pressurizer water level shall be determined to be within its limit at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

4.4.3.2 The capacity of each of the above required groups of pressurizer heaters shall be verified by energizing the heaters and measuring circuit power (kW) at least once per 18 months.

SHEARON HARRIS - UNIT 1 3/4 4-10 Amendment No. ____

TABLE 3.7-2 STEAM LINE SAFETY VALVES PER LOOP VALVE NUMBER LIFT SETTING (+/- 3%)* ORIFICE SIZE (IN.2)

STEAM GENERATOR A B C 1MS-43 1MS-44 1MS-45 1170 psig 16.0 1MS-46 1MS-47 1MS-48 1185 psig 16.0 1MS-49 1MS-50 1MS-51 1200 psig 16.0 1MS-52 1MS-53 1MS-54 1215 psig 16.0 1MS-55 1MS-56 1MS-57 1230 psig 16.0

The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.

SHEARON HARRIS - UNIT 1 3/4 7-3 Amendment No. ____

U.S. Nuclear Regulatory Commission Serial HNP-15-038 SERIAL HNP-15-038 ENCLOSURE 4 PROPOSED TECHNICAL SPECIFICATION BASES CHANGES SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 DOCKET NO. 50-400 RENEWED LICENSE NUMBER NPF-63

No changes to this page. Included for information only.

SAFETY LIMITS BASES 2.1.2 REACTOR COOLANT SYSTEM PRESSURE The restriction of this Safety Limit protects the integrity of the Reactor Coolant System (RCS) from overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere.

The reactor pressure vessel, pressurizer, and the RCS piping, pumps, valves and fittings are designed to Section III, Division I of the ASME Code for Nuclear Power Plants, which permits a maximum transient pressure of 110% to 125% of design pressure (2485 psig) depending on component. The Safety Limit of 2735 psig (110% of design pressure) is therefore consistent with the design criteria and associated Code requirements.

The entire RCS is hydrotested at 125% (3107 psig) of design pressure, to demonstrate integrity prior to initial operation.

2.2 LIMITING SAFETY SYSTEM SETTINGS 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS The Reactor Trip Setpoint Limits specified in Table 2.2-1 are the nominal values at which the Reactor trips are set for each functional unit. The Trip Setpoints have been selected to ensure that the core and Reactor Coolant System are prevented from exceeding their Safety Limits during normal operation and design basis anticipated operational occurrences and to assist the Engineered Safety Features Actuation System in mitigating the consequences of accidents. The Setpoint for a Reactor Trip System or interlock function is considered to be adjusted consistent with the nominal value when the "as measured" Setpoint is within the band allowed for calibration accuracy. For example, if a bistable has a trip setpoint of 100%, a span of 125%, and a calibration accuracy of 0.5% of span, then the bistable is considered to be adjusted to the trip setpoint as long as the "as measured" value for the bistable is 100.62%.

To accommodate the instrument drift assumed to occur between operational tests and the accuracy to which Setpoints can be measured and calibrated, Allowable Values for the Reactor Trip Setpoints have been specified in Table 2.2-1. Operation with Setpoints less conservative than the Trip Setpoint but within the Allowable Value is acceptable since an allowance has been made in the safety analysis to accommodate this error. An optional provision has been included for determining the OPERABILITY of a channel when its Trip Setpoint is found to exceed the Allowable Value. The methodology of this option utilizes the "as measured" deviation from the specified calibration point for rack and sensor components in conjunction with a statistical combination of the other uncertainties of the instrumentation to measure the process variable and the uncertainties in calibrating the instrumentation. In Equation 2.2-1, Z + R + S TA, the interactive effects of the errors in the rack and the sensor, and the "as measured" values of the errors are considered. Z, as specified in Table 2.2-1, in percent span, is the statistical summation of SHEARON HARRIS - UNIT 1 B 2-2

LIMITING SAFETY SYSTEM SETTINGS BASES REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS (Continued) errors assumed in the analysis excluding those associated with the sensor and rack drift and the accuracy of their measurement. TA or Total Allowance is the difference, in percent span, between the Trip Setpoint and the value used in the analysis for Reactor trip. R or Rack Error is the "as measured" deviation, in percent span, for the affected channel from the specified Trip Setpoint. S or Sensor Error is either the "as measured" deviation of the sensor from its calibration point or the value specified in Table 2.2-1, in percent span, from the analysis assumptions. Use of Equation 2.2-1 allows for a sensor drift factor and an increased rack drift factor, and provides a threshold value for determination of OPERABILITY.

The methodology to derive the Trip Setpoints is based upon combining all of the uncertainties in the channels. Inherent to the determination of the Trip Setpoints are the magnitudes of these channel uncertainties. Sensors and other instrumentation utilized in these channels are expected to be capable of operating within the allowances of these uncertainty magnitudes.

Rack drift in excess of the Allowable Value exhibits the behavior that the rack has not met its allowance. Being that there is a small statistical chance that this will happen, an infrequent excessive drift is expected. Rack or sensor drift, in excess of the allowance that is more than occasional, may be indicative of more serious problems and should warrant further investigation.

The various Reactor trip circuits automatically open the Reactor trip breakers whenever a condition monitored by the Reactor Trip System reaches a preset or calculated level. In addition to redundant channels and trains, the design approach provides a Reactor Trip System which monitors numerous system variables, therefore providing Trip System functional diversity. The functional capability at the specified trip setting is required for those anticipatory or diverse Reactor trips for which no direct credit was assumed in the safety analysis to enhance the overall reliability of the Reactor Trip System. The Reactor Trip System initiates a Turbine trip signal whenever Reactor trip is initiated. This prevents the reactivity insertion that would otherwise result from excessive Reactor Coolant System cooldown and thus avoids unnecessary actuation of the Engineered Safety Features Actuation System.

Reactor Trip System Instrumentation Setpoints and TSTF-493 This section applies only to the Functional Units to which Notes 7 and 8 in the Trip Setpoint Column are applicable. Those Functional Units have revisions in accordance with Technical Specification Task Force Traveler 493 (TSTF-493). Clarify Application of Setpoint Methodology for LSSS Functions. Those Functional Units are limited to x Power Range, Neutron Flux High Setpoint x Power Range, Neutron Flux Low Setpoint x Power Range, Neutron Flux High Positive Rate, and x Power Range, Neutron Flux High Negative Rate Notes 7 and 8 have been added to Table 2.2-1 that require verifying both trip setpoint setting as-found and as-left values during surveillance testing. In accordance with 10 CFR 50.36, these functions are Limiting Safety System Settings. Adding test requirements ensures that instruments will function as required to initiate protective systems or actuate mitigating systems at the point assumed in the applicable safety analysis. These notes address NRC staff concerns with Technical Specification Allowable Values. Specifically, calculated Allowable Values may be non-conservative depending upon the evaluation of instrument performance history, and the as-left requirements of the calibration procedures could have an adverse effect on equipment INSERT

Pressurizer Water Level - High Setpoint SHEARON HARRIS - UNIT 1 B 2-3 Amendment No. 139

No changes to this page. Included for information only.

LIMITING SAFETY SYSTEM SETTINGS BASES REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS (Continued) operability. In addition, using Allowable Values as the limiting setting for assessing instrument channel operability may not be fully in compliance with the intent of 10 CFR 50.36, and the existing surveillance requirements would not provide adequate assurance that instruments will always actuate safety functions at the point assumed in the applicable safety analysis. In the Harris Technical Specifications, the term Trip Setpoint is analogous to Nominal Trip Setpoint (NTSP) in TSTF-493.

Note 7 requires a channel performance evaluation when the as-found setting is outside its as-found tolerance. The performance evaluation verifies that the channel will continue to behave in accordance with safety analysis and instrument performance assumptions in the setpoint methodology. The purpose of this evaluation is to provide confidence in the performance prior to returning the channel to service. If the as-found setting is non-conservative with respect to the Allowable Value, the channel is INOPERABLE. If the as-found setting is conservative with respect to the Allowable Value but is outside the as-found tolerance band, the channel is OPERABLE but degraded. The degraded channel condition will be further evaluated during performance of the surveillance. This evaluation will consist of resetting the channel setpoint to within the as-left tolerances applicable to the actual setpoint implemented in the surveillance procedures (field setting), and evaluating the channel response. If the channel is functioning as required and is expected to pass the next surveillance, then the channel is OPERABLE and can be restored to service at the completion of the surveillance. After the surveillance is completed, the channel as-found condition is entered into the corrective action program for further analysis and trending.

Note 8 requires that the as-left channel setting be reset to a value that is within the as-left tolerances about the Trip Setpoint in Table 2.2-1 or within as-left tolerances about a more conservative actual (field) setpoint. As-left channel settings outside the as-left tolerances of PLP-106 and the surveillance procedures cause the channel to be INOPERABLE.

A tolerance is necessary because no device perfectly measures the process. Additionally, it is not possible to read and adjust a setting to an absolute value due to the readability and/or accuracy of the test instruments or the ability to adjust potentiometers. The as-left tolerance is considered in the setpoint calculation. Failure to set the actual plant trip setpoint to within as-left the tolerances of the NTSP or within as-left tolerances of a more conservative actual field setpoint would invalidate the assumptions in the setpoint calculation, because any subsequent instrument drift would not start from the expected as-left setpoint. The determination will consider whether the instrument is degraded or is capable of being reset and performing its specified safety function. If the channel is determined to be functioning as required (i.e., the channel can be adjusted to within the as-left tolerance and is determined to be functioning normally based on the determination performed prior to returning the channel to service), then the channel is OPERABLE and can be restored to service.

If the as-left instrument setting cannot be returned to a setting within the prescribed as-left tolerance band, the instrument would be declared INOPERABLE.

The methodologies for calculating the as-found tolerances and as-left tolerances about the Trip Setpoint or more conservative actual field setpoint are specified in EGR-NGGC-0153, Engineering Instrument Setpoints, which is incorporated by reference into the FSAR. The actual field setpoint and the associated as-found and as-left tolerances are specified in PLP-106, Technical Specification Equipment List Program and Core Operating Limits Report, the applicable section of which is incorporated by reference into the FSAR.

Limiting Trip Setpoint (LTSP) is generic terminology for the setpoint value calculated by means of the setpoint methodology documented in EGR-NGGC-0153. HNP uses the plant-specific term Nominal Trip Setpoint (NTSP) in place of the generic term LTSP. The NTSP is the LTSP with SHEARON HARRIS - UNIT 1 B 2-3a Amendment No. 139

No changes to this page. Included for information only.

LIMITING SAFETY SYSTEM SETTINGS BASES REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS (Continued) margin added, and is always equal to or more conservative than the LTSP. The NTSP may use a setting value that is more conservative than the LTSP, but for Technical Specification compliance with 10 CFR 50.36, the plant-specific setpoint term NTSP is cited in Note 8.

The NTSP meets the definition of a Limiting Safety System Setting per 10 CFR 50.36 and is a predetermined setting for a protective channel chosen to ensure that automatic protective actions will prevent exceeding Safety Limits during normal operation and design basis anticipated operational occurrences, and assist the Engineered Safety Features Actuation System in mitigating the consequences of accidents. The Allowable Value is the least conservative value of the as-found setpoint that the channel can have when tested, such that a channel is OPERABLE if the as-found setpoint is within the as-found tolerance and is conservative with respect to the Allowable Value during a CHANNEL CALIBRATION or CHANNEL OPERATIONAL TEST. As such, the Allowable Value differs from the NTSP by an amount greater than or equal to the expected instrument channel uncertainties, such as drift, during the surveillance interval.

In this manner, the actual NTSP setting ensures that a Safety Limit is not exceeded at any given point of time as long as the channel has not drifted beyond expected tolerances during the surveillance interval. Although the channel is OPERABLE under these circumstances, the trip setpoint must be left adjusted to a value within the as-left tolerance band, in accordance with uncertainty assumptions stated in the setpoint methodology (as-left criteria), and confirmed to be operating within the statistical allowances of the uncertainty terms assigned (as-found criteria).

Field setting is the term used for the actual setpoint implemented in the plant surveillance procedures, where margin has been added to the calculated field setting. The as-found and as-left tolerances apply to the field settings implemented in the surveillance procedures to confirm channel performance. A trip setpoint may be set more conservative than the NTSP as necessary in response to plant conditions. However, in this case, the instrument operability must be verified based on the field setting and not the NTSP.

Manual Reactor Trip The Reactor Trip System includes manual Reactor trip capability.

Power Range, Neutron Flux In each of the Power Range Neutron Flux channels there are two independent bistables, each with its own trip setting used for a High and Low Range trip setting. The Low Setpoint trip provides protection during subcritical and low power operations to mitigate the consequences of a power excursion beginning from low power, and the High Setpoint trip provides protection during power operations to mitigate the consequences of a reactivity excursion from all power levels.

SHEARON HARRIS - UNIT 1 B 2-3b Amendment No. 139

No changes to this page. Included for information only.

LIMITING SAFETY SYSTEM SETTINGS BASES Power Range, Neutron Flux (Continued)

The Low Setpoint trip may be manually blocked above P-10 (a power level of approximately 10%

of RATED THERMAL POWER) and is automatically reinstated below the P-10 Setpoint.

Power Range, Neutron Flux, High Rates The Power Range Positive Rate trip provides protection against rapid flux increases which are characteristic of a rupture of a control rod drive housing. Specifically, this trip complements the Power Range Neutron Flux High and Low trips to ensure that the criteria are met for rod ejection from mid power.

The Power Range Negative Rate trip provides protection for control rod drop accidents. At high power a single or multiple rod drop accident could cause local flux peaking which could cause an unconservative local DNBR to exist. The Power Range Negative Rate trip will prevent this from occurring by tripping the reactor.

Intermediate and Source Range, Neutron Flux The Intermediate and Source Range, Neutron Flux trips provide core protection during reactor startup to mitigate the consequences of an uncontrolled rod cluster control assembly bank withdrawal from a subcritical condition. These trips provide redundant protection to the Low Setpoint trip of the Power Range, Neutron Flux channels. The Source Range channels will initiate a Reactor trip at about 105 counts per second unless manually blocked when P-6 becomes active. The Intermediate Range channels will initiate a Reactor trip at a current level equivalent to approximately 25% of RATED THERMAL POWER unless manually blocked when P-10 becomes active.

Overtemperature T The Overtemperature T trip provides core protection to prevent DNB for all combinations of pressure, power, coolant temperature, and axial power distribution, provided that the transient is slow with respect to transport to and response time of the temperature detectors (about 4 seconds), and pressure is within the range between the Pressurizer High and Low Pressure trips.

The Setpoint is automatically varied with: (1) coolant temperature to correct for temperature induced changes in density and heat capacity of water and includes dynamic compensation for transport to and response time of the loop temperature detectors, (2) pressurizer pressure, and (3) axial power distribution. With normal axial power distribution, this Reactor trip limit is always below the core Safety Limit as shown in Figure 2.1-1. If axial peaks are greater than design, as indicated by the difference between top and bottom power range nuclear detectors, the Reactor trip is automatically reduced according to the notations in Table 2.2-1.

SHEARON HARRIS - UNIT 1 B 2-4 Amendment No. 46

No changes to this page. Included for information only.

LIMITING SAFETY SYSTEM SETTINGS BASES Overpower T The Overpower T trip provides assurance of fuel integrity (e.g., no fuel pellet melting and less than 1% cladding strain) under all possible overpower conditions, limits the required range for Overtemperature T trip, and provides a backup to the High Neutron Flux trip. The Setpoint is automatically varied with: (1) coolant temperature to correct for temperature induced changes in density and heat capacity of water, and (2) rate of change of temperature for dynamic compensation for transport to and response time of the loop temperature detectors, to ensure that the allowable heat generation rate (kW/ft) is not exceeded.

Pressurizer Pressure In each of the pressurizer pressure channels, there are two independent bistables, each with its own trip setting to provide for a High and Low Pressure trip thus limiting the pressure range in which reactor operation is permitted. The Low Setpoint trip protects against low pressure which could lead to DNB by tripping the reactor in the event of a loss of reactor coolant pressure.

On decreasing power the Low Setpoint trip is automatically blocked by the loss of P-7 (a power level of approximately 10% of RATED THERMAL POWER or turbine inlet pressure at approximately 10% of full power equivalent); and on increasing power, automatically reinstated by P-7.

The High Setpoint trip functions in conjunction with the pressurizer relief and safety valves to protect the Reactor Coolant System against system overpressure.

Pressurizer Water Level The Pressurizer High Water Level trip is provided to prevent water relief through the pressurizer safety valves. On decreasing power the Pressurizer High Water Level trip is automatically blocked by the loss of P-7 (a power level of approximately 10% of RATED THERMAL POWER or a turbine inlet pressure at approximately 10% of full power equivalent); and on increasing power, automatically reinstated by P-7.

Reactor Coolant Flow The Reactor Coolant Low Flow trips provide core protection to prevent DNB by mitigating the consequences of a loss of flow resulting from the loss of one or more reactor coolant pumps.

On increasing power above P-7 (a power level of approximately 10% of RATED THERMAL POWER or a turbine inlet pressure at approximately 10% of full power equivalent), an automatic Reactor trip will occur if the flow in more than one loop drops below 90.5% of nominal full loop flow. Above P-8 SHEARON HARRIS - UNIT 1 B 2-5 Amendment No. 139

No changes to this page. Included for information only.

LIMITING SAFETY SYSTEM SETTINGS BASES Reactor Coolant Flow (Continued)

(a power level of approximately 49% of RATED THERMAL POWER) an automatic Reactor trip will occur if the flow in any single loop drops below 90.5% of nominal full loop flow. Conversely, on decreasing power between P-8 and P-7, an automatic Reactor trip will occur on low reactor coolant flow in more than one loop; and below P-7, the trip function is automatically blocked.

Steam Generator Water Level The Steam Generator Water Level Low-Low trip protects the reactor from loss of heat sink in the event of a sustained steam/feedwater flow mismatch resulting from loss of normal feedwater.

The specified Setpoint provides allowances for starting delays of the Auxiliary Feedwater System.

Steam/Feedwater Flow Mismatch and Low Steam Generator Water Level The Steam/Feedwater Flow Mismatch in coincidence with a Steam Generator Low Water Level trip is not used in the transient and accident analyses but is included in Table 2.2-1 to ensure the functional capability of the specified trip settings and thereby enhance the overall reliability of the Reactor Trip System. This trip is redundant to the Steam Generator Water Level Low-Low trip.

The Steam/Feedwater Flow Mismatch portion of this trip is activated when the steam flow exceeds the feedwater flow by the setpoint value. The Steam Generator Low Water level portion of the trip is activated when the setpoint value is reached, as indicated by the narrow range instrument. These trip values include sufficient allowance in excess of normal operating values to preclude spurious trips but will initiate a Reactor trip before the steam generators are dry.

Therefore, the required capacity and starting time requirements of the auxiliary feedwater pumps are reduced and the resulting thermal transient on the Reactor Coolant System and steam generators is minimized.

Undervoltage and Underfrequency - Reactor Coolant Pump Buses The Undervoltage and Underfrequency Reactor Coolant Pump Bus trips provide core protection against DNB as a result of complete loss of forced coolant flow. The specified Setpoints assure a Reactor trip signal is generated before the Low Flow Trip Setpoint is reached. Time delays are incorporated in the Underfrequency and Undervoltage trips to prevent spurious Reactor trips from momentary electrical power transients.

On decreasing power the Undervoltage and Underfrequency Reactor Coolant Pump Bus trips are automatically blocked by the loss of P-7 (a power level of approximately 10% of RATED THERMAL POWER or a turbine inlet pressure SHEARON HARRIS - UNIT 1 B 2-6 Amendment No. 139

No changes to this page. Included for information only.

LIMITING SAFETY SYSTEM SETTINGS BASES Undervoltage and Underfrequency - Reactor Coolant Pump Buses (Continued) at approximately 10% of full power equivalent); and on increasing power, reinstated automatically by P-7.

Turbine Trip A Turbine trip initiates a Reactor trip. On decreasing power the Reactor trip from the Turbine trip is automatically blocked by the loss of P-7 (a power level of approximately 10% of RATED THERMAL POWER or a turbine inlet pressure at approximately 10% of full power equivalent);

and on increasing power, reinstated automatically by P-7.

Safety Injection Input from ESF If a Reactor trip has not already been generated by the Reactor Trip System instrumentation, the ESF automatic actuation logic channels will initiate a Reactor trip upon any signal which initiates a Safety Injection. The ESF instrumentation channels which initiate a Safety Injection signal are shown in Table 3.3-3.

Reactor Trip System Interlocks The Reactor Trip System interlocks perform the following functions:

P-6 On increasing power P-6 allows the manual block of the Source Range trip (i.e.,

prevents premature block of Source Range trip), and deenergizes the high voltage to the detectors. On decreasing power, Source Range Level trips are automatically reactivated and high voltage restored.

P-7 On increasing power P-7 automatically enables Reactor trips on low flow in more than one reactor coolant loop, reactor coolant pump motor undervoltage and underfrequency, turbine trip, pressurizer low pressure and pressurizer high level.

On decreasing power, the above listed trips are automatically blocked.

P-8 On increasing power, P-8 automatically enables Reactor trips on low flow in one or more reactor coolant loops. On decreasing power, the P-8 automatically blocks the above listed trips.

P-10 On increasing power, P-10 allows the manual block of the Intermediate Range trip and the Low Setpoint Power Range trip; and automatically blocks the Source Range trip and deenergizes the Source Range high voltage power. On decreasing power, the Intermediate Range trip and the Low Setpoint Power Range trip are automatically reactivated. Provides input to P-7.

P-13 Provides input to P-7.

SHEARON HARRIS - UNIT 1 B 2-7 Amendment No. 139

REACTOR COOLANT SYSTEM BASES SAFETY VALVES (Continued) overpressure condition which could occur during shutdown. In the event that no safety valves are OPERABLE, an operating RHR loop, connected to the RCS, provides overpressure relief capability and will prevent RCS overpressurization. In addition, the Overpressure Protection System provides a diverse means of protection against RCS overpressurization at low temperatures.

During operation, all pressurizer Code safety valves must be OPERABLE to prevent the RCS from being pressurized above its Safety Limit of 2735 psig. The combined relief capacity of all of these valves is greater than the maximum surge rate resulting from a complete loss-of-load assuming no reactor trip until the second Reactor Trip System trip setpoint is reached (i.e., no credit is taken for a direct Reactor trip on the loss-of-load) and also assuming no operation of the power-operated relief valves or steam dump valves.

Demonstration of the safety valves' lift settings will occur only during shutdown and will be performed in accordance with the provisions of Section XI of the ASME Boiler and Pressure Code.

INSERT A 3/4.4.3 PRESSURIZER The limit on the maximum water level in the pressurizer assures that the parameter is maintained within the normal steady-state envelope of operation assumed in the SAR. The limit is consistent with the initial SAR assumptions. The 12-hour periodic surveillance is sufficient to ensure that the parameter is restored to within its limit following expected transient operation. The maximum water level also ensures that a steam bubble is formed and thus the RCS is not a hydraulically solid system. The requirement that a minimum number of pressurizer heaters be OPERABLE enhances the capability of the plant to control Reactor Coolant System pressure and establish natural circulation.

3/4.4.4 RELIEF VALVES In MODES 1, 2, and 3 the power-operated relief values (PORVs) provide an RCS pressure boundary, manual RCS pressure control for mitigation of accidents, and automatic RCS pressure relief to minimize challenges to the safety valves.

Providing an RCS pressure boundary and manual RCS pressure control for mitigation of a steam generator tube rupture (SGTR) are the safety-related functions of the PORVs in MODES 1, 2, and

3. The capability of the PORV to perform its function of providing an RCS pressure boundary requires that the PORV or its associated block valve is closed. The capability of the PORV to perform manual RCS pressure control for mitigation of a SGTR accident is based on manual actuation and does not require the automatic RCS pressure control function. The automatic RCS pressure control function of the PORVs is not a safety-related function in MODES 1, 2, and 3. The automatic pressure control function limits the number of challenges to the safety valves, but the safety valves perform the safety function of RCS overpressure protection. Therefore, the automatic RCS pressure control function of the PORVs does not have to be available for the PORVs to be operable.

SHEARON HARRIS - UNIT 1 B 3/4 4-2 Amendment No. 109

INSERT A In MODES 1, 2 and 3 the LCO requirement for a steam bubble is reflected implicitly in the accident analyses. Safety analyses performed for lower MODES are not limiting. All analyses performed from a critical reactor condition assume the existence of a steam bubble and saturated conditions in the pressurizer. In making this assumption, the analyses neglect the small fraction of non-condensable gases normally present.

Safety analyses presented in the FSAR do not take credit for pressurizer heater operation; however, an implicit initial condition assumption of the safety analyses is that the RCS is operating at normal pressure.

The maximum pressurizer water level limit, which ensures that a steam bubble exits in the pressurizer, is an initial condition for the RCS overpressurization that occurs during Turbine Trip in MODE 1. The initial pressurizer water level for other FSAR events is in accordance with applicable methodologies. This satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii). Although the heaters are not specifically used in accident analysis, the need to maintain subcooling in the long term during loss of offsite power, as indicated in NUREG-0737, is the reason for providing an LCO.

U.S. Nuclear Regulatory Commission Serial HNP-15-038 SERIAL HNP-15-038 ENCLOSURE 5 AFFIDAVIT SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 DOCKET NO. 50-400 RENEWED LICENSE NUMBER NPF-63

Enclosure 5 HNP-15-038 AFFIDAVIT of Benjamin C. Waldrep

1. I am Vice President of Harris Nuclear Plant, and as such have the responsibility of reviewing the proprietary information sought to be withheld from public disclosure in connection with nuclear plant licensing and am authorized to apply for its withholding on behalf of Duke Energy.

2. I am making this affidavit in conformance with the provisions of 10 CFR 2.390 of the regulations of the Nuclear Regulatory Commission (NRC) and in conjunction with Duke Energy's application for withholding which accompanies this affidavit.

3. I have knowledge of the criteria used by Duke Energy in designating information as proprietary or confidential. I am familiar with the Duke Energy information contained in the proprietary version of the Harris Turbine Trip Methodology Qualification.

4. Pursuant to the provisions of paragraph (b)(4) of 10 CFR 2.390, the following is furnished for consideration by the NRC in determining whether the information sought to be withheld from public disclosure should be withheld.

(i) The information sought to be withheld from public disclosure is owned by Duke Energy and has been held in confidence by Duke Energy and its consultants.

(ii) The information is of a type that would customarily be held in confidence by Duke Energy. Information is held in confidence if it falls in one or more of the following categories.

(a) The information requested to be withheld reveals distinguishing aspects of a process (or component, structure, tool, method, etc.) whose use by a vendor or consultant, without a license from Duke Energy, would constitute a competitive economic advantage to that vendor or consultant.

(b) The information requested to be withheld consist of supporting data, including test data, relative to a process (or component, structure, tool, method, etc.), and the application of the data secures a competitive economic advantage for example by requiring the vendor or consultant to perform test measurements, and process and analyze the measured test data.

(c) Use by a competitor of the information requested to be withheld would reduce the competitor's expenditure of resources, or improve its competitive position, in the design, manufacture, shipment, installation assurance of quality or licensing of a similar product.

(d) The information requested to be withheld reveals cost or price information, production capacities, budget levels or commercial strategies of Duke Energy or its customers or suppliers.

(e) The information requested to be withheld reveals aspects of the Duke Energy funded (either wholly or as part of a consortium ) development plans or programs of commercial value to Duke Energy.

Page 1 of 3

Enclosure 5 HNP-15-038 (f) The information requested to be withheld consists of patentable ideas.

The information in this presentation is held in confidence for the reasons set forth in paragraphs 4(ii)(a) and 4(ii)(c) above. Rationale for holding this information in confidence is that public disclosure of this information would provide a competitive advantage if the information was used by vendors or consultants without a license from Duke Energy. Public disclosure of this information would diminish the information's marketability, and its use by a vendor or consultant would reduce their expenses to duplicate similar information. The information consists of analysis methodology details, analysis results, supporting data, and aspects of development programs, relative to a method of analysis that provides a competitive advantage to Duke Energy.

(iii) The information was transmitted to the NRC in confidence and under the provisions of 10 CFR 2.390, it is to be received in confidence by the NRC.

(iv) The information sought to be protected is not available in public to the best of our knowledge and belief.

(v) The proprietary information sought to be withheld is that which is marked in the proprietary version of the Harris Turbine Trip Methodology Qualification. This information enables Duke Energy to:

(a) Support the license amendment request for changes to Technical Specifications 2.2.1, 3.4.3, and 3.7.1.1) for Shearon Harris Nuclear Power Plant, Unit 1 (HNP).

(b) Support turbine trip analysis calculations for its HNP.

(vi) The proprietary information sought to be withheld from public disclosure has substantial commercial value to Duke Energy.

(a) Duke Energy uses this information to reduce vendor and consultant expenses associated with supporting the operation and licensing of nuclear power plants.

(b) Duke Energy can sell the information to nuclear utilities, vendors, and consultants for the purpose of supporting the operation and licensing of nuclear power plants.

(c) The subject information could only be duplicated by competitors at similar expense to that incurred by Duke Energy.

5. Public disclosure of this information is likely to cause harm to Duke Energy because it would allow competitors in the nuclear industry to benefit from the results of a significant development program without requiring a commensurate expense or allowing Duke Energy to recoup a portion of its expenditures or benefit from the sale of the information.

Page 2 of 3

Enclosure 5 HNP-15-038 Benjamin C. Waldrep affirms that he is the person who subscribed his name to the foregoing statement, and that all the matters and facts set forth herein are true and correct to the best of his knowledge.

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

Executed on /i/;:r/!J' r I

¥twD Benjamin C. Waldrep Page 3 of 3

Enclosure 5 HNP-15-038 Benjamin C. Waldrep affirms that he is the person who subscribed his name to the foregoing statement, and that all the matters and facts set forth herein are true and correct to the best of his knowledge.

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

Executed on __________________.

Benjamin C. Waldrep Page 3 of 3

U.S. Nuclear Regulatory Commission Serial HNP-15-038 Attachment A SERIAL HNP-15-038 ATTACHMENT A HNP-I/INST-1010, EVALUATION OF RTS/ESFAS TECH SPEC RELATED SETPOINTS, ALLOWABLE VALUES, AND UNCERTAINTIES, TABLE 3.8 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 DOCKET NO. 50-400 RENEWED LICENSE NUMBER NPF-63

CALCULATION NO. HNP-I/INST-1010 PAGE 47 REV. 6 TABLE 3-8 PRESSURIZER WATER LEVEL - HIGH Summary of CSA and Five-Column Tech Spec Terms Revision 6 to this calculation supports a proposed License Amendment Request (LAR) affecting the RTS Pressurizer Water Level - High function in HNP Technical Specification Table 2.2-1. Reference 16 provides the driver for this change, by reducing the Safety Analysis Limit for this function from 100% to 95%. LARs impacting RTS Setpoints are currently required to accommodate TSTF-493 criteria (as described in Reference EGR-NGGC-0153, Section 9.10). The computation below:

Establishes new limits for the Trip Setpoint and Allowable Value.

Confirms that the associated TA, Z and S values are unchanged.

Relative to TSTF-493, establishes supplemental 1.5% "as-found" and 'as-left" tolerance criteria (tighter than Allowable Values) to be incorporated into Surveillance Tests for the transmitters and rack portions of the instrument channels. (Values outside these tolerances invoke specific actions required per TSTF-493).

The computations below are based upon the equations shown per Table 1-2 herein.

Values for the specific uncertainty terms are derived from Reference 2.9.d.

CSA = [ (PEA)2 + (SMTE + SD)2 + (STE)2 + (SPE)2 + (SCA + SMTE)2 + (SRA)2 +

(RMTE + RD)2 + (RTE)2 + (RCA + RMTE)2 ]1/2 + PMAPressure + PMARefLegTemp

= [ (0.00)2 + (0.56 + 1.25)2 + (0.50)2 + (0.50)2 + (0.50 + 0.56)2 + (0.50)2 +

(0.20 + 1.00)2 + (0.50)2 + (0.50 + 0.20)2 ]1/2 + 0.87 + 1.68

= 5.25 % Span [Reference 2.9.d & Reference 2.8 (WCAP Table 3-8)]

TS = 87.0 % Level Span [As of Revision 6, this TS value is selected to maintain the previous TA value of 8.0% Span; i.e., maintains previous margin of 2.75% Span)]

SAL = 95.0 % Level Span [Reference 16, Appendix N]

TA = { ( SAL - TS ) / 100 % level } x 100 % Span = 8.0 % Span Margin = TA - CSA = 2.75 % Span S = { (SD) + (SCA) } = { (1.25) + (0.5) } = 1.75 % Span Z = (A)1/2 + Biases = { (PEA)2 + (SPE)2 + (STE)2 + (RTE)2 }1/2 + EA + PMABiases

= { 02 + (0.5)2 + (0.5)2 + (0.5)2 }1/2 + 0 + (0.87 + 1.68) = 3.416 % Span R = T is the lesser of:

T1 = { RD + RCA } = { (1.0) + (0.5) } = 1.50 % Span T2 = TA - S - Z = 8.00 - 1.75 - 3.42 = 2.83 % Span AV = { TS + [ R/100%Span ] x 100 % Level }

= { 87.0% + [ 1.5/100%Span ] x 100 % Level } 88.5 % Level Span The new SAL, TS and AV values established above supersede the corresponding values previously established in Reference 2.9.d. All other assumptions, data, calculation and conclusions stated Reference 2.9.d remain applicable.

CALCULATION NO. HNP-I/INST-1010 PAGE 47A REV. 6 TABLE 3-8 (Cont'd)

PRESSURIZER WATER LEVEL - HIGH Summary of CSA and Five-Column Tech Spec Terms A comparison of current and proposed values 5 Column Tech Spec values are summarized as follows:

Tech Spec Term Value Value (Current per TS (Proposed change Amend 139) per Rev. 6)

Total Allowance (TA) 8.0 % Span 8.0 % Span Z Term 3.42 % Span 3.42 % Span Sensor Error (S) 1.75 % Span 1.75 % Span Trip Setpoint (TS) < 92.0 % level span < 87.0 % level span Allowable Value (AV) < 93.5 % level span < 88.5 % level span TSFT-493 Supplemental As-Found and As-Left Tolerances for Surveillance Testing:

For sensor-only surveillance:

ALT (As-Left Tolerance) = +/- Sensor Reference Accuracy [Ref. 7, Sect.9.10.4]

= SRA = +/- 0.50 % Span AFT (As-Found Tolerance) = +/- (ALT2 + Sensor Drift2 + M&TE2) 1/2

[Ref. 7, Sect.9.10.3]

= [ (ALT)2 + (SD)2 + (SMTE)2 ]1/2

= [ (0.50)2 + (1.25)2 + (0.56)2 ]1/2 = +/- 1.46 % Span For rack-only surveillance:

ALT (As-Left Tolerance) = +/- Rack Calibration Accuracy [Ref. 7, Sect.9.10.4]

= RCA = +/- 0.50 % Span AFT (As-Found Tolerance) = +/- (ALT2 + Rack Drift2 + M&TE2) 1/2

[Ref. 7, Sect.9.10.3]

= [ (ALT)2 + (RD)2 + (RMTE)2 ]1/2

= [ (0.50)2 + (1.00)2 + (0.20)2 ]1/2 = +/- 1.14 % Span In cases, where the above ALT or AFTs are not satisfied during a Surveillance Test, the response actions defined in Reference 7, Section 9.10.6 apply.

U.S. Nuclear Regulatory Commission Serial HNP-15-038 Attachment B SERIAL HNP-15-038 ATTACHMENT B HARRIS TURBINE TRIP ANALYSIS TO ADDRESS REVISED SAFETY VALVE TOLERANCES SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 DOCKET NO. 50-400 RENEWED LICENSE NUMBER NPF-63

Attachment B Attachment B - Harris Turbine Trip Analysis to Address Revised Safety Valve Tolerances B.

1.0 INTRODUCTION

The Harris FSAR Section 15.2.3 turbine trip analysis is reanalyzed to evaluate changes to the primary and secondary system safety valve tolerances. Two cases are analyzed for this event:

one challenging the primary overpressurization criterion and one challenging the secondary system overpressurization criterion. In addition, a sensitivity case is performed to confirm the requirements of Standard Review Plan, Chapter 5.2.2 - Overpressure Protection, continue to be satisfied. An evaluation of the DNB analysis is also performed.

B.2.0 ANALYSIS METHOD The turbine trip reanalysis is performed using the RETRAN-3D computer code. The RETRAN-3D modeling is based on previously approved Duke Energy methodology DPC-NE-3000-P-A (Reference B-1) with minor changes as described in Attachment C. The Harris RETRAN-3D plant model is assessed against the existing AREVA turbine trip analysis of record as described in Attachment C. Good agreement was obtained for the transient sequence of events, the system parameter responses, and the peak primary and secondary pressure results. This benchmark analysis provides confidence that the RETRAN computer code and model are adequate to assess the impact of changes in the safety valve tolerances for the turbine trip event.

The benchmarked Harris RETRAN-3D model is then modified to evaluate changes to the safety valve tolerance for the two turbine trip events (primary system overpressure and secondary system overpressure cases). An example of a change to the benchmarked input model is to include modeling of the individual steam lines according to the Harris plant configuration (refer to Attachment C, Section C.2.0).

The neutronics parameters are updated to use more representative values for MTC, DTC, ,

and /l (Table B-1). Other initial conditions, such as pressurizer pressure, reactor vessel Tavg, and total RCS flow, are selected based on guidance provided by Duke Energy methodology report DPC-NE-3002-A (Reference B-2). Sensitivity calculations are also performed to ensure conservative input values are selected. The conservative initial conditions used in these analyses are presented in Table B-2.

Page B-1 of B-19

Attachment B B.3.0 ANALYSIS The trip setpoints and time delays assumed in the analysis of this event are unchanged for those provided in FSAR Table 15.0.6-2 (Reference B-3) with the exception of the pressurizer high level trip setpoint. Other major assumptions adopted from the analysis of record are listed below:

1. Reactor Control - From the standpoint of the maximum pressures attained it is conservative to assume that the reactor is in manual control. If the reactor were in automatic control, the control rod banks would insert prior to trip and reduce the severity of the transient.

2. Steam Release - No credit is taken for the operation of the Steam Dump System or steam generator Power Operated Relief Valves (PORVs).

3. Pressurizer Spray and Power Operated Relief Valves:

a. For the secondary side overpressurization and the MDNBR cases, the pressurizer spray and power operated relief valves are conservatively assumed to operate in reducing or limiting the reactor coolant pressure. The Pressurizer Safety Valves (PSVs) are also available.

b. For the primary side overpressurization case, no credit is taken for the effect of pressurizer spray and power operated relief valves in reducing or limiting the reactor coolant pressure. The Pressurizer Safety Valves are operable.

4. Feedwater Flow - Main feedwater flow to the steam generators is assumed to be lost at the time of turbine trip. No credit is taken for auxiliary feedwater flow since a stabilized plant condition will be reached before auxiliary feedwater initiation would occur.

5. No credit is taken for the reactor trip on the turbine trip. Trip signals are expected due to high pressurizer pressure, over-temperature T, high neutron flux, high pressurizer water level, and low-low steam generator water level.

The revised turbine trip analysis evaluated a pressurizer and main steam safety valve tolerance of +/-3%. The specific input parameters used and bias assumptions are provided in Table B-3.

B.4.0 RESULTS B.4.1 Primary System Overpressurization Case The event summary is provided in Table B-4. Figure B-1 shows the transient reactor power.

Before the reactor trips, the reactor power is fairly constant at 2958 MWth with a slight power decrease resulting from the primary system heatup. The reactor trips at 7.07 seconds on high pressurizer pressure. After the reactor trips, the reactor power decreases quickly as the control rods insert.

Figure B-2 presents RCS temperatures, Thot, Tcold, and Tavg, for a representative loop. Given that the event is a symmetric transient, all loops behave in a similar manner. The cold leg temperature begins to increase soon after turbine trip due to the degraded primary-to-secondary Page B-2 of B-19

Attachment B heat transfer. The loop average temperature increases and reaches its maximum condition at about 10.5 seconds, then it decreases steadily until the end of the simulation.

The indicated pressurizer level (Figure B-3) increases after the turbine trips. The level increases after turbine trip as the reactor coolant expands and causes an in-surge of water to the pressurizer. At about 12 seconds, the pressurizer level begins decreasing from reactor coolant shrinkage caused by the reactor trip. The pressurizer level decreases steadily until the end of the calculation.

The primary pressure (Figure B-4) increases after turbine trip from the reduced heat transfer to the steam generators. The pressure increase causes PSVs to start opening at 6.9 seconds which is shortly before the occurrence of the peak primary pressure. The PSVs close at 9.4 seconds. The bottom of the reactor vessel is the location of the maximum peak primary pressure. The peak primary pressure of 2738.25 psia, which is below the acceptance criterion of 2750 psia, occurs at 7.8 seconds.

B.4.1.1 Sensitivity Case to Evaluate SRP Chapter 5.2.2 In accordance with Standard Review Plan (SRP) Chapter 5.2.2 Section II.3.B.iii (Reference B-4), a sensitivity case is performed assuming reactor trip on the second safety-grade trip from the reactor protection system. As a result, no credit is taken for the high pressurizer pressure reactor trip. All other assumptions are identical to the case described in Section B.4.1. The event summary is provided in Table B-5. The primary pressure (Figure B-5) increases after turbine trip due to reduced heat transfer to the steam generators. The pressure increase causes the PSVs to start opening at 6.9 seconds which is shortly before the occurrence of the peak primary pressure. The PSVs open and close twice during the event. Reactor trip signal on high pressurizer level is reached at 10.91 seconds, and the reactor trip occurs at 12.91 seconds. The peak primary pressure for the sensitivity case is 2738.7 psia and occurs at 7.85 seconds. The peak pressure for this case is slightly higher than the case with credit for the high pressurizer pressure trip but still below the acceptance criterion of 2750 psia.

B.4.2 Secondary System Overpressurization Case The event summary is provided in Table B-6. Figure B-6 shows the transient reactor power.

Before the reactor trips, the reactor power shows a mild power decrease caused by the negative temperature feedback resulting from the primary system heatup. The reactor trips at 11.34 seconds on the high pressurizer level trip. After the reactor trips, the reactor power decreases quickly as the control rods insert.

Figure B-7 presents RCS temperatures, Thot, Tcold, and Tavg, for a representative loop. The loop average temperature increases and reaches its maximum condition at about 14 seconds, then it decreases throughout the end of the simulation.

The indicated pressurizer level (Figure B-8) increases after the turbine trips. At about 14 seconds, the pressurizer level reaches 99.0% and begins decreasing due to reactor coolant shrinkage resulting from reactor trip. The pressurizer level decreases steadily until the end of the simulation.

The pressurizer pressure increases after turbine trip due to degraded heat transfer to the steam generators. Given the high initial pressurizer pressure assumed in this case, pressurizer spray Page B-3 of B-19

Attachment B initiates at the start of the analysis. As pressure increases, it causes the pressurizer PORVs to start opening and closing (cycling) at 2.4 seconds. The pressure response reflects the cycling of PORVs. After the reactor trips, the pressurizer pressure begins to decrease and continues decreasing until the end of the simulation. The PSVs are not challenged due to opening of the pressurizer PORVs.

The SG secondary system pressure increases rapidly upon turbine trip (Figure B-9) until the Main Steam Safety Valves (MSSVs) begin opening to relieve pressure and the reactor trip decreases heat generation. At 17.3 seconds, the pressure reaches its peak. The peak secondary pressure of 1304.66 psia, which is below the acceptance criterion of 1320 psia, occurs at the bottom of steam generator downcomer.

B.4.3 DNB Evaluation The PSVs are not challenged during the system DNB analysis due to the operation of the pressurizer PORVs. Sensitivity cases with different MSSV drift setpoints have been performed using the HNP RETRAN-3D model for the DNB analysis. The sensitivity study result shows that the DNBR results are insensitive to the MSSV setpoints. Thus, it is concluded that the MSSV setpoint tolerance has no negative impact on the MDNBR results, and the current HNP DNB analysis remains valid for the MSSV setpoint tolerance change.

B.

5.0 CONCLUSION

The turbine trip event has been reanalyzed to evaluate changes to the primary and secondary system safety valve tolerances. The cases analyzed demonstrate that the acceptance criteria are satisfied assuming a safety valve tolerance of +/- 3%. In addition, the sensitivity case presented in Section B.4.1.1 demonstrates that the peak primary overpressure criterion is met, and therefore the design and sizing of the pressurizer safety valves meets the overpressure design criterion cited in SRP Chapter 5.2.2.

B.

6.0 REFERENCES

B-1. Duke Energy Topical Report DPC-NE-3000-PA, Oconee Nuclear Station, McGuire Nuclear Station, Catawba Nuclear Station, Thermal-Hydraulic Transient Analysis Methodology, Revision 5a, October 2012.

B-2. Duke Energy Topical Report DPC-NE-3002-A, UFSAR Chapter 15 System Transient Methodology, Revision 4b, September 2010.

B-3. Shearon Harris Nuclear Generation Station FSAR Section 15.2.3.

B-4. NUREG-0800, Standard Review Plan for the Review of Safety Analysis Report for Nuclear Power Plants, Section 5.2.2, Revision 3, March 2007.

Page B-4 of B-19

Attachment B Table B-1 Range of Neutronic Parameters supported by the Turbine Trip Analysis Primary Side Secondary Side Overpressure Overpressure Parameter Case Value Case Value Moderator temperature Coefficient (pcm/ oF) 0.0 0.0 Doppler Coefficient (pcm/oF) -0.90 -0.90 0.007 0.007

/ l * (sec-1) 466.67 466.67 Shutdown Margin (pcm) 1770 1770 Page B-5 of B-19

Attachment B Table B-2 Initial Conditions Used in Turbine Trip Analysis Secondary Primary System System Overpressure Overpressure Parameter Case Value Case Value Initial reactor power (MW) 2958 2958 Initial pressurizer pressure (psia) 2212 2300 Initial pressurizer level (% of level span) 81.75 81.75 Initial reactor vessel Tavg (oF) 580.8 595.6 Initial total RCS flow rate (lbm/sec) 30854 32912 (293,540 gpm) (321,300 gpm)

Initial steam generator pressure (psia) 928 1050 Initial feedwater flowrate per SG 1210.8 1217.5 (lbm/sec)

Feedwater temperature (oF) 440 440 Page B-6 of B-19

Attachment B Table B Input Parameters and Bias Assumption for the Turbine Trip Primary Secondary Overpressurization Overpressurization Parameter Case Case 2958 MW 2958 MW Core power (MUR + 0.34%) (MUR + 0.34%)

Pressurizer pressure Nominal - Uncertainty Nominal + Uncertainty Pressurizer level 75% + Uncertainty 75% + Uncertainty o

Reactor vessel Tavg Minimum (580.8 F) Nominal + Uncertainty RCS flow rate Tech Spec minimum Design maximum Steam generator pressure Low High Initial feedwater flow rate Nominal Nominal Feedwater temperature Nominal Nominal Steam generator NR level Nominal High Cycle exposure BOC BOC o o Moderator Temperature Coefficient 0.0 pcm/ F 0.0 pcm/ F o o Doppler coefficient -0.9 pcm/ F (BOC least negative) -0.9 pcm/ F (BOC least negative)

Delayed neutron Maximum at BOC Maximum at BOC fraction,

-1 -1

/l* 466.67 sec 466.67 sec Minimum (bounds the most Minimum (bounds the most Reactor trip reactivity insertion reactive rod stuck out of the core) reactive rod stuck out of the core)

Pellet-to-cladding heat transfer coefficient N/A N/A Average core fuel temperature BOC maximum BOC maximum Steam generator tube plugging Maximum Minimum High pressurizer level trip setpoint 87% + allowance = 95% 87% + allowance = 95%

Pressurizer SV setpoint PSVs with 3% drift Nominal + tolerance Pressurizer PORV setpoints Disabled Nominal MSSV setpoints Banks 1-5 with 3% drift Banks 1-5 with 3% drift Rod position controller Manual Manual Pressurizer heaters Available Available Pressurizer spray Disabled Available Main feedwater Auto Auto Auxiliary feedwater Disabled Disabled Page B-7 of B-19

Attachment B Table B-4 Event Summary for Turbine Trip Primary System Overpressurization Event Time (sec)

Turbine trips 0.0 Pressurizer high pressure trip signal reached 5.07 PSVs open 6.9 Reactor trips on pressurizer high pressure (rod motion starts) 7.07 Peak primary pressure at bottom of reactor vessel reached 7.84 PSVs close 9.4 Bank 1 MSSVs open 12.9 Bank 2 MSSVs open 14.0 End of simulation 60.0 Page B-8 of B-19

Attachment B Table B-5 Event Summary for Turbine Trip Primary System Overpressurization With Pressurizer High Pressure Trip Unavailable Event Time (sec)

Turbine trips 0.0 PSVs open 6.9 Peak primary pressure at bottom of reactor vessel reached 7.85 PSVs close 9.5 PSVs open 10.0 High pressurizer level trip signal reached 10.91 Bank 1 MSSVs open 12.7 Reactor trips on high pressurizer level (control rod motion starts) 12.91 Bank 2 MSSVs open 13.6 Bank 3 MSSVs open 14.9 PSVs close 16.3 Bank 4 MSSVs open 17.1 Bank 4 MSSVs close 39.0 Bank 3 MSSVs close 41.7 Bank 2 MSSVs close 47.3 Auxiliary feedwater on lo-lo level 58.3 End of simulation 60.0 Page B-9 of B-19

Attachment B Table B-6 Event Summary for Turbine Trip Secondary System Overpressurization Event Time (sec)

Turbine trips 0.0 Pressurizer spray initiates 0.0 Pressurizer compensated and non-compensated PORVs open and cycle 2.4 Bank 1 MSSVs open 5.4 Bank 2 MSSVs open 6.2 Bank 3 MSSVs open 7.3 High pressurizer level trip signal reached 9.34 Bank 4 MSSVs open 10.1 Reactor trips on high pressurizer level (rod motion starts) 11.34 Bank 5 MSSVs open 14.6 Pressurizer non-compensated and compensated PORVs close 15.2 Pressurizer spray terminates 15.4 Peak secondary pressure occurs at bottom of the SG downcomer 17.30 Bank 5 MSSVs close 33.4 Bank 4 MSSVs close 35.3 Bank 3 MSSVs close 38.3 Bank 2 MSSVs close 46.7 AFW on lo-lo SG level 57.9 End of simulation 60.0 Page B-10 of B-19

Attachment B Figure B-1 Peak Primary Side Overpressurization Case Reactor Power 3500 3000 PSVs with +3% Drift MSSVs with +3% Drift 2500 Reactor Power (MWt) 2000 1500 1000 500 0

0 10 20 30 40 50 60 Time (seconds)

Page B-11 of B-19

Attachment B Figure B-2 Peak Primary Side Overpressurization Case RCS Temperature 640 PSVs with +3% Drift 620 MSSVs with +3% Drift RCS Temperature (oF) 600 580 Thot 560 Tcold Tavg 540 0 10 20 30 40 50 60 Time (seconds)

Page B-12 of B-19

Attachment B Figure B-3 Peak Primary Side Overpressurization Case Pressurizer Level 100 90 80 70 60 PSVs with +3% Drift Pressurizer Level (%)

MSSVs with +3% Drift 50 40 30 20 10 0

0 10 20 30 40 50 60 Time (seconds)

Page B-13 of B-19

Attachment B Figure B-4 Peak Primary Side Overpressurization Case Primary Pressure 2800 Primary Pressurization Acceptance Criterion = 2750 psia 2600 PSVs with +3% Drift MSSVs with +3% Drift 2400 Pressure (psia) 2200 2000 Pressurizer Pressure (psia)

Pressure at Bottom of Reactor Vessel (psia) 1800 1600 0 10 20 30 40 50 60 Time (seconds)

Page B-14 of B-19

Attachment B Figure B-5 Peak Primary Side Overpressurization SRP 5.2.2 Sensitivity Case Primary Pressure 2800 Primary Pressurization Acceptance Criterion = 2750 psia 2600 2400 Pressure (psia)

PSVs with +3% Drift 2200 MSSVs with +3% Drift 2000 1800 Pressurizer Pressure (psia)

Pressure at Bottom of Reactor Vessel (psia) 1600 0 10 20 30 40 50 60 Time (seconds)

Page B-15 of B-19

Attachment B Figure B-6 Peak Secondary Side Overpressurization Case - Reactor Power 3500 3000 MSSVs with +3% Drift 2500 Reactor Power (MWt) 2000 1500 1000 500 0

0 10 20 30 40 50 60 Time (seconds)

Page B-16 of B-19

Attachment B Figure B-7 Peak Secondary Side Overpressurization Case - RCS Temperature 640 MSSVs with +3% Drift 620 RCS Temperature (oF)

Thot 600 Tcold Tavg 580 560 0 10 20 30 40 50 60 Time (seconds)

Page B-17 of B-19

Attachment B Figure B-8 Peak Secondary Side Overpressurization Case - Pressurizer Level 100 90 MSSVs with +3% Drift 80 70 60 Pressurizer Level (%)

50 40 30 20 10 0

0 10 20 30 40 50 60 Time (seconds)

Page B-18 of B-19

Attachment B Figure B-9 Peak Secondary Side Overpressurization Case - Pressure 1400 Secondary Pressurization Acceptance Criterion = 1320 psia 1300 Pressure at Bottom of Steam Generator Downcomer (psia)

MSSVs with +3% Drift 1200 1100 1000 900 800 0 10 20 30 40 50 60 Time (seconds)

Page B-19 of B-19

U.S. Nuclear Regulatory Commission Serial HNP-15-038 Attachment D SERIAL HNP-15-038 ATTACHMENT D HARRIS TURBINE TRIP METHODOLOGY QUALIFICATION (REDACTED)

SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 DOCKET NO. 50-400 RENEWED LICENSE NUMBER NPF-63

Non-Proprietary Version of Attachment C Attachment D Attachment D - Harris Turbine Trip Methodology Qualification (Non-Proprietary Version)

D.

1.0 INTRODUCTION

The methodology report DPC-NE-3000-PA, hereafter DPC-NE-3000, presents the development and qualification of Dukes thermal-hydraulic models for transient analysis (Reference D-1). DPC-NE-3000 describes RETRAN and VIPRE-01 models for the Oconee (ONS), McGuire (MNS), and Catawba Nuclear Stations (CNS) and qualifies these models for licensing applications.

The material presented herein provides a description of the RETRAN-3D plant model for Shearon Harris Nuclear Power Plant, Unit 1 (HNP). The RETRAN-3D model for HNP is similar to the MNS and CNS models presented in DPC-NE-3000. The RETRAN-3D model for HNP is evaluated by comparing RETRAN-3D calculations to the analysis of record (AOR) for the turbine trip event (Reference D-2). The conditions and limitations in the NRCs generic Safety Evaluation Report (SER) for the RETRAN-3D computer code (Reference D-3) are evaluated for the HNP RETRAN-3D model as applied for the turbine trip analysis. Together, these evaluations qualify the HNP RETRAN-3D model to perform the FSAR analysis for the turbine trip event.

D.1.1 Proprietary Notice Certain data in this report is proprietary to Duke Energy. Proprietary data is denoted by brackets in text, tables and figures, and is deleted. Footnote letters associated with bracketed information refer to categories of proprietary information described in Section 4(ii) of the included Affidavit.

D.1.2 Evaluation of the RETRAN-3D SER Conditions and Limitations RETRAN-3D was developed to enhance and extend the simulation capabilities of the RETRAN-02 code. Some of the improvements include a three-dimensional reactor kinetics model, improved two-phase models, an improved heat transfer correlation package, and an implicit numerical solution method. RETRAN-3D was approved by the NRC staff in Reference D-3 with 45 limitations and conditions of use. Subsequent updates to RETRAN-3D add new features as well as correct errors (Reference D-4).

Page D-1 of D-17

Non-Proprietary Version of Attachment C Attachment D The limitations and conditions of use described in the NRCs generic SER for the RETRAN-3D computer code (Reference D-3) are assessed for the HNP RETRAN-3D model as applied for the turbine trip analysis. The assessment is organized into two categories as described below.

1) Limitations and conditions of use considered not applicable for the HNP turbine trip analysis or for which the NRC staff or previous Duke Energy resolutions apply (refer to Reference D-1, Appendix C).

2) HNP-specific evaluations of the limitations and conditions of use for which further explanation is warranted (8 total).

a. Condition 14: The HNP RETRAN-3D model uses [

]a, c. This usage is consistent with the NRC Staff Position.

b. Condition 16: The HNP RETRAN-3D model uses an algebraic equation for velocity difference based on the Chexal-Lellouche drift flux correlation. The HNP RETRAN-3D model applies the algebraic slip model [

]a, c. For a turbine trip analysis, there are negligible effects on the results by [

]a, c.

c. Condition 18: In the HNP RETRAN-3D model, wall heat transfer is modeled in the pressurizer. This usage is consistent with the NRC Staff position.

d. Condition 20: The HNP RETRAN-3D model for the reactor coolant pumps uses the same pump homologous curves as MNS, described in Reference D-1, Section 3.2.6.2. MNS and HNP have Westinghouse Model 93A reactor coolant pumps with similar characteristics.

e. Condition 24: The HNP RETRAN-3D model configures the [

]a, c differently from the MNS and CNS models described in DPC-NE-3000 (Reference D-1, Section 3.2.2). However, the HNP RETRAN-3D model [

]a, c.

f. Condition 28: The local conditions heat transfer model described in DPC-NE-3000 (Reference D-1, Section 3.2.6.7) is retained in the HNP RETRAN-3D model with the following change. In the HNP RETRAN-3D model, the local conditions heat transfer model is [ ]a, c. As in the MNS and CNS models, the local conditions heat transfer model is [

]a, c.

This usage complies with the limitation or conditions of use.

Page D-2 of D-17

Non-Proprietary Version of Attachment C Attachment D

g. Condition 40: Updates to RETRAN-3D subsequent to DPC-NE-3000 include the addition of new control blocks. The HNP RETRAN-3D model uses the following control blocks, which have not been reviewed previously by the NRC staff.

SSM - Super summer SMN - Super minimum SMX - Super maximum The use of these control block models enhances and simplifies applications. In addition, the accumulator model is changed to incorporate a polytropic expansion model. However, the accumulator actuation setpoints are not reached during the time period of interest for the turbine trip analysis.

h. Condition 45: The HNP turbine trip analysis is submitted for review. The turbine trip analysis is not a best-estimate analysis and includes assumptions commensurate with a conservative, traditional FSAR Chapter 15 analysis.

D.2.0 OVERVIEW OF RETRAN-3D PLANT MODEL This section describes the nodalization of the RETRAN-3D base model for the HNP. Changes for the benchmark analysis and turbine trip reanalysis are discussed in Section D.3.0 and Attachment B, respectively.

Figure D-1 shows the layout of the RETRAN-3D volumes and junctions used to model the primary system. Each of the three reactor coolant loops is modeled explicitly, with X or Y used to designate a corresponding set of volumes or junctions for Loops 1, 2 and 3. [

]a, c.

Figure D-2 shows the layout of the RETRAN-3D volumes and junctions used to model the secondary system. Each of the three steam generators is modeled explicitly, as are the steam lines connecting each steam generator to the common steam header. [

]a, c.

The level of modeling detail shown in Figure D-1 and Figure D-2 is similar to that shown in Figures 3.2-1 to 3.2-3 of Reference D-1 for the McGuire and Catawba Nuclear Stations. In addition to explicit modeling of the three reactor coolant loops, steam generators, and steam lines according to the HNP plant configuration, the other major change is [

]a, c as a, c shown in Figures 3.2-1 to 3.2-3 of Reference D-1 [ ] as shown in Figure D-2. This change [

Page D-3 of D-17

Non-Proprietary Version of Attachment C Attachment D

]a, c.

Not shown in Figure D-1 and Figure D-2 are the RETRAN-3D heat conductors used to represent the fuel rods, steam generator tubes and other structures. Heat conductors are used to evaluate the heat transfer between structures and fluid on the primary and secondary sides and are modeled as having either rectangular or cylindrical geometry. Details such as the heat transfer area, number of radial nodes and thermal properties are determined by the analyst and provided through appropriate specification of the RETRAN-3D input. The level of modeling detail is similar to that described in Tables 3.2-1 and 3.2-2 of Reference D-1, with various changes such as [

]a, c.

D.3.0 RETRAN-3D BENCHMARK ANALYSIS Duke Energy has developed a RETRAN-3D transient analysis model for the HNP as described in the previous section. As a part of the HNP RETRAN-3D model validation, a RETRAN-3D turbine trip benchmark analysis is performed consistently with the HNP Turbine Trip FSAR analysis (Reference D-2). Two cases are analyzed from the AOR: a primary overpressurization case and a secondary overpressurization case.

The turbine trip event is initiated by a rapid closure of the turbine stop valves. The FSAR analysis assumes that a direct reactor trip from turbine trip does not occur, and the reactor trip is delayed until conditions in the RCS cause another reactor protection system trip setpoint to be reached. Only the high pressurizer pressure trip, high pressurizer level trip, high neutron flux trip, low-low steam generator water level, and OTT are credited in the analysis. Main feedwater flow is terminated at the start of the event, and auxiliary feedwater flow is not available during the analysis period. No credit is taken for non-safety grade systems or equipment such as the steam dump system and steam line PORVs. In addition, no credit is taken for the pressurizer PORVs in the peak primary pressure case. Therefore, for the case that challenges the secondary pressure limit, only the main steam safety valves are available for pressure relief; for the case that challenges the primary pressure limit, only the pressurizer safety valves are available for pressure relief.

The HNP RETRAN-3D transient analysis model is an explicit representation of the three RCS and main steam system loops, with the main steam line piping modeled based on actual HNP plant configuration. However, the RETRAN-3D model used in the benchmark differs slightly from the model presented in Section D.2.0. For example, in order to closely simulate the transient response time in the AOR, the main steam lines downstream of the steam header are removed from the RETRAN-3D base model.

In the RETRAN-3D turbine trip benchmark analysis, [

Page D-4 of D-17

Non-Proprietary Version of Attachment C Attachment D

]a, c.

The RETRAN-3D models of pressurizer safety valves and main steam safety valves are justified by comparing the valve flows with the results documented in the AOR. Key parameters for the turbine trip event are compared between the RETRAN-3D calculated value and the AOR value.

The good agreement in the results presented in Table D-1 and Table D-2 and Figure D-3 through Figure D-10 show that the HNP RETRAN-3D model is able to predict transient responses and qualified to perform the FSAR analysis.

D.

4.0 REFERENCES

D-1. Duke Energy Methodology Report DPC-NE-3000-PA, Oconee Nuclear Station, McGuire Nuclear Station, Catawba Nuclear Station, Thermal-Hydraulic Transient Analysis Methodology, Revision 5a, October 2012.

D-2. Shearon Harris Nuclear Power Plant FSAR Section 15.2.3.

D-3. Letter, S. A. Richards (NRC) to G. L. Vine (EPRI), Safety Evaluation Report on EPRI Topical Report NP-7450(P), Revision 4, RETRAN-3D - A Program for Transient Thermal-Hydraulic Analysis of Complex Fluid Flow Systems (TAC No. MA4311),

January 2001.

D-4. RETRAN-3D - A Program for Transient Thermal-Hydraulic Analysis of Complex Fluid Flow Systems, EPRI, NP-7450(A), Volumes 1-4, Rev. 6.3, July 2007.

Page D-5 of D-17

Non-Proprietary Version of Attachment C Attachment D Table D-1 Primary Side Overpressurization - Sequence of Events Time (s)

Event FSAR RETRAN-3D Initiate turbine trip 0.0 0.01 Activate reactor trip signal (high pressure) 5.03 4.76 Pressurizer safety valve setpoint reached 6.5 6.0 Scram Initiation 7.04 6.76 Reach full flow through pressurizer safety 7.6 7.1 valve

Reach peak primary side pressure 7.8 7.8 (FSAR value for peak pressurizer pressure)

Open SG 1st bank MSSVs 8.4 8.8 nd Open SG 2 bank MSSVs 9.3 10.4 rd Open SG 3 bank MSSVs 10.8 11.8 Open SG 4th bank MSSVs th Open SG 5 bank MSSVs

there is a loop seal purge time delay after the setpoint is reached Page D-6 of D-17

Non-Proprietary Version of Attachment C Attachment D Table D-2 Secondary Side Overpressurization - Sequence of Events Time (s)

Event FSAR RETRAN-3D Initiate turbine trip 0.0 0.01 Activate pressurizer spray 1.0 0.9 Open pressurizer compensated PORV 1.2 1.2 Open pressurizer uncompensated PORV 4.3 4.0 st Open SG 1 bank MSSVs 5.4 5.3 nd Open SG 2 bank MSSVs 6.5 5.9 rd Open SG 3 bank MSSVs 7.9 7.0 th Open SG 4 bank MSSVs 10.1 9.7 Activate OTT trip 11.16 12.06 Scram initiation 12.41 13.32 th Open SG 5 bank MSSVs 13.2 13.8 Reach peak pressurizer level 16.2 17.7 Reach peak SG secondary pressure 18.9 19.3 Page D-7 of D-17

Non-Proprietary Version of Attachment C Attachment D Figure D-1 RETRAN-3D Volumes and Junctions for Primary System a, c Page D-8 of D-17

Non-Proprietary Version of Attachment C Attachment D Figure D-2 RETRAN-3D Volumes and Junctions for Secondary System a, c Page D-9 of D-17

Non-Proprietary Version of Attachment C Attachment D Figure D-3 Reactor Power - Primary Side Overpressurization RETRAN-3D Page D-10 of D-17

Non-Proprietary Version of Attachment C Attachment D Figure D-4 Average Temperature - Primary Side Overpressurization Thot_RETRAN-3D Tcold_RETRAN-3D Tavg_RETRAN-3D Page D-11 of D-17

Non-Proprietary Version of Attachment C Attachment D Figure D-5 Primary Pressure - Primary Side Overpressurization Bottom of Lower Plenum RETRAN-3D Pressurizer RETRAN-3D Page D-12 of D-17

Non-Proprietary Version of Attachment C Attachment D Figure D-6 Pressurizer Level - Primary Side Overpressurization RETRAN-3D Page D-13 of D-17

Non-Proprietary Version of Attachment C Attachment D Figure D-7 Reactor Power - Secondary Side Overpressurization RETRAN-3D Page D-14 of D-17

Non-Proprietary Version of Attachment C Attachment D Figure D-8 Average Temperature - Secondary Side Overpressurization Thot_RETRAN-3D Tcold_RETRAN-3D Tavg_RETRAN-3D Page D-15 of D-17

Non-Proprietary Version of Attachment C Attachment D Figure D-9 Pressurizer Level - Secondary Side Overpressurization RETRAN-3D Page D-16 of D-17

Non-Proprietary Version of Attachment C Attachment D Figure D-10 Pressure at Bottom of SG Downcomer - Secondary Side Overpressurization RETRAN-3D Page D-17 of D-17