Text

Report of Changes Pursuant to 10 CFR 50.59 and Summary of Commitment Changes
	ML18122A071
Person / Time
Site:	Harris
Issue date:	05/02/2018
From:	Bradley Jones; Duke Energy Progress
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	HNP-18-023
	Download: ML18122A071 (30)
	v • d • e

( ~ DUKE Bentley K. Jones ENERGYa Director, Organizational Effectiveness Harris Nuclear Plant 5413 Shearon Harris Road New Hill, NC 27562-9300 919.362 2305 MAY O 2 20'18 10 CFR 50. 59(d)(2)

HNP-18-023 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:

Report of Changes Pursuant to 10 CFR 50.59 and Summary of Commit ment Changes Ladies and Gentlemen:

In accordance with 10 CFR 50.59(d)(2), Duke Energy Progress, LLC, submits the attached report for Shearon Harris Nuclear Power Plant, Unit 1 (HNP). The enclosure provides a brief description of changes to the facility and a summary of the evaluations required per 10 CFR 50.59 for those items, regardless of implementation status, between April 12, 2016, and April 5, 2018.

This letter also informs the NRC that there have been no unreported changes in commitments made during the period from April 12, 2016, through April 5, 2018.

This letter contains no new regulatory commitments.

If you have any questions regarding this submittal, please contact Jeff Roberts on, Manager -

Regulatory Affairs, at (919) 362-3137.

Sincerely,

&u l"1 ~

Bentley K. Jones

Enclosure:

Report of Changes Pursuant to 10 CFR 50.59 cc: J. Zeiler, NRC Sr. Resident Inspector, HNP M. Barillas, NRC Project Manager, HNP NRC Regional Administrator, Region II

Bentley K. Jones Director, Organizational Effectiveness Harris Nuclear Plant 5413 Shearon Harris Road New Hill, NC 27562-9300 919.362.2305 10 CFR 50.59(d)(2)

HNP-18-023 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:

Report of Changes Pursuant to 10 CFR 50.59 and Summary of Commitment Changes Ladies and Gentlemen:

In accordance with 10 CFR 50.59(d)(2), Duke Energy Progress, LLC, submits the attached report for Shearon Harris Nuclear Power Plant, Unit 1 (HNP). The enclosure provides a brief description of changes to the facility and a summary of the evaluations required per 10 CFR 50.59 for those items, regardless of implementation status, between April 12, 2016, and April 5, 2018.

This letter also informs the NRC that there have been no unreported changes in commitments made during the period from April 12, 2016, through April 5, 2018.

This letter contains no new regulatory commitments.

If you have any questions regarding this submittal, please contact Jeff Robertson, Manager -

Regulatory Affairs, at (919) 362-3137.

Sincerely, Bentley K. Jones

Enclosure:

Report of Changes Pursuant to 10 CFR 50.59 cc: J. Zeiler, NRC Sr. Resident Inspector, HNP M. Barillas, NRC Project Manager, HNP NRC Regional Administrator, Region II

U.S. Nuclear Regulatory Commission HNP-18-023 Enclosure HNP-18-023 ENCLOSURE SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 DOCKET NO. 50-400/RENEWED LICENSE NO. NPF-63 REPORT OF CHANGES PURSUANT TO 10 CFR 50.59 (27 pages plus cover)

Enclosure to HNP-18-023 Page 1 of 27 Report of Changes Pursuant to 10 CFR 50.59 Log Number / Description of Change Evaluation Summary Implementing Document 01962459/ While engineering personnel were evaluating EC 284102 results in an increase to the maximum Engineering operational experience associated with the required RWST switchover volume and decreases in CT Change Susquehanna Plant, it was discovered that the Harris pump and RHR pump NPSH margins. These (EC) 284102, Nuclear Power Plant, Unit 1 (HNP), Technical components are used for accident mitigation and are not Revision 0 Specifications (TS) contained non-conservative potential accident initiators. Therefore, this activity has no surveillance test acceptance criteria for determining impact on the frequency of occurrence of any accident Emergency Diesel Generator (EDG) operability. The previously evaluated in the FSAR.

associated surveillance test procedures utilize EDG transient ranges for voltage and frequency, i.e. 6900 Built-in margin exists to compensate for the increase in volts +/- 10% and 60 Hz +/- 2%. These values are the maximum required RWST switchover volume. Part of applicable when the EDG is operating in the the volume between the RWST Lo-Lo and Empty set isochronous mode (i.e. isolated from the offsite points is margin that is uncredited by analysis. The source) and only when the generator is coming up to currently available switchover margin is approximately speed or is being loaded (i.e. transient). Steady state 20,600 gallons. This will decrease to approximately frequency and voltage conditions were not identified. 19,300 gallons. Since some of the RWST margin can be Therefore, the existing TS ranges for frequency and credited to compensate for the increase in analytical voltage are too wide for steady-state conditions. This outflow, none of the existing RWST set points are condition was entered into the Corrective Action affected; all associated automatic and procedural actions Program as Nuclear Condition Report (NCR) 461896. remain unchanged. Therefore, the increase in maximum The HNP EDG voltage regulators are set at 6900 required switchover volume and corresponding decrease volts alternating current (VAC) +/-120 volts. ECs in switchover margin have no impact on the likelihood of 69609 and 82877 replaced the originally-supplied occurrence of a malfunction of a structure, system, or EDG Woodward analog speed control system with a component (SSC) important to safety previously new Woodward 2301A electronic speed control evaluated in the FSAR.

governor. The steady state speed band of the governor is +/- 0.25%, which results in a steady state EC 284102 results in a reduction to the injection and frequency range between 60.15 hertz (Hz) and 59.85 recirculation-mode NPSH margin for the CT pump.

Hz. Insufficient NPSH margin can result in pump cavitation and performance degradation. Although there is a EC 284102 provides the basis for changes to new reduction in the available NPSH for the CT pumps, the voltage (+/-4%) and frequency (+/-0.8%) tolerances, available NPSH is still greater than the required NPSH in which are more restrictive than current limits, and both modes of operation. Therefore, this has a minimal

Enclosure to HNP-18-023 Page 2 of 27 Report of Changes Pursuant to 10 CFR 50.59 Log Number / Description of Change Evaluation Summary Implementing Document updates all affected documents accordingly. The TS impact on the likelihood of CT pump malfunction.

revision process will take place independently from EC 284102. EC 284102 results in a reduction to the RHR pump NPSHA in recirculation mode to be consistent with the existing design basis. Insufficient NPSH margin can The steady-state frequency tolerance change was result in pump cavitation and performance degradation.

incorporated into plant design-basis documents Although there is a reduction in the available NPSH for (calculations) in order to account for the effect on the RHR pumps as shown in the FSAR, the available EDG-driven safety-related components such as NPSH is still greater than the required NPSH. Therefore, pumps, fans, and motor-operated valves and updates this has a minimal impact on the likelihood of an RHR all affected plant documents accordingly. pump malfunction.

The reductions in RWST switchover margin and CT and When +0.8% frequency was incorporated into RHR NPSH margin do not have an impact on the ability calculations for maximum Containment Spray (CT), of any equipment to perform their accident and dose Residual Heat Removal (RHR), and mitigating functions. Although RWST switchover margin Charging/Safety Injection (CSIP) flow rates during was reduced, significant positive margin still remains.

switchover, it resulted in an increase to the Likewise, the NPSH margins for the CT and RHR pumps maximum required Refueling Water Storage Tank also remain positive, so there is no impact on pump (RWST) switchover volume from 63,360 gallons to performance or their ability to mitigate an accident.

64,688 gallons. Since the HNP Final Safety Therefore, this activity does not result in more than a Analysis Report (FSAR), Section 6.3.2 cites this minimal increase in the consequences of an accident volume, and since an increase in this volume is previously evaluated in the FSAR.

non-conservative, this was identified as an adverse effect to an FSAR-described design function. Also, The consequences of a failure of a CT pump or an RHR the increase in the CT pump flow rate resulted in a pump do not change as a result of this activity.

decrease in CT pump net positive suction head Therefore, this activity does not result in more than a (NPSH) margin during injection and recirculation minimal increase in the consequences of a malfunction of modes. In injection mode, the net positive suction an SSC important to safety previously evaluated in the head available (NPSHA) decreased from 92.3 feet FSAR.

to 92.0 feet and net positive suction head required (NPSHR) increased from 12.5 feet to 13.0 feet. In There are no new failure modes established by EC recirculation mode, NPSHA decreased from 27.1 284102 and no new equipment added to the plant.

feet to 25.5 feet and NPSHR increased from 12.0 Reductions in the RWST switchover margin and the CT

Enclosure to HNP-18-023 Page 3 of 27 Report of Changes Pursuant to 10 CFR 50.59 Log Number / Description of Change Evaluation Summary Implementing Document feet to 12.4 feet. Since FSAR Section 6.2.2.3.2.1 and RHR pump NPSH margins do not result in new cites these values, and since this reduction in accident types. Therefore, this activity does not create a NPSH margin is non-conservative, this was possibility for an accident of a different type than any identified as an adverse effect to an FSAR- previously evaluated in the FSAR.

described design function.

This activity does not affect the design function of the During EC 284102 development, it was noted that RWST, the CT pumps, or the RHR pumps. Therefore, the minimum NPSHA for the RHR pumps this activity does not create a possibility for a malfunction immediately following switchover to recirculation is of an SSC important to safety with a different result than cited as 22.14 feet in FSAR Table 6.3.2-1. This any previously evaluated in FSAR.

does not agree with the value of 20.85 feet shown in the existing plant calculation, SI-0043. So, the This activity does not result in a design basis limit for a value in the FSAR will be corrected by EC 284102 fission product barrier as described in the FSAR being to match SI-0043. This reduction in RHR pump exceeded or altered. The method of calculating these NPSHA as shown in the FSAR is considered an margins was unchanged - only the flow-rate inputs were adverse effect to an FSAR-described design revised to account for increased pump speeds function. associated with +0.8% frequency tolerance. Therefore, this activity does not result in a departure from a method of evaluation described in the FSAR used in establishing the design bases or in the safety analyses.

01981672/ In order to resolve the non-conforming condition This activity concerns the HNP's response to a SGTR EC 296193, described in Nuclear Condition Report (NCR) event, regardless of its frequency of occurrence. The Revision 0 626242, EC 296193 amends the HNP steam MTO analysis is based on existing plant design features generator tube rupture (SGTR) margin-to-overfill and existing emergency operating procedures. The (MTO) analysis of record and, subsequently, revises activity does not add, delete, or modify any plant the FSAR, Section 15.6.3, and affected plant components. Therefore, this activity has no impact on the procedures. As identified in NCR 626242, a credible frequency of occurrence of a SGTR event or any other failure in the turbine-driven auxiliary feedwater pump accident previously evaluated in the FSAR. The (TDAFWP) speed control system could cause the evaluation also concludes that the reduction in required TDAFWP to run at the upper end of its speed-control AFW isolation time for a SGTR event from 10 minutes to range of 4,100 revolutions per minute (RPM), rather 8.8 minutes does not result in more than a minimal

Enclosure to HNP-18-023 Page 4 of 27 Report of Changes Pursuant to 10 CFR 50.59 Log Number / Description of Change Evaluation Summary Implementing Document than at the normal steady-state speed of increase in the likelihood of occurrence of a malfunction approximately 3,500 RPM as is implicitly assumed by of an SSC important to safety as evaluated in the FSAR.

the current FSAR Section 15 SGTR MTO analysis.

Such a failure would result in additional feedwater This activity does not revise the SGTR dose analysis; it is delivery to a faulted steam generator and could limited to the MTO analysis and the single-failure adversely impact the calculated margin to overfill. assumptions within the MTO analysis. This activity has This single failure scenario involving the TDAFWP determined that the TDAFWP speed controller failure is speed controller is not new. It was originally the most limiting single failure with respect to MTO considered in calculation HNP-M/MECH-1049, following a SGTR and has shown that acceptable MTO is Revision 0, in 2001, prior to the performance of the maintained with this limiting equipment malfunction.

current MTO analysis. However, when the MTO Other plausible equipment malfunctions associated with analysis was supplemented in 2010, the supplement MTO result in greater MTO. Therefore, positive MTO is failed to incorporate or consider this credible failure. maintained regardless of the equipment malfunction and this basic assumption in the dose analysis remains valid.

The maximum allowable SGTR AFW isolation time is None of the results of this activity impact the SGTR dose being reduced. This action is credited and described analysis. Therefore, this activity does not result in more in the SGTR FSAR Chapter 15 analysis and is a than a minimal increase in the consequences of an design function. Reducing this time has an adverse accident previously evaluated in the FSAR. This activity impact on a design function and on the control of this also does not result in more than a minimal increase in design function. the consequences of a malfunction of an SSC important to safety previously evaluated in the FSAR.

This activity does not add, delete, or modify components within the plant. This activity is limited in scope to the re-evaluation of an existing accident, a SGTR event, given a different input value for AFW delivery. No new accident types are considered or can be introduced. Therefore, this activity does not create the possibility of an accident of a different type not previously evaluated in the FSAR.

This activity also does not create the possibility for a malfunction of an SSC important to safety with different results than any previously evaluated in the FSAR.

Enclosure to HNP-18-023 Page 5 of 27 Report of Changes Pursuant to 10 CFR 50.59 Log Number / Description of Change Evaluation Summary Implementing Document A SGTR event assumes the failure of a portion of the reactor coolant system (RCS) pressure boundary (a steam generator tube). This is an existing FSAR Chapter 15 analysis. The AFW delivery value modified in the most recent version of the SGTR MTO analysis in HNP-M/MECH-1049, Revision 1, does not affect safety injection inputs, operator action types, or other parameters that would adversely impact fuel cladding integrity. Therefore, this activity does not result in a design basis limit for an FSAR-described fission product barrier being exceeded or altered.

For this activity, the SGTR MTO supplemental analysis in HNP-M/MECH-1049, Revision 1, uses the existing analysis of record (AOR), which is contained in calculation CN-CRA-10-31, as its basis. CN-CRA-10-31 was based on, and supplements, the previous AOR identified in calculation CN-CRA-99-80, which was based on the methodology of WCAP-10698, as described in FSAR Section 15.6.3.

The evaluation performed in HNP-M/MECH-1049, Revision 1, is a disposition of a single failure not previously considered in calculation CN-CRA-10-31. The evaluation and calculation both follow the NRC approved methodology of WCAP-10698.

Although the evaluation performed in HNP-M/MECH-1049 is not a mechanistic code run as are the cases in calculation CN-CRA-10-31, the evaluation presents the expected results should the mechanistic run be performed. The evaluation determines whether various input changes that affect the ruptured SG mass yield a

Enclosure to HNP-18-023 Page 6 of 27 Report of Changes Pursuant to 10 CFR 50.59 Log Number / Description of Change Evaluation Summary Implementing Document different margin to overfill than the prior calculation in CN-CRA-10-31. The evaluation is based on the previous results of the mechanistic calculation which follows the methodology. Because the only changes being made are inputs and are not elements of the method, these changes are not considered a departure from the method of evaluation described in the FSAR.

02053968/ EC 296136 resolves a non-conforming condition The pressure and temperature changes for the EC apply EC 296136, identified by the Westinghouse Nuclear Safety to the analysis of a post-LOCA containment atmosphere, Revision 0 Advisory Letter (NSAL) 2, Westinghouse Loss- after an accident has already occurred. There are no of-Coolant Accident Mass and Energy Release additions, deletions, or modifications to any SSCs as a Calculation Issue for Steam Generator Tube Material result of this activity. Therefore, there is no increase in the Properties. In NSAL-14-2, Westinghouse identifies frequency of occurrence of an accident previously an error in their calculation of mass and energy evaluated in the FSAR.

(M&E) release histories for large-break loss-of-coolant accidents (LOCAs) applicable to the HNP, The peak containment pressure is increased from 41.8 among other nuclear power plants. Specifically, psig to 42.0 psig. 42.0 psig is less than the design NSAL-14-2 notes that LOCA M&E analyses are pressure of 45 psig specified in TS 5.2.2 and in HNP-sensitive to the energy stored in the RCS metal M/MECH-1008. The pressure margin, as shown in FSAR mass, which includes the mass of the steam Table 6.2.1-3, decreases from 7.1% to 6.7% at 42.0 psig.

generator (SG) tubes. The Westinghouse M&E 42.0 psig is less than the initial containment pressure analysis for the HNP has historically assumed the SG used during EST-210, the integrated leak rate test for tubes to be stainless steel. The HNP SG tubes are containment, which pressurizes containment to 44-45 Alloy 690. psig. Based upon this, the new maximum calculated post-LOCA containment pressure of 42.0 psig is As a result of this NSAL, peak post-LOCA acceptable and does not represent more than a minimum containment pressure and temperature at HNP have increase in the likelihood of occurrence of a malfunction increased by small amounts in order to compensate of the containment pressure boundary.

for the error discovered by Westinghouse in the M&E analysis. Specifically, the peak post-accident The new maximum calculated post-LOCA sump and containment pressure for the LOCA double-ended spray pH values remain within the 7.0 to 11.0 range

Enclosure to HNP-18-023 Page 7 of 27 Report of Changes Pursuant to 10 CFR 50.59 Log Number / Description of Change Evaluation Summary Implementing Document hot-leg break case increases from 41.8 pounds per specified in Design Basis Document (DBD) -106, TS square inch gauge (psig) to 42.0 psig. This bounds Bases, Section 3/4.1.2, and the FSAR, Sections 3.11.5.1, all cases (LOCA and Main Steam Line Break) and is 6.1.1.2, and 6.5.2.1.2. The upper pH limit is intended to the most- limiting containment pressure case. The preclude excessive corrosion of equipment inside maximum LOCA-based containment atmospheric containment. Increases of the magnitudes noted, temperature increases from 270.2 degrees combined with the remaining pH margin (pH remains less Fahrenheit (o F) to 270.4o F. However, this is not the than 11.0), indicates that there will be no practical or bounding containment temperature case, as the unacceptable change in the amount of corrosion bounding case is based upon the Main Steam Line expected inside containment. Based upon this, the new Break and remains unchanged. maximum calculated post-LOCA sump and spray pH values are acceptable and do not represent more than a EC 296136 incorporates pressure and temperature minimal increase in the likelihood of occurrence of a penalties into the HNP containment analysis and malfunction of equipment inside containment due to evaluates the impact on plant documents. The peak excessive post-LOCA corrosion.

post-LOCA containment pressure for HNP is identified in HNP TS, Section 6.8.4.k. EC 296136 HNP Dose Analysis is independent of peak containment provides the basis for the change to the peak post- pressure and relies instead on the leak rate limit from TS.

LOCA containment pressure value identified in TS. A The dose analysis in HNP-F/NFSA-0072 was not revised TS change is necessary to implement EC 296136. for this activity. Also, the containment leak rate assumed Containment integrated leak rate testing is controlled in the dose analysis remains bounding since completed through Engineering Surveillance Tests (ESTs), integrated leak rate tests and local leak rate tests have which will be revised to reflect the pressure change used test pressures higher than new analytical limit. The as a result of the TS Change. Specifically, EST-209, changes to the sump and spray pH profiles are a factor in EST-210, EST-212, EST-219, EST-220, EST-221, the calculation of chemical precipitate formation in the and EST-222. EPT-221 will be revised. These recirculation pool. The quantities of these precipitates procedures are used to ensure containment integrity affect the pressure drop across the strainers and, and to ensure that the structure continues to perform consequently, core cooling through the RHR Pumps.

its pressure-boundary design function. In each of However, EC 296136 shows that the quantities of these procedures, peak accident pressure (Pa) is precipitates used during strainer testing remain bounding used as an acceptance criteria for the measured end- compared to the revised calculated amounts. Therefore, of-test pressure or as the minimum pressure to be there is no impact to the analyzed strainer pressure drop maintained during testing. The existing Pa value of or core cooling. Based on the above, this activity does

Enclosure to HNP-18-023 Page 8 of 27 Report of Changes Pursuant to 10 CFR 50.59 Log Number / Description of Change Evaluation Summary Implementing Document 41.8 psig will be revised to 42.0 psig. EC 296136 has not result in an increase in the consequences of an shown that existing (most recent) test results will still accident previously evaluated in the FSAR.

meet the revised criteria.

The increases in calculated post-LOCA containment The pressure penalty used to compensate for the pressure and sump and spray pH do not have any impact M&E error described in NSAL-14-2 is 0.2 psi. When on any SSC failure effects. Therefore, this activity does added to the existing containment analysis results, the not result in more than a minimal increase in the peak containment pressure increases from 41.8 psig consequences of a malfunction of an SSC important to to 42.0 psig. While this remains less than the 45 psig safety as previously evaluated in the FSAR.

design pressure for the containment structure, the increase represents an adverse effect on a design This activity does not make any physical changes to the function described in the FSAR. plant. The analyses revised for this activity involve post-LOCA conditions where an accident has already been The post-accident pH analysis for the containment assumed to occur. Slight increases in post-LOCA sump is a function of containment pressure. containment pressure and sump and spray pH do not Specifically, containment pressure affects the result in any new accident types. Therefore, this activity calculated rates of injection of sodium hydroxide does not create the possibility of an accident different (NaOH) solution from the Containment Spray Additive from any previously evaluated in the FSAR, nor does it Tank and borated water from the RWST. When the create the possibility for a malfunction of an important pressure profile was adjusted for NSAL-14-2, and SSC with a result that is different from that previously when a latent non-conservatism in the pH analysis evaluated in the FSAR.

was corrected, the resulting maximum pH values in the sump and in the containment spray system went In addition, this activity does not result in a design basis up slightly. For the sump, maximum pH went from limit for a fission product barrier being exceeded. This 9.420 to 9.422. For the spray, maximum pH when up activity does not alter the existing containment design from 10.578 to 10.606. Although the final pH values pressure of 45 psig.

are within the design range of 7.0 to 11.0 from FSAR Section 6.5.2, the change represents an adverse This activity revises the containment analysis in HNP-effect on a design function described in the FSAR. M/MECH-1008 to note that the NSAL-14-2 pressure and temperature penalties are to be applied to the existing results. This revision is an amendment to the existing analysis to require the manual addition of pressure and

Enclosure to HNP-18-023 Page 9 of 27 Report of Changes Pursuant to 10 CFR 50.59 Log Number / Description of Change Evaluation Summary Implementing Document temperature penalties; thus there is no reanalysis performed.

As a result of the containment analysis changes, a revision is also required to the containment sump and spray pH analysis in calculation 14.06.000-021. The change to this calculation involves an input (containment pressure). The existing analytical method is not modified, only rerun. Based on the above, this activity does not result in a departure from a method of evaluation described in the FSAR that is used to establish a design basis or used in a safety analysis.

02055046/ The cycle-specific thermal-hydraulic analysis results The activity does not modify or remove any SSC other HNP-F/NFSA- for the HNP Cycle 21 reload core show that for the than fuel. The results of an accident analysis do not 0264, rod ejection accident documented in the FSAR under affect the frequency of its occurrence. Therefore, this Revision 1, Section 15.4.8, there is a change from no fuel activity does not affect the frequency any accident.

HNP Cycle 21 assemblies failing to all the rods in one fuel assembly Loading having failed cladding as a result of a departure from This change to the rod ejection accident remains less Pattern and nucleate boiling (DNB). This is an American Nuclear than the number of failed fuel assemblies evaluated in Core Models Society (ANS) Condition IV accident and the cycle- the dose analysis documented in the FSAR. This activity specific results are within the fuel failure assumptions does not involve change to the RCS pressure boundary, specified by the dose analysis. fuel, or any other SSC which would affect the likelihood of a rod ejection. Further, this activity does not modify, add, or remove any SSC (other than fuel) nor change how SSCs are used during normal operation or to mitigate an accident. Therefore, this activity does not result in more than a minimal increase in the likelihood of a malfunction of an SSC important to safety.

Since the dose analysis assumed cladding failures bound the estimated failures from the safety analysis, the

Enclosure to HNP-18-023 Page 10 of 27 Report of Changes Pursuant to 10 CFR 50.59 Log Number / Description of Change Evaluation Summary Implementing Document predicted dose consequences remain the same.

Therefore, the proposed activity does not result in a more than minimal increase in the consequences of an accident, nor does it result in a more than minimal increase in the consequences of a malfunction of an SSC.

A change to analysis results in FSAR, Section 15.4.8, does not constitute a new or different type of accident.

Thus, the proposed activity does not create the possibility of an accident of a different type. The change to analysis results in FSAR 15.4.8 does not constitute a new malfunction. Thus, the proposed activity does not create the possibility for a malfunction of an SSC with a different result.

The change from no fuel assemblies failing to all the rods in one fuel assembly having failed cladding as a result of DNB in the rod ejection accident involves an ANS Condition IV event where fuel failures are allowed.

Evaluations have been performed as necessary to ensure the fission product barrier (fuel cladding, RCS boundary, and containment) limits are not compromised except where allowed in the design basis accident dose analyses. In addition, this activity does not represent a departure in a method of evaluation.

02079940/ NRC Bulletin 2012-01, "Design Vulnerability In Installation of the OPP system was evaluated under 10 EC 402237, Electric Power System," dated July 27, 2012, CFR 50.59 in accordance with guidance provided in Revision 0 requires licensees to install open-phase protection on Nuclear Energy Institute (NEI) 96-07, Revision 1, station transformers supplying offsite power to Guidelines for 10 CFR 50.59 Implementation, and NEI essential plant safety equipment. An open-phase 01-01 (EPRI TR-102348, Revision 1), Guideline on

Enclosure to HNP-18-023 Page 11 of 27 Report of Changes Pursuant to 10 CFR 50.59 Log Number / Description of Change Evaluation Summary Implementing Document condition (OPC) present in the offsite power system Licensing Digital Upgrades.

may not be detectable on the low side of the startup transformer (SUT) when the transformer load is not The OPP system provides SUT protection similar to the large enough to cause the emergency bus voltage to existing SUT protective relays. Inadvertent operation of drop below the under-voltage protective relay setting. the existing SUT protective relays can lead to lockout of a The transformers at HNP that may be susceptible to SUT resulting in partial or complete loss of non-an OPC are limited to the startup transformers, SUT- emergency AC power and subsequently partial or A and SUT-B, as they are the primary transformers complete loss of forced reactor coolant flow. Inadvertent used to supply offsite power to the 6.9 kilovolt (kV) operation of the new OPP system can also result in essential buses. lockout of a SUT, but to no more extent than the existing SUT protective relaying.

A previous EC 296261 installed a digital-based Open Phase Protection (OPP) system on the high-voltage The OPP system has been subjected to analyses, tests, side of SUT-1A and SUT-1B. The OPP system and requirements typically applied to equipment used in installed under EC 296261 provides OPC monitoring safety related applications (demonstrating a high quality only and is not capable of locking-out (i.e., tripping) a threshold). Additionally, the OPP system employs two-SUT. The intent of the new OPP system is to out-of-four coincidence trip logic, which provides added enhance protection of the Class 1E (safety-related) reliability and further assures that a valid trip signal will power system from a potential degraded condition be processed while an invalid trip signal will be caused by an OPC that could adversely affect both disregarded. Existing SUT protective relaying employs Class 1E and non-Class 1E systems. one-out-of-one coincidence trip logic, which is less reliable. With the existing protection scheme, failure The OPP system consists of four separate cabinets (malfunction) of a single protective relay could result in per SUT with each cabinet housing one of four SUT lockout while a single failure within the OPP system separate OPC sensing/trip channels. Of the four will not result in a SUT lockout.

channels per SUT, two channels employ one type of controller platform and central processing unit (CPU) Trip setpoints for the new OPP system were established architecture while the other two channels employ a to maintain coordination with other protective relay different type of controller platform and CPU schemes and to accommodate normal plant operation architecture. A SUT lockout command is generated (such as equipment starts and stops) to ensure spurious when any two of the four channels detects an OPC. actuation of the OPP system does not occur. The Thus, satisfying the SUT trip logic when a valid OPC setpoint values will be monitored and validated during the

Enclosure to HNP-18-023 Page 12 of 27 Report of Changes Pursuant to 10 CFR 50.59 Log Number / Description of Change Evaluation Summary Implementing Document is detected does not rely on a single type of controller OPP system monitoring phase of operation.

platform or CPU architecture.

Based on the robust design, considerable testing, and This activity (EC 402237) will physically connect the analyses performed on the OPP system to be installed as OPP system installed under EC 296261 to the part of this activity, it can be reasonably concluded that corresponding SUT lockout relay. the quality and reliability of the OPP system is at least as good as the existing SUT protective relays. Therefore, implementation of the proposed activity will not result in more than a minimal increase in the frequency of occurrence of an accident previously evaluated in the FSAR, nor will it result in more than a minimal increase in the likelihood of occurrence of a malfunction of an SSC important to safety previously evaluated in the FSAR.

The new OPP system will not increase operator burden or place constraints on an operators ability to adequately respond to an accident. The initial accident assessments contained in the FSAR remain valid and unchanged as a result of the implementing activity. The new equipment installed by this activity will have no adverse impact on its installed environment or another plant SSC. Therefore, the proposed activity will not result in more than a minimal increase in the consequences of an accident previously evaluated in the FSAR, nor will this activity result in more than a minimal increase in the consequences of a malfunction of an SSC important to safety previously evaluated in the FSAR.

Failure of the new OPP system can result in loss of SUT, but to no more extent than failure of an existing SUT protective relay. Since only the SUTs are affected by the

Enclosure to HNP-18-023 Page 13 of 27 Report of Changes Pursuant to 10 CFR 50.59 Log Number / Description of Change Evaluation Summary Implementing Document proposed activity, and the types of accidents resulting from a loss of SUT have already been analyzed in the safety analysis, the proposed activity cannot create the possibility for an accident of a different type than previously evaluated in the FSAR. No new outcomes have been introduced and the proposed activity to provide SUT open-phase protection cannot create the possibility for a malfunction with a different result than previously evaluated in the FSAR. The proposed activity will not result in a design basis limit for a fission product barrier as described in the FSAR being exceeded or altered. The proposed activity does not involve a change to any element of the analytical methods described in the FSAR used to demonstrate the design meets the design basis or that the safety analysis is acceptable, nor does this change involve use of a method or evaluation not already approved by the NRC. Therefore, the proposed activity will not result in a departure from a method of evaluation described in the FSAR used in establishing the design bases or in the safety analyses.

02093286/ This evaluation is a revision to an evaluation There is no impact to the 10 CFR 50.59 evaluation EC 296136, completed under Log Number 02053968, which is conclusions presented under Log Number 02053968 as Revision 1 described in this Enclosure. The revision adds EPT- a result of the revised evaluation, which is described in 222 to the list of impacted procedures. The initial this Enclosure.

issuance of this procedure occurred during EC 296136, Revision 0, development and was not identified in the first revision of the EC. EC 296136, Revision 1, includes EPT-222 as an impacted procedure.

Enclosure to HNP-18-023 Page 14 of 27 Report of Changes Pursuant to 10 CFR 50.59 Log Number / Description of Change Evaluation Summary Implementing Document 02092509/ This evaluation is prepared for EC 284243, HNP The new digital TCS was evaluated under 10 CFR 50.59 EC 284243, Turbine Control System Upgrade (TCSU) Integration. in accordance with guidance provided in NEI 96-07, Revision 0 The existing Westinghouse Digital Electro-Hydraulic Revision 1, and NEI 01-01 (EPRI TR-102348, Revision Control (DEH) system for the turbine-generator is 1). The new digital TCS performs the same turbine speed being replaced with an Invensys Triconex Digital and load control functions and interfaces with the same Electro-Hydraulic Turbine Control System (TCS) components and systems as the existing analog DEH utilizing triple modular redundant (TMR) digital system. The new TCS design assures that a single controllers, redundant input sensors and output component failure within the system will not result in a actuators to control and protect the turbine. The loss of steam or load control, or prevent a valid trip controls and electro-hydraulic interface include stand- response. A Failure Modes and Effects Analysis (FMEA) alone, fault-tolerant, and online maintainable trip was conducted, which concludes that the TCS contains block assemblies that will hydraulically trip the turbine no single points of vulnerability and that there is no single on overspeed conditions sensed by either the failure, which on its own, could result in a turbine trip.

Turbine Controller or the diverse Secondary Therefore, this activity does not result in more than a Overspeed Protection System (SOPS) system, or will minimal increase in the frequency of occurrence of any act to slow down the turbine speed by closing the accident previously evaluated in the FSAR.

control valves during certain scenarios (load rejection).

The existing turbine mechanical and electrical hydraulic trip components are replaced in the new TCS design with This modification is being implemented to improve equipment of equal or greater reliability, and will be plant reliability. The existing DEH control system controlled through redundant components and single reflects a relatively old design provided by failure proof voting logic. The new system improves the Westinghouse. In addition to obsolescence issues reliability of the entire TCS system, will not result in a with the existing system, a number of single failure system-level failure, and either will not affect or will points exist since fault tolerance was not a significant reduce the likelihood of occurrence of a malfunction as consideration during its development. The new TCS postulated in the FSAR.

design provides a state-of-the-art, fault-tolerant control system.

The consequences of a failure of the new TCS are bounded by the consequences of a failure of the existing The new TCS Triconex network is composed of the TCS. Failure of the new TCS could result in an increase Turbine and Valve Control System (TVCS), Turbine or decrease in heat removal from the secondary system, Protection System (TPS), SOPS, Human-System but no more than the failure of the existing TCS. Thus,

Enclosure to HNP-18-023 Page 15 of 27 Report of Changes Pursuant to 10 CFR 50.59 Log Number / Description of Change Evaluation Summary Implementing Document Interface (HSI) cabinet, Distributed Control System replacement of the TCS will not result in more than a (DCS) equipment, Engineering Workstation (EWS), minimal increase in the consequences of an accident and Main Control Room (MCR) Operator HSIs. The previously evaluated in the FSAR. The new TCS also TVCS is a subsystem of the TCS that includes all does not result in more than a minimal increase in the critical control functions and turbine protection: speed consequences of a malfunction of an SSC important to and load control, turbine protection trips (other than safety previously evaluated in the FSAR.

diverse SOPS and any other trips performed by the TPS that are external to the TVCS) and valve This activity does not introduce any components with management. The TPS is a subsystem of the TCS new failure modes and effects that are not bounded by that includes the hydraulic trip functions and the the accidents evaluated in the FSAR. This activity does diverse and independent SOPS. The SOPS is a not alter the failure modes and effects of the existing digital trip system that replaces the mechanical components. The overspeed trip portion of the new overspeed trip system and is diverse and system remains independent from the control portion of independent from the other control and protective the system. System-level failure modes for the features of the TCS. equipment are immediate or result in initiation of a turbine trip. The FMEA for the new TCS also concludes Both overspeed trip systems rely on independent that at the system level the most severe effects of triple speed sensing inputs and voting logic, including failures are a turbine trip and loss of some functionality at sensor health monitoring and fault notification alarms the HSI. No new failure modes are introduced by and warnings. Both systems will trip the turbine on replacement of the existing equipment with the TCS, and loss of or diverging speed signals. The design of the this replacement makes no change to the most limiting new turbine control provides the plant operators with scenario of the turbine trip previously evaluated in the a better graphical interface on a common set of FSAR. Therefore, the new TCS does not create the monitors using a trackpad, keypad, and pointing possibility of an accident of a different type than devices rather than discrete switches and indication. previously evaluated in the FSAR.

The TCS upgrade includes the following functional The TCS upgrade does not introduce any new failure or differences that are conservatively treated as operating modes, functions, interfaces, or operating adverse: parameters that would create a possibility of a (1) The change from functionally diverse mechanical malfunction for any SSC important to safety with a and electrical overspeed turbine trip mechanisms to different result than any previously analyzed. Thus, the redundant and electrically diverse overspeed trip failure effects of the new digital TCS are consistent with

Enclosure to HNP-18-023 Page 16 of 27 Report of Changes Pursuant to 10 CFR 50.59 Log Number / Description of Change Evaluation Summary Implementing Document mechanisms. the failure effects of the existing DEH system and the (2) Conversion from hard controls to soft controls results of these malfunctions are the same as previously because it involves more than minimal differences in evaluated in the FSAR. Therefore, the new TCS does not the HSI. create the possibility of a malfunction for any SSC important to safety with a different result than any previously analyzed in the FSAR. The new TCS also does not result in a design basis limit for a fission product barrier as described in the FSAR being exceeded or altered.

The proposed activity does not necessitate revision nor replacement of any evaluation methodology used in establishing any design basis or in the safety analysis.

No new methods of evaluation are required to assess the new equipment installed as part of this activity. No alternative or new methods of evaluation are required or employed for this activity. Therefore, the TCS upgrade does not impact existing evaluation methodology used in establishing any design basis or in the safety analysis.

02100628/ This Evaluation addresses a revised post-LOCA From engineering evaluation, the RADTRAD-NAI code EC 298102, Emergency Core Cooling System (ECCS) RWST implements the same methods as TITAN5 and Revision 1 backleakage dose assessment based upon EC generates essentially the same results as TITAN5 for 298102, Revision 1. The implementation of EC the RWST dose component of the total LOCA dose.

298102, Revision 1, involves revising or replacing certain evaluation methodologies described in the From engineering evaluation, the IODEX-NAI code FSAR, which were used in the design basis post- implements the same NUREG/CR-5950 methods as the LOCA ECCS RWST backleakage dose analysis. This Duke IODEX code, generates the same results as methodology change required a 10 CFR 50.59 IODEX for the RWST iodine releases, and there are no evaluation; no other aspect of this activity required 10 restraints or restrictions imposed on IODEXs use for CFR 50.59 evaluation. The methodology changes iodine release from review of NRC safety evaluation are: considerations. Based on criteria established by NEI 96-

Enclosure to HNP-18-023 Page 17 of 27 Report of Changes Pursuant to 10 CFR 50.59 Log Number / Description of Change Evaluation Summary Implementing Document (1) The Westinghouse TITAN5 dose analysis 07, Revision 1, it was concluded that the use of both the computer code results for the RWST component of RADTRAD-NAI code and the IODEX-NAI codes as total LOCA dose will be replaced with RADTRAD-NAI alternate methodologies is acceptable to implement dose analysis computer code results. without prior NRC review and approval.

(2) The Westinghouse approximations for iodine gas conversion and iodine gas partitioning between the The EC 298102, Revision 1, update to the AOR dose RWST liquid inventory and the RWST air inventory assessment demonstrates that the ECCS backleakage will be replaced with IODEX-NAI computer code allowable value, considering both onsite and offsite dose calculations. This code is derived from Duke criteria, can be increased from its current evaluation developed IODEX, which has been used for NRC basis limit of 3.0 gallons per minute (gpm). The change approved applications at other Duke nuclear sites. in the ECCS backleakage allowable value is evaluated under Log Number 02128760, which is described in this Enclosure. The revised post-LOCA ECCS RWST backleakage dose assessment ensures that the allowable backleakage flows to the RWST remain below the dose consequence results in the FSAR, Table 15.6.5-16.

02118552/ This evaluation addresses the Security Information The SIEM was evaluated under 10 CFR 50.59 in EC 405128, and Event Manager (SIEM), which will collect logging accordance with guidance provided in NEI 96-07, Revision 0 data through the existing Plant Process Network Revision 1, and NEI 01-01 (EPRI TR-102348, Revision (PNET) infrastructure. It was determined that the 1). The SIEM performs a monitoring function through a SIEM has the potential to fail or malfunction in a network that performs no control functions and cannot manner that results in a multicast/broadcast data initiate any plant transients or FSAR-described transmission (data storm) which could adversely accidents. Therefore, the proposed activity will result in affect the reliability of PNET and interfacing SSCs, no increase in the frequency of occurrence of any such as the Emergency Response Facility accident previously evaluated in the FSAR.

Information System (ERFIS) and Leading Edge Flowmeter (LEFM). Therefore, the scope of this The SIEM is qualitatively determined to be at least as evaluation is limited to SSCs with FSAR described dependable as the SSCs to which it is connected. The design functions that depend on PNET because only failure modes of the SIEM, and the likelihood of those functions were identified to be adversely malfunction, are indistinguishable from those of the

Enclosure to HNP-18-023 Page 18 of 27 Report of Changes Pursuant to 10 CFR 50.59 Log Number / Description of Change Evaluation Summary Implementing Document effected by EC 405128. existing equipment. Since there is no clear trend toward increasing the likelihood of failure, the proposed change is considered to have a negligible effect on the likelihood of malfunction. As a result, there is no credible malfunction of the SIEM that can increase the dose consequences of any FSAR-described accident. Based on the above, the proposed activity does not result in more than a minimal increase in the consequences of an accident previously evaluated in the FSAR. In addition, there is no credible malfunction of the SIEM that can increase the dose consequences of the malfunction of any SSC. Based on the above, the proposed activity does not result in more than a minimal increase in the consequences of malfunction of an SSC important to safety.

The SIEM does not have its own computing network; rather, it uses the same network as the components that it is monitoring, to collect the log data. A FMEA was performed for the SIEM. It concludes that there are no new failure modes or failure modes with a different result. Therefore, the proposed activity does not create a possibility for an accident of a different type than previously evaluated in the FSAR, nor does it create a possibility for a malfunction of an SSC important to safety with a different result than any previously evaluated in the FSAR.

The proposed activity does not directly or indirectly involve the fuel, the RCS pressure boundary, the containment, or any of the design basis limits associated with these fission product barriers. Consequently, the

Enclosure to HNP-18-023 Page 19 of 27 Report of Changes Pursuant to 10 CFR 50.59 Log Number / Description of Change Evaluation Summary Implementing Document activity cannot result in a design basis limit for a fission product barrier as described in the FSAR being exceeded or altered. The proposed activity neither involves a change to any element of the analytical methods described in the FSAR used to demonstrate the design meets the design bases or that the safety analyses are acceptable, nor involves use of a method or evaluation not already approved by the NRC. Therefore, the proposed activity will not result in a departure from a method of evaluation described in the FSAR used in establishing the design bases or in the safety analyses.

02128760/ This evaluation addresses a revised post-LOCA The proposed activity addresses the outcome of EC 298102, recirculation sump water level decrease due to accidents previously evaluated in FSAR, Chapter 15.

Revision 1 increasing the allowable backleakage to the RWST The relevant accidents include LOCAs. After the water during post-LOCA recirculation. For above RWST level of the RWST reaches a minimum allowable value, water line backleakage, allowable seat leakage has coolant for long-term cooling of the core is obtained by been increased for the Containment Spray (CT) switching from the injection mode to the cold leg system and Charging/Safety Injection System (CS) recirculation mode of operation in which spilled borated boundary isolation valves. This increase in allowable water is drawn from the containment sump by the low seat leakage supports initiation of Category A seat head safety injection (RHR) pump and returned to the leakage testing of the subject valves. For below RCS cold legs. The CT System continues to operate to RWST water line backleakage, the allowable value further reduce containment pressure. It is during the has also been increased to support future leakage recirculation mode of operation where the activities assessments. No change is being made to the associated with EC 298102, Revision 1, are applicable.

allowed ECCS leakage within the Reactor Auxiliary The potential sump water level decrease due to Building (RAB). increasing the allowable backleakage to the RWST during post-LOCA recirculation occurs during accident The minimum sump water level calculation, SD-0022, mitigation (not initiation). Therefore, these changes are previously assumed an ECCS leakage rate back to limited to accident mitigation (not initiation) and do not the RWST of 520 cubic centimeters per hour (cc/hr) increase the frequency of occurrence of an accident or 0.0023 gpm. This has been increased to 17.30 previously evaluated in the FSAR. Component

Enclosure to HNP-18-023 Page 20 of 27 Report of Changes Pursuant to 10 CFR 50.59 Log Number / Description of Change Evaluation Summary Implementing Document gpm, and the previously assumed ECCS leakage to manipulations needed to support re- injection would not the RAB of 5643 cc/hr (0.025 gpm) has been increase the likelihood of a malfunction of any increased to 1.0 gpm, as documented in SD-0022, components.

Revision 2. SD-0022, Revision 1, concluded that after 30 days of ECCS leakage at 0.027 gpm, the As shown in the FSAR, Table 15.6.5-2, the safety recirculation sump water level remained above the analysis for a large break LOCA is terminated at 829.7 recirculation sump strainer ECCS strainer vortex seconds (0.23 hours2.662037e-4 days 0.00639 hours 3.80291e-5 weeks 8.7515e-6 months ). As shown in the FSAR, Figure suppressor. With the newly established allowable 15.6.5-32, the safety analysis for a small break LOCA is ECCS leakage of 18.30 gpm, sump water level will terminated at 6000 seconds (1.7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months ). The earliest drop below the vortex suppressor before 30 days potential need for re-injection from the RWST to the (31.9 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months is the minimum duration based upon recirculation sump is 31.9 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months . This is well past Calculation SD-0022, Revision 2). This requires re- termination of LOCA accident analyses and therefore injection from the RWST to the containment sump there is no impact to accident analyses. Since accident (using the water that had leaked back to the RWST) analyses remain valid, there can be no increase in using emergency operations procedure guidance, to accident dose consequences. The dose consequences ensure the vortex suppressor remains covered. resulting from increased RWST backleakage allowed under EC 298102, Revision 1, remain bounded by the EC 298102, Revision 1, may require periodic re- dose consequences identified in the FSAR, Table 15.6.5-injection from the RWST to the recirculation sump 16.

during recirculation to accommodate higher allowed ECCS leakage out of containment. As determined in The revision to EOP-ES-1.3, to utilize a CT pump to re-Calculation SD-0022, Revision 2, the minimum time inject from the RWST to the recirculation sump to following start of recirculation, until re-injection is maintain sump inventory in the event of significant needed ranges from approximately 31 to 41 hours4.74537e-4 days 0.0114 hours 6.779101e-5 weeks 1.56005e-5 months backleakage to the RWST, ensures that adequate sump depending on break size. If re-injection is not level is maintained to support ECCS and CT pump initiated, sump water level would eventually decrease operation to mitigate the accident. In the event of failure to the point where vortexing and/or incomplete of a CT train (pump, valve, etc.), the redundant CT train submergence of the ECCS strainers would result. would be available to support re-injection, should it be required due to significant RWST backleakage.

Emergency operations procedure guidance for the Containment sump level is provided with redundant transfer to cold leg recirculation, EOP-ES-1.3, safety-related instrumentation, so the redundant currently supports re-injection from the RWST to the instrument can be used by control room personnel to

Enclosure to HNP-18-023 Page 21 of 27 Report of Changes Pursuant to 10 CFR 50.59 Log Number / Description of Change Evaluation Summary Implementing Document recirculation sump using a CSIP and is being revised monitor sump level and assess the need for re-injection.

to support re-injection using a CT pump as the Equipment malfunctions will therefore not increase preferred method. However, this is an adverse accident doses since the ability to maintain long-term change (i.e., method used to perform post-LOCA core and containment cooling is assured by maintaining recirculation design function is adversely affected) adequate sump inventory.

from the current post-LOCA leakage assessment in SD-0022, Revision 1, where no re-injection is shown The changes being addressed in this evaluation are to be necessary for the 30-day LOCA. FSAR, Section associated with LOCA consequence mitigation. So the 6.3.2.8, will be revised to acknowledge that re- changes are applicable only after the accident has injection from the RWST may be required following occurred. There is no credible mechanism for another switchover to sump recirculation, to compensate for accident to occur following a LOCA. Therefore, the significant ECCS leakage outside containment in proposed activity does not create an accident of a order to maintain adequate sump inventory. different type than previously evaluated.

The CT system is designed for single failure. In the event of a failure that renders one train non-functional, the other train is capable of providing adequate injection flow from the RWST or recirculation flow from the sump. Re-alignment of a CT pump from recirculation back to injection (if needed to compensate for significant RWST backleakage), does not change this capability. The consequences of failure of a CT train are not changed by incorporating the ability to swap back to RWST injection if required. The remaining train remains adequate to perform its design basis function of containment cooling to ensure that containment integrity is maintained as assumed in the dose calculation. The remaining train is also available to be realigned for RWST re-injection if necessary. Therefore, the proposed change does not create a possibility for a malfunction of an SSC important to safety with a different result than any previously evaluated in the FSAR.

Enclosure to HNP-18-023 Page 22 of 27 Report of Changes Pursuant to 10 CFR 50.59 Log Number / Description of Change Evaluation Summary Implementing Document The changes being addressed in this evaluation are associated with LOCA consequence mitigation. So the changes are applicable only after the LOCA has occurred, at which point, two of the three fission product barriers (RCS and fuel cladding) are assumed to have been breached. The third barrier, containment, remains intact, experiencing only design basis leakage. The changes associated with this evaluation will not affect the ability of ECCS and CT to remove post-accident decay heat from containment. Therefore, containment pressure will remain within design limits. Thus, these changes do not result in a design basis limit for a fission product barrier as described in the FSAR being exceeded or altered. In addition, the proposed activity does not result in departure form a method of evaluation described in the FSAR.

02153055/ EC 409889 implements changes to the PNET The activity has been evaluated under 10 CFR 50.59 in EC 409889, configuration such that some of the PNET devices accordance with guidance provided in NEI 96-07, Revision 0 will be reconfigured to separate the primary ERFIS Revision 1, and NEI 01-01 (EPRI TR-102348, Revision related devices, including the Multiplexor Fiber Ring, 1). ERFIS provides monitoring, alarming, displaying, from the remaining portion of the PNET using a reporting and archiving capabilities to the Control Room Firewall and Intrusion Detection System. Additionally, operators, the Technical Support Center and the one of the ERFIS workstations has additional Emergency Operations Facility through a network and functionality as a "QNX" workstation used to manage performs no control functions. ERFIS is not an initiator of the Waste Processing Building computer. any FSAR-described accidents. Therefore, the proposed activity will not result in an increase in the frequency of occurrence of any accident previously evaluated in the FSAR. The failure modes of ERFIS, and the likelihood of malfunction, are indistinguishable from those of the existing equipment. Since there is no clear trend toward increasing the likelihood of malfunction, the proposed

Enclosure to HNP-18-023 Page 23 of 27 Report of Changes Pursuant to 10 CFR 50.59 Log Number / Description of Change Evaluation Summary Implementing Document change is considered to have a negligible effect on the likelihood of occurrence of a malfunction of an SSC important to safety previously evaluated in the FSAR.

ERFIS is not credited for mitigating the consequences of an accident. The changes to ERFIS per EC 409889 are not visible to the end user. Post implementation of EC 409888, ERFIS will still not be credited for mitigating the consequences of an accident. Therefore, the proposed activity has no impact on the consequences of an accident previously evaluated in the FSAR. ERFIS is not credited for mitigating the consequences of an accident.

Based on the above, the proposed activity does not result in more than a minimal increase in the consequences of a malfunction of an SSC important to safety previously evaluated in the FSAR. A FMEA was performed for the ERFIS related network changes under EC 409889. It concludes that there are no new failure modes and existing failure modes are not accident initiators. Consequently, the proposed activity does not create a possibility for an accident of a different type than previously evaluated in the FSAR and there is no possibility for a malfunction of an SSC important to safety with a different result than any previously evaluated in the FSAR. The proposed activity does not directly or indirectly involve the fuel, the RCS pressure boundary, the containment, or any of the design basis limits associated with these fission product barriers.

Consequently, the activity cannot result in a design basis limit for a fission product barrier as described in the FSAR being exceeded or altered. The proposed activity neither involves a change to any element of the analytical methods described in the FSAR used to demonstrate the

Enclosure to HNP-18-023 Page 24 of 27 Report of Changes Pursuant to 10 CFR 50.59 Log Number / Description of Change Evaluation Summary Implementing Document design meets the design bases or that the safety analyses are acceptable, nor involves use of a method or evaluation not already approved by the NRC. Therefore, the proposed activity will not result in a departure from a method of evaluation described in the FSAR used in establishing the design bases or in the safety analyses.

02183576/ The proposed activity is the revision to the FSAR The changes that require a revision to the FSAR analysis Revision to Section 15.8, Anticipated Transients without Scram are a result of the response to NSAL-11-01 to address FSAR, (ATWS), as a result of a revision to the respective potentially inadequate prediction of the SG low-low water Section 15.8 underlying calculation. Revision 0 of the level ATWS AMSAC setpoint utilizing SG mass. The Westinghouse ATWS analysis predicted the SG low- changes are best characterized as changes in input to low water level ATWS Mitigation System Actuation the ATWS analysis. It is not possible to characterize the Circuitry (AMSAC) setpoint utilizing SG mass. This proposed activity as an accident initiator. There are also may not have adequately addressed the transient no SSCs involved in this analysis related activity. Thus, behavior for HNP, as identified in NSAL-11-01, no SSCs could initiate an accident. Without changing the Calculation of the Steam Generator Mass for the frequency of occurrence of any accident initiators, no Low-Low Water Level Setpoint in LOFTRAN change in the classification of the accidents can occur.

Analyses. Based upon NSAL-11-01, the trip mass at Since there is no impact on the frequency of occurrence the low-low SG level setpoint needs to be of an accident, it can be concluded that the proposed reanalyzed. Thus, the ATWS analysis has been activity does not result in more than a minimal increase in recalculated by Westinghouse to accommodate the the frequency of occurrence of an accident previously issue discussed above for HNP. evaluated in the FSAR.

The revision to the FSAR ATWS analysis for peak RCS From NEI 96-07, Revision 1, this activity required a pressure is not related to the performance of SSCs. The 10 CFR 50.59 evaluation because the change to the proposed activity does not affect any aspect of SSC peak RCS pressure in the reanalysis for the Loss of design, including changing material or construction Normal Feedwater ATWS adversely affects the standards of SSCs. The proposed activity is solely design function of the RCS, as described in the focused on updating the FSAR analysis to reflect FSAR. One of the design functions of the RCS is to changes and the associated impacts on the ATWS provide a second barrier against fission product accident peak RCS pressure analysis result. Results are release in the event of fuel cladding failure. Another

Enclosure to HNP-18-023 Page 25 of 27 Report of Changes Pursuant to 10 CFR 50.59 Log Number / Description of Change Evaluation Summary Implementing Document design function of the RCS is to help ensure that still within the design limit, which does not change.

coolant (water) is available to remove heat from the Therefore, it can be concluded the proposed activity does fuel. The peak RCS pressure must not exceed the not result in more than a minimal increase in the American Society of Mechanical Engineers (ASME) likelihood of occurrence of a malfunction of an SSC Boiler and Pressure Vessel (B&PV) Code, Service important to safety previously evaluated in the FSAR.

Level C, stress limit criterion of 3,215 pounds per square inch absolute or 3,200 psig. In order for the The net effect of this change was an increase in the peak RCS to perform its design function, RCS pressure RCS pressure. DNB is not a concern for the ATWS needs to stay within the acceptance criteria. analysis. Thus, there is no fuel failure associated with the ATWS analysis. Therefore, there will be no impact on dose analyses. Also, since the RCS pressure increase does not exceed the ASME B&PV Code, Service Level C stress criterion of 3,200 psig, the RCS would not have been breached and any potential fuel failures would not have been released from the RCS. Thus, the proposed activity will have no impact on any dose analyses of record. Factors which influence dose calculations and environmental consequences remain consistent with the FSAR analyses. Therefore, the proposed activity does not result in more than a minimal increase in the consequences of an accident evaluated in the FSAR.

There is no fuel failure associated with the ATWS analysis; however, even if fuel failure was present, the RCS pressure increase will not cause a breach to the RCS. Thus, any potential fuel failure will be contained in the RCS. Therefore, the proposed activity will have no impact on any dose analyses of record and the consequences of a malfunction of an SSC important to safety previously evaluated in the FSAR will not be increased. The proposed activity does not create a possibility for an accident of a different type than

Enclosure to HNP-18-023 Page 26 of 27 Report of Changes Pursuant to 10 CFR 50.59 Log Number / Description of Change Evaluation Summary Implementing Document previously evaluated in the FSAR, does not create a possibility for a malfunction of an SSC important to safety with a different result than any previously evaluated in the FSAR, does not result in a design basis limit for a fission product barrier as described in the FSAR being exceeded or altered, and does not result in a departure from a method of evaluation described in the FSAR.

02189627/ The HNP Cycle 22 reload core analysis results show The activity does not modify or remove any SSC other HNP-F/NFSA- that for the single control rod withdrawal accident, than fuel. The results of an accident analysis do not 0284, documented in the FSAR under Section 15.4.3, there affect the frequency of its occurrence. Therefore, this Revision 0, is a change such that the predicted minimum activity does not affect the frequency any accident.

HNP Cycle 22 departure from nucleate boiling ratio (MDNBR) is less Loading than the 95/95 safety limit. For HNP Cycle 22, less This change to the single control rod withdrawal accident Pattern and than 4% of the fuel is predicted to fail based on the remains less than the number of failed fuel assemblies Core Models DNB criteria whereas previously for HNP Cycle 21, evaluated in the dose analysis documented in the FSAR.

no fuel was predicted to fail because the MDNBR This activity does not involve change to the RCS was greater than the 95/95 safety limit. The single pressure boundary, fuel, or any other SSC which would control rod withdrawal accident is an ANS Condition affect the likelihood of a single control rod withdrawal.

III accident and the cycle-specific results are within Further, this activity does not modify, add, or remove any the fuel failure assumptions specified by the dose SSC (other than fuel) nor change how SSCs are used analysis. during normal operation or to mitigate an accident.

Therefore, this activity does not result in more than a minimal increase in the likelihood of a malfunction of an SSC important to safety.

Since the dose analysis assumed cladding failures bound the estimated failures from the safety analysis, the predicted dose consequences remain the same.

Therefore, the proposed activity does not result in a more than minimal increase in the consequences of an accident, nor does it result in a more than minimal

Enclosure to HNP-18-023 Page 27 of 27 Report of Changes Pursuant to 10 CFR 50.59 Log Number / Description of Change Evaluation Summary Implementing Document increase in the consequences of a malfunction of an SSC.

A change to analysis results in FSAR, Section 15.4.3, does not constitute a new or different type of accident.

Thus, the proposed activity does not create the possibility of an accident of a different type. The change to analysis results in FSAR 15.4.3 does not constitute a new malfunction. Thus, the proposed activity does not create the possibility for a malfunction of an SSC with a different result.

The change from no fuel assemblies failing to one assembly exceeding the DNB cladding failure criteria for the single rod withdrawal accident involves an ANS Condition III event where a small fraction of fuel rod failures are acceptable. Evaluations have been performed as necessary to ensure the fission product barrier (fuel cladding, RCS boundary, and containment) limits are not compromised except where assumed in the design basis accident dose analyses. In addition, this activity does not represent a departure in a method of evaluation.