HNP-17-077, License Amendment Request Regarding Rod Control Movable Assemblies Technical Specifications

License Amendment Request Regarding Rod Control Movable Assemblies Technical Specifications

ML17283A159
Person / Time
Site:	Harris
Issue date:	10/10/2017
From:	Hamilton T Duke Energy Progress
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
HNP-17-077
Download: ML17283A159 (24)

Text

Tanya M. Hamilton Vice President Harris Nuclear Plant 5413 Shearon Harris Road New Hill, NC 27562-9300 919-362-2502 10 CFR 50.90 October 10, 2017 Serial: HNP-17-077 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 Regarding Rod Control Movable Assemblies Technical Specifications Ladies and Gentlemen:

Pursuant to 10 CFR 50.90, Duke Energy Progress, LLC (Duke Energy) hereby requests a revision to the Technical Specifications (TS) for the Shearon Harris Nuclear Power Plant, Unit 1 (HNP). The proposed license amendment modifies the TS for moveable control assemblies and for maintaining shutdown margin to align the designation of an OPERABLE rod control assembly with that of the industry. Specifically, Duke Energy is proposing changes to TS 3/4.1.1, Reactivity Control Systems Boration Control, and TS 3/4.1.3, Reactivity Control Systems Movable Control Assemblies Group Height, to align the TS more closely to improved Standard Technical Specifications for rod control and to initial conditions in HNP safety analyses.

Duke Energy proposes deleting TS action statement 3.1.3.1.c and modifying TS action statement 3.1.3.1.d. The action statements requirements include a plant shut down to address rods that are immovable but still trippable, such as upon receipt of a Rod Control Urgent Failure alarm. These actions are unnecessary, as rods that are immovable but still trippable still meet the associated Limiting Condition for Operation (LCO). Surveillance requirement (SR) 4.1.1.1.1.a and SR 4.1.1.2.a also contain actions to take for immovable rods that may still be trippable. Revision to these SRs is proposed to clarify actions are not necessary if rods are immovable but still trippable. These changes are consistent with Revision 4 of NUREG-1431, Standard Technical Specifications - Westinghouse Plants (Agencywide Documents Access and Management System (ADAMS) Accession No. ML12100A222).

provides a description and assessment of the proposed changes. Enclosure 2 provides the existing TS pages marked to show the proposed changes. Enclosure 3 provides existing TS Bases pages marked to show the proposed changes for information only.

Duke Energy requests approval of this License Amendment Request by October 10, 2018, with a 30-day implementation period. The proposed changes have been evaluated in accordance

U.S. Nuclear Regulatory Commission Serial HNP-17-077 Page 2 with 10 CFR 50.91(a)(1) using the criteria in 10 CFR 50.92(c), and it has been concluded that the proposed changes involve no significant hazards consideration.

In accordance with 10 CFR 50.91, a copy of this application, with attachments, is being provided to the designated North Carolina State Official.

This document contains no new Regulatory Commitments.

Should you have any questions regarding this submittal, please contact Jeff Robertson, HNP Regulatory Affairs Manager, at (919)362-3137.

I declare under penalty of perjury that the foregoing is true and correct. Executed on October 10, 2017.

Sincerely, Tanya M. Hamilton

Enclosures:

1. Evaluation of the Proposed Change.

2. Proposed Technical Specification Changes.

3. Technical Specification Bases Change (For Information Only).

cc:

J. Zeiler, NRC Sr. Resident Inspector, HNP W. L. Cox, III, Section Chief, N.C. DHSR M. Barillas, NRC Project Manager, HNP C. Haney, NRC Regional Administrator, Region II

U.S. Nuclear Regulatory Commission Serial HNP-17-077, Enclosure 1 SERIAL HNP-17-077 ENCLOSURE 1 EVALUATION OF THE PROPOSED CHANGE SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 DOCKET NO. 50-400 RENEWED LICENSE NUMBER NPF-63 15 PAGES PLUS THE COVER

U.S. Nuclear Regulatory Commission Page 1 of 15 Serial HNP-17-077, Enclosure 1 Evaluation of the Proposed Change License Amendment Request Regarding Rod Control Movable Assemblies 1.0 Summary Description Pursuant to 10 CFR 50.90, Duke Energy Progress, LLC (Duke Energy) is proposing changes to the Shearon Harris Nuclear Power Plant, Unit 1, (HNP) Technical Specification (TS) 3/4.1.1, Reactivity Control Systems Boration Control, and to TS 3/4.1.3, Reactivity Control Systems Movable Control Assemblies Group Height.

The proposed changes will delete TS action statement 3.1.3.1.c and modify TS action statement 3.1.3.1.d. The action statements requirements include a plant shut down to address rods that are immovable but still trippable, such as upon receipt of a Rod Control Urgent Failure alarm.

These actions are unnecessary, as rods that are immovable but still trippable still meet the associated Limiting Condition for Operation (LCO). Surveillance requirements (SR) 4.1.1.1.1.a and SR 4.1.1.2.a also contain actions to take for immovable rods that may still be trippable.

Revision to these SRs is proposed to clarify that actions are not necessary if rods are immovable but still trippable. These changes are consistent with NUREG-1431, Revision 4, Standard Technical Specifications - Westinghouse Plants, Volumes 1 and 2. (Reference 6.1 and 6.2, respectively.)

Reference 6.2 specifies that an immovable rod is inoperable only if it is also untrippable. A rod that is immovable but can be tripped, such as one impacted by a Rod Control Urgent Failure alarm or by an obvious electrical problem, would be considered operable. An immovable but trippable rod also has no adverse impact on shutdown margin (SDM).

HNP TS differ from Reference 6.2, in that it contains language defining an immovable but trippable rod as inoperable. Action statement 3.1.3.1.c establishes a shut down action for Rod Control Urgent Failure alarm or obvious electrical problems, two conditions that render a rod immovable but trippable. Action statement 3.1.3.1.d includes actions for a rod that is trippable but inoperable. SR 4.1.1.1.1.a and SR 4.1.1.2.a. include extra actions for a rod that is immovable or untrippable. Immovable rods that are still trippable should be considered operable, per Reference 6.2. These TS actions are unnecessary, as Reference 6.2 explains per Section B 3.1.4, and can lead to unnecessary plant shut downs and/or transients.

2.0 Detailed Description 2.1 System Design and Operation Rod Control System Design The Control Rod System provides for reactor power modulation by manual or automatic control of control rod banks in a preselected sequence and for manual operation of individual rod banks from the control room. The primary function of the Control Rod System is to provide a method of controlling the reactivity of the reactor core and to shut down the reactor.

Associated with this system is the Control Rod Drive Mechanism (CRDM), a three-coil electromagnetic jack that steps the control rod clusters into or out of the reactor core by raising

U.S. Nuclear Regulatory Commission Page 2 of 15 Serial HNP-17-077, Enclosure 1 or lowering a 144-inch drive rod to which the cluster assembly is attached. The power output of the reactor is controlled by varying rod cluster positions in the reactor core.

The CRDMs withdraw and insert rod cluster control assemblies (RCCAs) at a designated speed in a controlled manner during all normal phases of reactor operation, including start-up and shut down. The RCCAs are held at any step position within the range of the drive rod assembly travel during normal operation by providing electrical power to the stationary gripper coil of the CRDM. The CRDM provides rapid insertion (by force of gravity) of the drive rod assembly and attached rod cluster when electrical power to the CRDM operation coils is interrupted, either deliberately in a reactor trip or due to accidental power failure.

The following indication of the Control Rod System is provided in the control room:

a. Alarms to alert the operator if the required core reactivity SDM is not available due to excessive control rod insertion.

b. Display of control rod position.

c. Alarms to alert the operator in the event of control rod deviation exceeding a preset limit.

d. Plant Computer and historical and trend analysis.

There are two separate systems used to determine control rod position, the demand position system and the digital rod position indication (DRPI) system. Operating procedures require the reactor operator to compare the demand and indicated (actual) readings from the DRPI system to verify operation of the Rod Control System. The demand position system counts pulses generated in the Rod Drive Control System to provide a digital readout of the demanded bank position. The DRPI system measures the actual position of each control rod using a detector, which consists of discrete coils mounted concentrically with the rod drive pressure housing.

The Rod Control Urgent Failure alarm alerts operators to rod control system trouble conditions caused by an electrical issue. This alarm annunciates in the control room and inhibits automatic rod motion in the group in which it occurs, ensuring the demand position system and the actual rod position remain aligned. This condition is different from a rod becoming immovable from excessive friction or mechanical interference or known to be untrippable. The rods are free to move mechanically but are immovable electrically. A loss of power to the CRDM will still cause a rod insert by gravity, as would occur during a reactor trip.

The reactivity control systems must be redundant and capable of holding the reactor core subcritical when in shutdown under cold conditions. Maintenance of the SDM ensures that postulated reactivity events will not damage the fuel.

SDM requirements provide sufficient reactivity margin to ensure that acceptable fuel design limits will not be exceeded for normal shut down and anticipated operational occurrences. As such, the SDM defines the degree of subcriticality that would be obtained immediately following the insertion or scram of all shutdown and control rods, assuming that the single rod cluster assembly of highest reactivity worth is fully withdrawn. In addition, the Control Rod System, together with the boration system, provides the SDM during power operation and is capable of making the core subcritical rapidly enough to prevent exceeding acceptable fuel damage limits, assuming that the rod of highest reactivity worth remains fully withdrawn.

U.S. Nuclear Regulatory Commission Page 3 of 15 Serial HNP-17-077, Enclosure 1 Thus, the Rod Control System affects the sites ability to maintain SDM within accident analysis assumptions. SDM is maintained if shutdown and control rods can be inserted in the core during a reactor trip. Rods that are immovable and untrippable (i.e. stuck) would not insert by gravity on loss of power to the CRDM, such as during a reactor trip. However, rods that are immovable but trippable, such as when Rod Control Urgent Failure alarm is received or during other electrical problems, would still insert during a reactor trip. Thus, untrippable rods impact SDM whereas rods that are immovable but trippable do not.

2.2 Current Technical Specifications Requirements There are TS that govern operation of the rod positioning systems and rod position indication systems. The proposed changes will impact the TS which governs rod positioning only.

Specifically, the proposed changes will be to TS 3/4.1.3, Reactivity Control Systems Movable Control Assemblies Group Height. The limiting condition for operation (LCO) per TS 3.1.3.1 is All shutdown and control rods shall be OPERABLE and positioned within +/- 12 steps (indicated position) of their group step counter demand position. Thus, this LCO addresses both control rod safety functions by requiring all rods to be operable and requiring them to be positioned within +/- 12 steps (indicated position) of their demand position.

Action statement 3.1.3.1.a lists actions to take if a rod is immovable as a result of excessive friction of mechanical interference or known to be untrippable. This action statement ensures appropriate actions are taken if a control rod can no longer freely fall into the reactor when power to the CRDM is interrupted, as is the case during a reactor trip.

Action statements 3.1.3.1.b and 3.1.3.1.d ensure actions are taken in the event one or more control rods are misaligned, which affects power distribution limits and potentially impacts accident analyses. Thus, these action statements also maintain a safety function of the rods.

Action statement 3.1.3.1.c requires actions if control rods are immovable due to electrical issues, including receipt of the Rod Control Urgent Failure alarm. Note that this condition does not prevent a loss of power to the CRDM from causing a rod to drop. Any misalignment in the rods that exceed the LCO for rod alignment will trigger either action statement 3.1.3.1.b or 3.1.3.1.d depending on the number of rods impacted. Immovable rods that remain trippable and aligned do not impact any safety analysis at HNP. Rod position indication would continue to display accurate rod position.

In addition to actions taken for misalignment, action statement 3.1.3.1.d also contains actions to take if control rods are trippable but inoperable.

There are also TS that govern boration control system operation to ensure adequate SDM is maintained. SRs within TS 3/4.1.1, Reactivity Control Systems Boration Control, require extra actions for verifying SDM in the event of inoperable rods. LCO 3.1.1.1 addresses SDM in modes 1 and 2. LCO 3.1.1.2 addresses SDM in modes 3, 4, and 5.

SR 4.1.1.1.1.a and SR 4.1.1.2.a specify actions to verify shutdown margin If the inoperable control rod is immovable or untrippable An immovable rod that is still trippable is still considered operable per Reference 6.2, and has no impact on SDM. The condition inoperable

U.S. Nuclear Regulatory Commission Page 4 of 15 Serial HNP-17-077, Enclosure 1 control rod will not be met and the SRs action is not necessary if the immovable rods remain trippable.

The following excerpts are from TS 3/4.1.1and 3/4.1.3 as currently written, with emphasis added:

LCO 3.1.3.1, Reactivity Control Systems Movable Control Assemblies Group Height:

U.S. Nuclear Regulatory Commission Page 5 of 15 Serial HNP-17-077, Enclosure 1 LCO 3.1.1.1, Reactivity Control Systems Boration Control Shutdown Margin - Modes 1 and 2:

LCO 3.1.1.2, Reactivity Control Systems Shutdown Margin Modes - 3, 4, and 5:

U.S. Nuclear Regulatory Commission Page 6 of 15 Serial HNP-17-077, Enclosure 1 2.3 Reason for Proposed Change Improved Standard Technical Specifications (STS), documented per Reference 6.2, specifies that an immovable rod is inoperable only if it is also untrippable. A rod that is immovable but can be tripped, such as one impacted by a Rod Control Urgent Failure alarm or by an obvious electrical problem, would not be inoperable and would not impact SDM.

HNP TS differ from Reference 6.2, in that it contains language defining an immovable but trippable rod as inoperable. Action statement 3.1.3.1.c establishes a shut down action for Rod Control Urgent Failure alarm or obvious electrical problems. Rods in this condition are still trippable, and should therefore be considered operable. This action statement is unnecessary and can lead to unnecessary plant shut downs and/or transients.

Also, action statement 3.1.3.1.d contains the statement, With one rod trippable but inoperableor misaligned. This statement is inconsistent with the definition of rod operability as established in Reference 6.2 Section B 3.1.4. A trippable rod should be considered operable, assuming the rod is aligned per its indicated position.

SR 4.1.1.1.1.a and SR 4.1.1.2.a both specify actions to verify shutdown margin If the inoperable control rod is immovable or untrippable. An immovable rod that is still trippable should be considered operable, and has no impact on SDM. The condition inoperable control rod will not be met and the SRs action is not necessary if the immovable rods remain trippable.

Per Reference 6.2 Section B 3.1.4, The limits on shut down or control rod alignments ensure that the assumptions in the safety analysis will remain valid. The requirements on control rod OPERABILITY ensure that upon reactor trip, the assumed [negative] reactivity will be available and will be inserted. The control rod OPERABILITY requirements (i.e., trippability) are separate from the alignment requirements, which ensure that the RCCAs and banks maintain the correct power distribution and rod alignment. The rod OPERABILITY requirement is satisfied provided the rod will fully insert in the required rod drop time assumed in the safety analysis. Rod control malfunctions that result in the inability to move a rod (e.g., rod lift coil failures), but that do not impact trippability, do not result in rod inoperability.

This LAR will not impact any TS action statements related to ensuring adequate rod alignment to their respective group step counter demand positions. This LAR will not implement any significant changes to how the plant operates relative to rod alignment. Determining SDM will still be required for any inoperable rod, either misaligned or untrippable, per SR 4.1.1.1.1.a and SR 4.1.1.2.a as amended.

2.4 Description of Proposed Change Duke Energy is proposing the following changes to HNP TS 3/4.1.3 to align the TS more closely to the control rod operability definition used in the development of Reference 6.1 and to the initial conditions in HNP safety analyses.

Delete TS action statement 3.1.3.1.c, With more than one rod inoperable, due to a rod control urgent failure alarm or obvious electrical problem in the rod control system existing for greater than 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />, be in HOT STANDBY within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

U.S. Nuclear Regulatory Commission Page 7 of 15 Serial HNP-17-077, Enclosure 1 Modify TS action statement 3.1.3.1.d to read as follows, With one rod misaligned from its group step counter demand height by more than +/- 12 steps (indicated position).

POWER OPERATION may continue.

Modify SR 4.1.1.1.1.a to read as follows, The SHUTDOWN MARGIN shall be determined to be greater than or equal to 1770 pcm: Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after detection of an inoperable control rod(s) and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter while the rod(s) is inoperable. If the inoperable control rod is untrippable, the above required SHUTDOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the untrippable control rod(s);

Modify SR 4.1.1.2.a to read as follows, The SHUTDOWN MARGIN shall be determined to be greater than or equal to the required value: Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after detection of an inoperable control rod(s) and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter while the rod(s) is inoperable. If the inoperable control rod is untrippable, the SHUTDOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the untrippable control rod(s); and.

In addition to reflecting the proposed changes to the TS, the TS 3/4.1.3 Bases are revised for clarity and consistency. Per 10 CFR 50.36, "A summary statement of the bases or reasons for such specifications, other than those covering administrative controls, shall also be included in the application, but shall not become part of the technical specifications." Changes to the TS Bases will be made in accordance with the Technical Specifications Bases Control Program following approval of the requested amendment and are provided in Enclosure 3 for information only.

3.0 Technical Evaluation Reference 6.2, Section B 3.1.4, explains:

The OPERABILITY (i.e., trippability) of the shutdown and control rods is an initial assumption in all safety analyses that assume rod insertion upon reactor trip. Maximum rod misalignment is an initial assumption in the safety analysis that directly affects core power distributions and assumptions of available SDM.

Rod inoperability or misalignment may cause increased power peaking, due to the asymmetric reactivity distribution and a reduction in the total available rod worth for reactor shut down. Therefore, rod alignment and OPERABILITY are related to core operation in design power peaking limits and the core design requirement of a minimum SDM.

Limits on rod alignment have been established, and all rod positions are monitored and controlled during power operation to ensure that the power distribution and reactivity limits defined by the design power peaking and SDM limits are preserved

U.S. Nuclear Regulatory Commission Page 8 of 15 Serial HNP-17-077, Enclosure 1 Shutdown and control rod OPERABILITY and alignment are directly related to power distributions and SDM, which are initial conditions assumed in safety analyses.

Therefore they satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii).

The above analysis in Reference 6.2 is applicable to HNP. HNP assumes the trippability of rods in accident analysis. HNP also uses maximum rod misalignment in safety analysis. These two aspects of rod control are already addressed per HNP TS 3/4.1.3 via the current action statements as follows:

Action statement 3.1.3.1.a delineates actions to be taken if one or more rods are found immovable as a result of excessive friction, mechanical interference, or otherwise known to be untrippable.

Action statements 3.1.3.1.b and 3.1.3.1.d delineates actions to be taken if one or more rods are found misaligned by greater than +/- 12 steps (indicated position).

Also, Reference 6.2, Section B 3.1.1, notes, The minimum required SDM is assumed as an initial condition in safety analyses. This is consistent with HNPs accident analyses. Conditions impacting rod operation that could impact SDM includes a rod being untrippable or a rod being misaligned. Impacts to the SDM will continue to be validated in the event of an untrippable or misaligned rod per SR 4.1.1.1.1.a and 4.1.1.2.a. Only conditions resulting in immovable but trippable rods will be impacted by the proposed change.

Further, TS LCOs remain for the following aspects of rod control and position:

TS LCOs 3.1.3.2 and 3.1.3.3 for Rod Position Indication Systems TS LCO 3.1.3.4 for Rod Drop Time TS LCOs 3.1.3.5 and 3.1.3.6 for shutdown and control rod insertion limits Thus, it is acceptable to remove TS action statement 3.1.3.1.c from the HNP TS scheme and to modify action statement 3.1.3.1.d, SR 4.1.1.1.1.a and SR 4.1.1.2.a as described. Rods that are immovable but trippable, such as when impacted by electrical issues or during the receipt of Rod Control Urgent Failure alarm, should still be considered operable as described in Reference 6.2. There is no impact on HNP accident analysis or SDM due to immovable, but trippable, rods. This change is consistent with Reference 6.1, as these TS actions do not have an analogous action statement within the NUREG-1431 template. All aspects of the Rod Control System that serve as an initial condition of a design basis accident or transient analysis will still be addressed by HNP TS and SR, including rods that are untrippable or are misaligned by greater than +/- 12 steps for any system condition.

Probabilistic Risk Assessment was used to gain risk insights into utilizing the proposed changes to TS. The Harris no-maintenance PRA model was solved accounting for the CRDM being unable to move in response to a demand position change, but with the ability to trip the reactor via rod insertion remaining. Based on this analysis, the incremental conditional core damage probability (ICCDP) for having immovable rods for an extended period of time has been shown to be very low.

The increase in risk associated with this condition results from a very small increase in the likelihood of a reactor trip due to the inability of the control rods to step. The dominant risk contributors are a reactor trip with failure of secondary side heat removal or failure of reactor

U.S. Nuclear Regulatory Commission Page 9 of 15 Serial HNP-17-077, Enclosure 1 coolant pump seal cooling. Note that with the dominant risk being reactor trip, the removal of action statement 3.1.3.1.c and amending of action statement 3.1.3.1.d as proposed will not significantly increase risk, as these action statements as currently written would lead to a reactor shut down for the specified condition. The proposed changes will reduce overall risk by removing unnecessary shut downs initiated by action statement 3.1.3.1.c.

4.0 Regulatory Evaluation 4.1 Applicable Regulatory Requirements/Criteria The following NRC requirements and guidance documents are applicable to the proposed change:

10 CFR 50.36 The regulations at Title 10 of the Code of Federal Regulations (10 CFR) Part 50.36 "Technical specifications," establish the requirements related to the content of the TS. Section 50.36(c)(2) states:

Limiting conditions for operation are the lowest functional capability or performance levels of equipment required for safe operation of the facility. When a limiting condition for operation of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the technical specifications until the condition can be met.

The proposed change will remove HNP TS action statement 3.1.3.1.c and revise action statement 3.1.3.1.d. In both cases, the site will be removing action statements for rods that are in conditions that would still be trippable and would not necessarily impact rod alignment. Thus, LCO 3.1.3.1 continues to be met and the associated actions per the two listed action statements are not required. The HNP TS for the Rod Control System as amended will continue to meet the requirements specified in 10 CFR 50.36(c)(2).

10 CFR 50.36(c)(3) establishes requirements related to SR and states, Surveillance requirements are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met.

The proposed change to SR 4.1.1.1.1.a and 4.1.1.2.a will remove a requirement to perform SDM verification for a condition (immovable but trippable rods) that does not impact SDM, does not impact the facilities ability to operate within the safety limits, and does not impact operation within established LCOs. HNP SR for SDM as amended will meet the requirements specified in 10 CFR 50.36(c)(3).

General Design Criteria 10 CFR Part 50 Appendix A GDC 10 states, The reactor core and associated coolant, control and protection systems shall be designed with appropriate margin to assure that specified

U.S. Nuclear Regulatory Commission Page 10 of 15 Serial HNP-17-077, Enclosure 1 acceptable fuel design limits are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences.

To meet this GDC, HNP credits the control rod drive mechanism as specified in Chapter 15 accident analyses of the HNP FSAR, including the ability to insert rods and rod alignment to control for adverse power distribution. The proposed changes to HNP TS will not alter the plant operation of the Rod Control System in a way that would significantly impact initial assumptions used in accident analysis. All accidents analyses, as documented in Chapter 15 of the HNP FSAR, remain valid.

10 CFR Part 50 Appendix A GDC 20 states, The protection systems shall be designed (1) to initiate automatically the operation of appropriate systems including the reactivity control systems, to assure that specified acceptable fuel design limits are not exceeded as a result of anticipated operational occurrences and (2) to sense accident conditions and to initiate the operation of systems and components important to safety.

The proposed changes to the HNP TS will not alter the design functions of the Rod Control System credited for meeting this GDC. A reactor trip will remove power to the CRDMs of all the full length rod cluster control assemblies. This causes the rods to insert by gravity, which rapidly reduces the reactor power output. Conditions impacting a rods ability to insert would result in entry into HNP TS action statement 3.1.3.1.a., which is not altered by the proposed change. If a rod is identified as untrippable, SR 4.1.1.1.1.a or 4.1.1.2.a (depending on mode) will continue to verify adequate SDM until this condition is addressed.

10 CFR Part 50 Appendix A GDC 23 states, The protection system shall be designed to fail into a safe state or into a state demonstrated to be acceptable on some other defined basis if conditions such as disconnection of the system, loss of energy (i.e., electric power, instrument air), or postulated adverse environments (i.e., extreme heat or cold, fire, pressure, steam, water and radiation) are experienced.

Conditions impacting a rods ability to insert on a loss of power to the CRDM would result in entry into HNP TS action statement 3.1.3.1.a., which is not altered by the proposed change.

10 CFR Part 50 Appendix A GDC 25 states, The protection system shall be designed to assure that specified acceptable fuel design limits are not exceeded for any single malfunction of the reactivity control systems, such as accidental withdrawal (not ejection or dropout) of control rods.

The proposed changes to the HNP TS will not alter the design functions of the Rod Control System credited for meeting this GDC. Reactor shut down by full-length rod insertion is completely independent of the normal control function since the trip breakers interrupt power to the rod mechanisms regardless of existing control signals. Thus, in the postulated accidental withdrawal (assumed to be initiated by a control malfunction), flux, temperature, pressure, level, and flow signals would be generated independently. Any of these signals (trip demands) would operate the breakers to trip the reactor. Conditions impacting a rods ability to insert would result in entry into HNP TS action statement 3.1.3.1.a., which is not altered by the proposed change.

An untrippable rod will still result in validation of SDM per SR 4.1.1.1.1.a or 4.1.1.2.a, depending on mode.

U.S. Nuclear Regulatory Commission Page 11 of 15 Serial HNP-17-077, Enclosure 1 10 CFR Part 50 Appendix A GDC 26 states, Two independent reactivity control systems of different design principles shall be provided. One of the systems shall use control rods, preferably including a positive means for inserting the rods, and shall be capable of reliably controlling reactivity changes to assure that under conditions of normal operation, including anticipated operational occurrences, and with appropriate margin for malfunctions such as stuck rods, specified acceptable fuel design limits are not exceeded. The second reactivity control system shall be capable of reliably controlling the rate of reactivity changes resulting from planned, normal power changes (including xenon burnout) to assure acceptable fuel design limits are not exceeded. One of the systems shall be capable of holding the reactor core subcritical under cold conditions.

To meet this criteria, HNP credits the Rod Control System and boration control systems for controlling reactivity. The proposed change will not alter components or change the design function of either system, and both will still be operated in a manner consistent with the GDC 26.

Thus, the proposed change will not alter HNPs compliance with GDC 26 in any significant way.

10 CFR Part 50 Appendix A GDC 27 states, The reactivity control systems shall be designed to have a combined capability, in conjunction with soluble poison addition by the Emergency Core Cooling System, of reliably controlling reactivity changes to assure that under postulated accident conditions and with appropriate margin for stuck rods the capability to cool the core is maintained.

The proposed changes to the HNP TS will not alter the design functions of the Rod Control System or boration control systems credited for meeting this GDC. TS will remain to ensure rod insertion and rod alignment is maintained, ensuring adequate SDM.

10 CFR Part 50 Appendix A GDC 28 states, The reactivity control systems shall be designed with appropriate limits on the potential amount and rate of reactivity increase to assure that the effects of postulated reactivity accidents can neither (1) result in damage to the reactor coolant pressure boundary greater than limited local yielding nor (2) sufficiently disturb the core, its support structures or other reactor pressure vessel internals to impair significantly the capability to cool the core. These postulated reactivity accidents shall include consideration of rod ejection (unless prevented by positive means), rod dropout, steam line rupture, changes in reactor coolant temperature and pressure, and cold water addition.

The proposed change will not alter rod control or boration control function, and the changes to TS and SR will not alter actions related to any of the initial assumptions used in accident analysis per Chapter 15 of the HNP FSAR.

10 CFR Part 50 Appendix A GDC 29 states, The protection and reactivity control systems shall be designed to assure an extremely high probability of accomplishing their safety functions in the event of anticipated operational occurrences.

The rod control system is designed to allow the rods to drop into the reactor on a loss of power to the CRDM. Conditions impacting a rods ability to insert on a loss of power would result in entry into HNP TS action statement 3.1.3.1.a., which is not altered by the proposed change.

Rod alignment is also not impacted, as it is governed by action statements 3.1.3.1.b and 3.1.3.1.d. SDM will still be verified per SR 4.1.1.1.1.a and 4.1.1.2.a for any conditions impacting

U.S. Nuclear Regulatory Commission Page 12 of 15 Serial HNP-17-077, Enclosure 1 a rods ability to insert on loss of power or maintain alignment relative to position indication. The proposed changes do not impact the systems reliability for accomplishing safety functions.

Thus, the HNP TS as amended will comply with all applicable regulatory requirements and guidance.

4.2 Precedent These proposed changes are based on References 6.1 and 6.2, which contain the improved STS for Westinghouse plants. The changes reflected in References 6.1 and 6.2 result from the experience gained from plant operation using the improved STS and extensive public technical meetings and discussions among the Nuclear Regulatory Commission staff and various nuclear power plant licensees and the Nuclear Steam Supply System Owners Groups.

The proposed changes to HNP TS are consistent with NRC-approved guidance defining the operability of the rod control system. Reference section B 3.1.4 of Reference 6.2.

4.3 No Significant Hazards Consideration Determination Analysis Pursuant to 10 CFR 50.90, Duke Energy Progress, LLC (Duke Energy), is proposing changes to the Shearon Harris Nuclear Power Plant, Unit 1, (HNP) Technical Specification (TS) 3/4.1.1, Reactivity Control Systems Boration Control and to TS 3/4.1.3, Reactivity Control Systems Movable Control Assemblies Group Height.

The proposed changes will delete TS action statement 3.1.3.1.c and modify TS action statement 3.1.3.1.d. The action statements requirements include a plant shut down to address rods that are immovable but still trippable, such as upon receipt of a Rod Control Urgent Failure alarm.

These actions are unnecessary, as rods that are immovable but still trippable still meet the associated Limiting Condition for Operation (LCO). Surveillance requirement (SR), 4.1.1.1.1.a and SR 4.1.1.2.a also contain actions to take for immovable rods that may still be trippable.

Revision to these SRs is proposed to clarify actions are not necessary if rods are immovable but still trippable. These changes are consistent with Revision 4 of NUREG-1431, Standard Technical Specifications - Westinghouse Plants, Volumes 1 and 2. (Agencywide Documents Access and Management System (ADAMS) Accession No. ML12100A222 and ML12100A228, respectively.)

NUREG-1431 Volume 2 specifies that an immovable rod is inoperable only if it is also untrippable. A rod that is immovable but can be tripped, such as one impacted by a Rod Control Urgent Failure alarm or by an obvious electrical problem, would be considered operable. An immovable but trippable rod also has no adverse impact on shutdown margin.

HNP TS differ from NUREG-1431 Volume 2, in that it contains language defining an immovable but trippable rod as inoperable. Action statement 3.1.3.1.c establishes a shut down action for Rod Control Urgent Failure alarm or obvious electrical problems, two conditions that render a rod immovable but trippable. Action statement 3.1.3.1.d includes actions for a rod that is trippable but inoperable. SR 4.1.1.1.1.a and SR 4.1.1.2.a include extra actions for a rod that is immovable or untrippable. Immovable rods that are still trippable should be considered

U.S. Nuclear Regulatory Commission Page 13 of 15 Serial HNP-17-077, Enclosure 1 operable. These TS actions are unnecessary, as NUREG-1431 Volume 2 explains per Section B 3.1.4, and can lead to unnecessary plant shut downs and/or transients.

Duke Energy has evaluated whether a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

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

Response: No.

The proposed activity will delete action statement 3.1.3.1.c from the HNP TS and amend action statement 3.1.3.1.d, SR 4.1.1.1.1.a, and SR 4.1.1.2.a. These TS actions address electrical problems that prevent the Control Rod Drive Mechanism (CRDM) from moving rods. These conditions do not affect the safety functions of the control rods or shutdown margin of the unit.

Rods will still insert into the core on an interruption of power to the CRDM, as occurs in a reactor trip. Also, rod alignment is not impacted, ensuring no change to reactivity.

The proposed activity is removing actions from the HNP TS for conditions that do not impact the plants safety analysis. Rods will still insert into the core on an interruption of power to the CRDM, as occurs in a reactor trip. Also, rod alignment is not impacted, ensuring no change to reactivity or shutdown margin. Since the conditions of these TS actions do not impact the plant safety analysis, the plant shutdown directed by them is unnecessary. The overall probability or consequence of an accident will not be significantly increased by removing the unnecessary TS actions.

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

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

Response: No.

The proposed activity will delete action statement 3.1.3.1.c from the HNP TS and amend action statements 3.1.3.1.d, SR 4.1.1.1.1.a, and SR 4.1.1.2.a. These TS actions address electrical problems that prevent the CRDM from moving rods. These conditions do not affect the safety functions of the control rods. Rods will still insert into the core on an interruption of power to the CRDM, as occurs in a reactor trip. Also, rod alignment is not impacted, ensuring no change to reactivity or shutdown margin.

The proposed change does not involve installation of new equipment or modification of existing equipment, so that no new equipment failure modes are introduced. Also, the proposed change in TS does not result in a change to the way that the equipment or facility is operated that would create new accident initiators.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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3) Does the proposed license amendment involve a significant reduction in a margin of safety?

Response: No.

The proposed activity will delete action statement 3.1.3.1.c from the HNP TS and amend action statement 3.1.3.1.d, SR 4.1.1.1.1.a, and SR 4.1.1.2.a. These actions address electrical problems that prevent the CRDM from moving rods. These conditions do not affect the safety functions of the control rods. Rods will still insert into the core on an interruption of power to the CRDM, as occurs in a reactor trip. Also, rod alignment is not impacted, ensuring no change to reactivity or shutdown margin.

The TS action statements as amended will continue to address the two required safety functions of rod control: to shut down the reactor in the event of a reactor trip, or to maintain proper alignment to ensure even power distribution. TS action statement 3.1.3.1.a will remain to drive actions if untrippable rods are identified. TS action statements 3.1.3.1.b and 3.1.3.1.d will remain to drive actions if misaligned rods are identified. The proposed changes to HNP TS do not significantly impact either rod safety function, and separate TS action statements for both functions will remain in place. Further, the impacted surveillances will continue to be applicable to conditions impacting either rod safety function.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

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.

5.0 Environmental Consideration Duke Energy has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined by 10 CFR 20, or it would change an inspection or surveillance requirement. 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 off site, 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

U.S. Nuclear Regulatory Commission Page 15 of 15 Serial HNP-17-077, Enclosure 1 statement or environmental assessment needs be prepared in connection with the proposed amendment.

6.0 References

1. NUREG-1431, Revision 4, Standard Technical Specifications, Westinghouse Plants, Volume 1, Specifications, dated April 2012 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML12100A222)

2. NUREG-1431, Revision 4, Standard Technical Specifications, Westinghouse Plants, Volume 2, Bases, dated April 2012 (ADAMS Accession No. ML12100A228)

U.S. Nuclear Regulatory Commission Serial HNP-17-077, Enclosure 2 SERIAL HNP-17-077 ENCLOSURE 2 PROPOSED TECHNICAL SPECIFICATION CHANGES SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 DOCKET NO. 50-400 RENEWED LICENSE NUMBER NPF-63 3 PAGES PLUS THE COVER

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U.S. Nuclear Regulatory Commission Serial HNP-17-077, Enclosure 3 SERIAL HNP-17-077 ENCLOSURE 3 TECHNICAL SPECIFICATION BASES CHANGE (FOR INFORMATION ONLY)

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

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