HNP-16-066, License Amendment Request to Relocate Technical Specification Cycle-Specific Parameters to the Core Operating Limits Report

License Amendment Request to Relocate Technical Specification Cycle-Specific Parameters to the Core Operating Limits Report

ML16337A249
Person / Time
Site:	Harris
Issue date:	12/02/2016
From:	Hamilton T Duke Energy Progress
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
HNP-16-066
Download: ML16337A249 (45)

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 December 2, 2016 Serial: HNP-16-066 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 to Relocate Technical Specification Cycle-Specific Parameters to the Core Operating Limits Report Ladies and Gentlemen:

In accordance with the provisions of 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 would relocate selected figures and values from the TS to the Core Operating Limits Report (COLR), including TS Figure 2.1-1 cited in TS 2.1.1, Safety Limits - Reactor Core, selected portions of Table 2.2-1 Note 1 on Overtemperature T (OTT) and Note 3 on Overpower T (OPT) cited in TS 2.2.1, Limiting Safety System Settings - Reactor Trip System Instrumentation Setpoints, and Departure from Nucleate Boiling (DNB) values cited in TS 3.2.5, DNB Parameters. These changes are consistent with the intent of Nuclear Regulatory Commission (NRC)-approved Technical Specification Task Force (TSTF) Improved Standard Technical Specifications Change Traveler TSTF-339, Relocate TS Parameters to COLR. As a result of the above changes, TS 6.9.1.6, Core Operating Limits Report, and associated TS Bases will be revised to reflect the above proposed changes.

Furthermore, editorial changes are proposed throughout the HNP TS to remove all reference to plant procedure PLP-106, Technical Specification Equipment List Program and Core Operating Limits Report, as it pertains to the COLR. Going forward, the COLR will no longer be contained in PLP-106, aligning with the Duke Energy fleet procedure governing core design and design deliverable documents.

This amendment request also proposes to delete requirements from HNP Administrative Control TS 6.7, Safety Limit Violation, that duplicate requirements found in regulation 10 CFR 50.36.

The proposed changes are consistent with the intent of NRC-approved TSTF-5, Delete Safety Limit Violation Notification Requirements, Revision 1.

The proposed changes have been evaluated in accordance with 10 CFR 50.91(a)(1) using criteria in 10 CFR 50.92(c), and it has been determined that the proposed changes involve no significant hazards consideration. Attachment 1 of this license amendment request provides Duke Energys evaluation of the proposed changes. Attachment 2 provides a copy of the

U.S. Nuclear Regulatory Commission Page 2 HNP-16-066 proposed TS changes. Attachment 3 provides a copy of the TS Bases markup based on the proposed changes.

Approval of the proposed license amendment is requested by December 1, 2017, to support the planned refueling outage scheduled for April 2018. The amendment shall be implemented within 90 days following approval.

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.

Please refer any questions regarding this submittal to 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 December , 2016.

Sincerely, Tanya M. Hamilton Attachments:

1. Evaluation of the Proposed Change

2. Proposed Technical Specification Changes

3. Proposed Technical Specification Bases Changes cc: Mr. C. Jones, NRC Sr. Resident Inspector, HNP Mr. W. L. Cox, III, Section Chief N.C. DHSR Ms. M. Barillas, NRC Project Manager, HNP Ms. C. Haney, NRC Regional Administrator, Region II

U.S. Nuclear Regulatory Commission Serial HNP-16-066 HNP-16-066 ATTACHMENT 1 EVALUATION OF THE PROPOSED CHANGE SHEARON HARRIS NUCLEAR POWER PLANT / UNIT 1 DOCKET NO. 50-400 RENEWED LICENSE NUMBER NPF-063

U.S. Nuclear Regulatory Commission Page 1 of 6 Serial HNP-16-066 Evaluation of the Proposed Change

Subject:

License Amendment Request to Relocate Technical Specification Cycle-Specific Parameters to the Core Operating Limits Report 1.0

SUMMARY

DESCRIPTION In accordance with the provisions of 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 would relocate selected figures and values from the TS to the Core Operating Limits Report (COLR), including TS Figure 2.1-1 cited in TS 2.1.1, Safety Limits - Reactor Core, selected portions of Table 2.2-1 Note 1 on Overtemperature T (OTT) and Note 3 on Overpower T (OPT) cited in TS 2.2.1, Limiting Safety System Settings - Reactor Trip System Instrumentation Setpoints, and Departure from Nucleate Boiling (DNB) values cited in TS 3.2.5, DNB Parameters. These changes are consistent with the intent of Nuclear Regulatory Commission (NRC)-approved Technical Specification Task Force (TSTF) Improved Standard Technical Specifications Change Traveler TSTF-339, Relocate TS Parameters to COLR. As a result of the above changes, TS 6.9.1.6, Core Operating Limits Report, and associated TS Bases will be revised to reflect the above proposed changes.

Furthermore, editorial changes are proposed throughout the HNP TS to remove all reference to plant procedure PLP-106, Technical Specification Equipment List Program and Core Operating Limits Report, as it pertains to the COLR. Going forward, the COLR will no longer be contained in PLP-106, aligning with the Duke Energy fleet procedure governing core design and design deliverable documents.

This amendment request also proposes to delete requirements from HNP Administrative Control TS 6.7, Safety Limit Violation, that duplicate requirements found in regulation 10 CFR 50.36.

The proposed changes are consistent with the intent of NRC-approved TSTF-5, Delete Safety Limit Violation Notification Requirements, Revision 1.

2.0 DETAILED DESCRIPTION Duke Energy proposes changes to the HNP TS as follows:

• TS Figure 2.1-1 is deleted and relocated to the COLR.

• TS Safety Limit 2.1.1 is revised to read:

2.1.1 The combination of THERMAL POWER, pressurizer pressure, and the highest operating loop coolant temperature (Tavg) shall not exceed the limits specified in the COLR; and the following Safety Limits shall not be exceeded:

a. The departure from nucleate boiling ratio (DNBR) shall be maintained 1.141 for the HTP DNB correlation.

U.S. Nuclear Regulatory Commission Page 2 of 6 Serial HNP-16-066

b. The peak centerline temperature shall be maintained < [(2790 -

17.9 x P - 3.2 x B) x 1.8 + 32] F where P is the maximum weight percent of Gadolinia (%) and B is the maximum pin burnup (GWD/MTU).

APPLICABILITY: MODES 1 and 2.

ACTION:

If Safety Limit 2.1.1 is violated, restore compliance and be in HOT STANDBY within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

• The overpower and overtemperature T trip constants and time constants in Table 2.1-1 of TS 2.2.1 are deleted and relocated to the COLR.

• TS 3.2.5 DNB-related parameters are deleted and relocated to the COLR.

• Administrative Control TS 6.9.1.6 is revised to reflect the above changes.

• Editorial changes throughout TS to remove reference to HNP plant procedure PLP-106 as it pertains to the COLR.

• Administrative Control TS 6.7.1 is deleted to remove duplicative reporting and restart requirements to those already contained in the regulations.

Relocation of cycle-specific parameters from the TS to the COLR, a licensee-controlled document subject to the requirements of TS 6.9.1.6 and the provisions of 10 CFR 50.59, would afford Duke Energy the flexibility to revise cycle-specific parameters that are in accordance with NRC-approved methodologies without the need for license amendments. The COLR is required to be submitted to the NRC for each reload cycle per TS 6.9.1.6, including any mid-cycle revisions or supplements to the NRC, unless otherwise approved by the Commission.

3.0 TECHNICAL EVALUATION

Basis for Proposed Change NRC Generic Letter 88-16, Removal of Cycle-Specific Parameter Limits From Technical Specifications, dated October 4, 1988, provides guidance to licensees for the removal of cycle-dependent variables from the TS provided that these values are included in a COLR and are determined with NRC-approved methodologies referenced in the TS. Westinghouse Electric Company (Westinghouse) subsequently developed WCAP-14483, Generic Methodology for Expanding Core Operating Limits Report, describing how cycle-specific parameters may be relocated to the COLR. WCAP-14483 was accepted for referencing by the NRC on January 19, 1999. The Safety Evaluation Report, contained in the January 19, 1999, NRC letter approving WCAP-14483-A, concluded that additional information contained in the TS may be relocated to the COLR.

The limits on the parameters which are removed from the TS and added to the COLR must be developed and justified using NRC-approved methodologies. All accident analyses, performed

U.S. Nuclear Regulatory Commission Page 3 of 6 Serial HNP-16-066 in accordance with these methodologies, must meet the applicable NRC-approved limits of the safety analysis. The removal of parameter limits from the TS and their addition to the COLR does not obviate the requirement to operate within these limits. Furthermore, any changes to those limits must be performed in accordance with TS 6.9.1.6.3. If any of the applicable limits of the safety analysis are not met, prior NRC approval of the change is required, as is the case for a license amendment request. For more routine modifications, where NRC-approved methodologies and limits of the safety analysis remain applicable, the potentially burdensome and lengthy process of amending the TS may be avoided. The requested changes are essentially administrative in nature; therefore, the required level of safety will be maintained.

Applicability of TSTF-339 The following requested changes are based upon the NRC-approved Westinghouse Owners Group (WOG) Technical Specifications Task Force (TSTF)-339, Relocated TS Parameters to the COLR Consistent with WCAP-14483, Revision 2, and Westinghouse WCAP-14483-A:

1. Revise TS 3.2.5 to relocate the pressurizer pressure, RCS average temperature (Tavg),

and RCS total flow rate values to the COLR. The minimum limit for total flow based on that used in the reference safety analysis will be retained in the TS.

2. Revise TS Table 2.2-1, Notes 1 and 3, to relocate the overtemperature T and overpower T (K) constant values and dynamic compensation () values, and the breakpoint and slope values for the f(l) penalty function(s) to the COLR.

3. Revise TS 2.1 Safety Limits, and the associated bases, to relocate Figure 2.1-1 to the COLR and replace it with more specific requirements regarding the safety limits (i.e., the fuel DNB design basis and the fuel centerline melt design basis). The NRC-approved methodologies used to derive the parameters in the figure will be referenced in the Reporting Requirements section of the TS.

Duke Energy has reviewed TSTF-339, Revision 2, and concluded that the TS changes as outlined in WCAP-14483-A are applicable to HNP. The above proposed changes are also consistent with Revision 4 of NUREG-1431, Standard Technical Specifications - Westinghouse Plants (Agencywide Documents Access and Management System (ADAMS) Accession No. ML12100A222), with minor format and structure differences due to HNP having not converted to Improved Technical Specifications (ITS), including the TS 2.1.1 numbering scheme (alphabetical versus numerical) and the use of TS 6.7.1 to identify Safety Limit Violation actions.

Editorial Changes related to PLP-106 The HNP COLR is currently contained within plant procedure PLP-106 as an attachment. To maintain consistency with the Duke Energy fleet procedure governing core design deliverable documents, the COLR will no longer be contained within PLP-106, but will rather be its own entity. As such, all reference to PLP-106 in HNP TS pertaining to the COLR will be deleted.

Reference to PLP-106 in TS as it pertains to the Technical Specification Equipment List Program is not affected by this license amendment request, aside from those instances where the current procedure title is given.

The COLR is a licensee-controlled document, subject to the requirements of TS 6.9.1.6 and the provisions of 10 CFR 50.59. It will continue to be submitted to the NRC for each reload cycle, including any mid-cycle revisions or supplements to the NRC, unless otherwise approved by the Commission. Therefore, the deletion from TS of reference PLP-106 as it pertains to the COLR is

U.S. Nuclear Regulatory Commission Page 4 of 6 Serial HNP-16-066 an administrative change that aligns with the corresponding wording of NUREG-1431, Volume 1, Revision 4.

Applicability of TSTF-5 The requested deletion of requirements from Administrative Control TS 6.7, Safety Limit Violation, is based upon the NRC-approved TSTF-5, Delete Safety Limit Violation Notification Requirements, Revision 1.

Duke Energy has reviewed TSTF-5, Revision 1, and concluded that the intent of the TS changes to be applicable to HNP. The proposed changes delete administrative actions from the TS that duplicate the requirements to report safety limit violations and requirements to preclude restart after a safety limit violation without NRC approval. These requirements are also more restrictive than those already contained in the regulations in that they require the report to be submitted to the NRC within 14 days of the violation. The 10 CFR 50.36 reporting requirements require the licensee to notify the NRC as required by 10 CFR 50.72 and submit a Licensee Event Report to the NRC as required by 10 CFR 50.73. Therefore, appropriate reporting would be made to the NRC in accordance with the regulations in the event a TS safety limit was violated. In addition, 10 CFR 50.36 states that operations must not be resumed until authorized by the Commission. The removal of the duplicate reporting and restart requirements from TS is a simplification of the TS and a reduction in administrative burden to track duplicated requirements. It also aligns with the requirements as presented in NUREG-1431, Volume 1, Revision 4.

4.0 REGULATORY ANALYSIS

4.1 No Significant Hazards Consideration Determination Pursuant to 10 CFR 50.90, Duke Energy Progress, LLC (Duke Energy), proposes a license amendment request (LAR) for the Shearon Harris Nuclear Power Plant, Unit 1 (HNP) Technical Specifications (TS). The proposed LAR relocates selected figures and values from the TS to the Core Operating Limits Report (COLR), including TS Figure 2.1-1 cited in TS 2.1.1, Safety Limits - Reactor Core, selected portions of Table 2.2-1 Note 1 on Overtemperature T (OTT) and Note 3 on Overpower T (OPT) cited in TS 2.2.1, Limiting Safety System Settings -

Reactor Trip System Instrumentation Setpoints, and Departure from Nucleate Boiling (DNB) values cited in TS 3.2.5, DNB Parameters. These changes are consistent with the intent of Nuclear Regulatory Commission (NRC)-approved Technical Specification Task Force (TSTF)

Improved Standard Technical Specifications Change Traveler TSTF-339, Relocate TS Parameters to COLR. This LAR also proposes the deletion of duplicative notification, reporting, and restart requirements from the Administrative Controls section of TS, consistent with the intent of NRC-approved TSTF-5, Delete Safety Limit Violation Notification Requirements.

Furthermore, editorial changes are proposed throughout the HNP TS to remove all reference to plant procedure PLP-106, Technical Specification Equipment List Program and Core Operating Limits Report, as it pertains to the COLR. As a result of the above changes, TS 6.9.1.6, Core Operating Limits Report, and associated TS Bases will be revised to reflect the above proposed changes.

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

U.S. Nuclear Regulatory Commission Page 5 of 6 Serial HNP-16-066

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

The proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated. The proposed changes are administrative in nature, facilitate improved content and presentation of Administrative controls, and alter only the format and location of cycle-specific parameter figures and limits from the TS to the COLR. This relocation does not result in the alteration of the design, material, or construction standards that were applicable prior to the change. The proposed changes will not result in modification of any system interface that would increase the likelihood of an accident since these events are independent of the proposed change. The proposed amendment will not change, degrade, or prevent actions, or alter any assumptions previously made in evaluating the radiological consequences of an accident described in the Final Safety Analysis Report (FSAR).

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

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

The proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated. The proposed changes do not involve any change to the configuration or method of operation of any plant equipment. Accordingly, no new failure modes have been defined for any plant system or component important to safety nor has any new limiting single failure been identified as a result of the proposed changes. Also, there will be no change in types or increase in the amounts of any effluents released offsite.

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

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

The proposed changes do not involve a significant reduction in a margin of safety.

Previously-approved methodologies will continue to be used in determination of cycle-specific core operating limits that are present in the COLR. The proposed changes are administrative in nature and will not affect the plant design or system operating parameters. As such, there is no detrimental impact on any equipment design parameter and the plant will continue to be operated within prescribed limits.

Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

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

U.S. Nuclear Regulatory Commission Page 6 of 6 Serial HNP-16-066 4.2 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 concluded that the proposed amendment meets the criteria provided by 10 CFR 51.22(c)(9) for categorical exclusion from the requirement for an Environmental Impact Statement. 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 effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure.

U.S. Nuclear Regulatory Commission Serial HNP-16-066 HNP-16-066 ATTACHMENT 2 PROPOSED TECHNICAL SPECIFICATION CHANGES SHEARON HARRIS NUCLEAR POWER PLANT / UNIT 1 DOCKET NO. 50-400 RENEWED LICENSE NUMBER NPF-063

specified in the COLR; and the following Safety Limits shall not be exceeded:

ADD: INSERT 2 ADD: INSERT 1

INSERT 1

a. dŽŽŽEZŽ HTP DNB correlation.

b. The peak centerline temperature shall be maintained < [ (2790 - W-  &

WŽ'Ž

(GWD/MTU).

INSERT 2 If Safety Limit ŽŽŽ,Kd^dEzŽ.

ADD:

"This figure is deleted from Technical Specifications and relocated to the COLR."

No Changes made to this page ADD: "The values denoted with [*] are specified in the COLR."

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ADD: "The values denoted with [*] are specified in the COLR."

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the limit specified in the COLR, and the limit specified in the COLR*, and and greater than or equal to the limit specified in the COLR.

TABLE 3.3-4 (Continued)

TABLE NOTATIONS

• Time constants utilized in the lead-lag controller for Steam Line Pressure--Low are

 50 seconds and 5 seconds. CHANNEL CALIBRATION shall ensure that these time constants are adjusted to these values.

• • The time constant utilized in the rate-lag controller for Steam Line Pressure-Negative Rate--High is 50 seconds. CHANNEL CALIBRATION shall ensure that this time constant is adjusted to this value.

The indicated values are the effective, cumulative, rate-compensated pressure drops as seen by the comparator.

NOTE 1: If the as-found channel setpoint is outside its predefined as-found tolerance, the channel shall be evaluated to verify that it is functioning as required before returning the channel to service.

NOTE 2: The instrument channel setpoint shall be reset to a value that is within the as-left tolerance around the Trip Setpoint in Table 3.3-4 (Nominal Trip Setpoint (NTSP))

at the completion of the surveillance; otherwise, the channel shall be declared inoperable. Setpoints more conservative than the NTSP are acceptable provided that the as-found and as-left tolerances apply to the actual setpoint implemented in the surveillance procedures (field setting) to confirm channel performance. The methodologies used to determine NTSPs and the as-found and the as-left tolerances are specified in EGR-NGGC-0153, Engineering Instrument Setpoints.

The as-found and as-left tolerances are specified in PLP-106, Technical Specification Equipment List Program and Core Operating Limits Report.

SHEARON HARRIS - UNIT 1 3/4 3-36 Amendment No. 146

ADD: "Deleted."

ADD: INSERT 3 INSERT3

i. ReactorCoreSafetyLimitsFigureforSpecification2.1.1.

j. OvertemperatureTandOverpowerTsetpointparametersandtimeconstantvaluesfor

Specification2.2.1.

k. ReactorCoolantSystempressure,temperature,andflowDeparturefromNucleateBoiling(DNB)

limitsforSpecification3/4.2.5.

U.S. Nuclear Regulatory Commission Serial HNP-16-066 HNP-16-066 ATTACHMENT 3 PROPOSED TECHNICAL SPECIFICATION BASES CHANGES SHEARON HARRIS NUCLEAR POWER PLANT / UNIT 1 DOCKET NO. 50-400 RENEWED LICENSE NUMBER NPF-063

The DNBR safety limit for high thermal performance fuel is 1.141 for the Siemens HTP correlation (Reference 1).

Insert new paragraph 1:

The restrictions of this safety limit also prevent fuel centerline melting. Fuel centerline melting occurs when the local LHR, or power peaking, in a region of the fuel is high enough to cause the fuel centerline temperature to reach the melting point of the fuel.

Expansion of the pellet upon centerline melting may cause the pellet to stress the cladding to the point of failure, allowing an uncontrolled release of activity to the reactor coolant. The fuel centerline temperature limit is a function of weight percent of Gadolinia and pin burnup as presented in Reference 2 and approved for use at HNP per Reference 3.

Insert new paragraph 2:

The safety limit figure provided in the COLR shows the loci of points of Fraction of Rated Thermal power, RCS Pressure, and average temperature for which the minimum DNBR is not less than the safety analyses limit, that fuel centerline temperature remains below melting, that the average enthalpy in the hot leg is less than or equal to the enthalpy of saturated liquid, and that the exit quality is within the limits defined by the DNBR correlation.

The reactor core safety limits are established to preclude violation of the following fuel design criteria:

a. There must be at least 95% probability at a 95% confidence level (the 95 / 95 DNB criteria) that the hot fuel rod in the core does not experience DNB; and

b. There must be at least a 95% probability at a 95% confidence level that the hot fuel pellet in the core does not experience centerline fuel melting.

The reactor core safety limits are used to define the various RPS functions such that the above criteria are satisfied during steady state operation and Condition I and II events. To ensure that the RPS precludes the violation of the above criteria, additional criteria are applied to the Over Temperature and Overpower T reactor trip functions. That is, it must be demonstrated that the average enthalpy in the hot leg is less than or equal to the saturation enthalpy and that the core exit quality is within the limits defined by the DNBR correlation.

Appropriate functioning of the RPS ensures that for variations in the THERMAL POWER, RCS Pressure, RCS average temperature, RCS flow rate, and I that the reactor core safety limits will be satisfied during steady state operation and Condition I and II events.

the COLR.

Insert new Section:

References

1. EMF-92-153(P)(A), "HTP: Departure from Nucleate Boiling Correlation for High Thermal Performance Fuel."

2. XN-NF-79-56(P)(A), Revision 1, "Gadolinia Fuel Properties for LWR Safety Evaluation."

3. XN-NF-85-92(P)(A), "Exxon Nuclear Uranium Dioxide/Gadolinia Irradiation Examination and Thermal Conductivity Results."

adjusted for measurement uncertainty then compared to the analytical limits specified in the COLR.

3/4.3 INSTRUMENTATION BASES Note 2 requires that the as-left channel setting be reset to a value that is within the as-left tolerances about the Trip Setpoint in Table 3.3-4 or within as-left tolerances about a more conservative actual (field) setpoint. As-left channel settings outside the as-left tolerances of PLP-106 and the surveillance procedures cause the channel to be INOPERABLE.

A tolerance is necessary because no device perfectly measures the process. Additionally, it is not possible to read and adjust a setting to an absolute value due to the readability and/or accuracy of the test instruments or the ability to adjust potentiometers. The as-left tolerance is considered in the setpoint calculation. Failure to set the actual plant trip setpoint to within as-left the tolerances of the NTSP or within as-left tolerances of a more conservative actual field setpoint would invalidate the assumptions in the setpoint calculation, because any subsequent instrument drift would not start from the expected as-left setpoint. The determination will consider whether the instrument is degraded or is capable of being reset and performing its specified safety function.

If the channel is determined to be functioning as required (i.e., the channel can be adjusted to within the as-left tolerance and is determined to be functioning normally based on the determination performed prior to returning the channel to service), then the channel is OPERABLE and can be restored to service. If the as-left instrument setting cannot be returned to a setting within the prescribed as-left tolerance band, the instrument would be declared inoperable.

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

Limiting Trip Setpoint (LTSP) is generic terminology for the setpoint value calculated by means of the setpoint methodology documented in EGR-NGGC-0153. HNP uses the plant-specific term NTSP in place of the generic term LTSP. The NTSP is the LTSP with margin added, and is always equal to or more conservative than the LTSP. The NTSP may use a setting value that is more conservative than the LTSP, but for Technical Specification compliance with 10 CFR 50.36, the plant-specific setpoint term NTSP is cited in Note 2. The NTSP meets the definition of a Limiting Safety System Setting per 10 CFR 50.36 and is a predetermined setting for a protective channel chosen to ensure that automatic protective actions will prevent exceeding Safety Limits during normal operation and design basis anticipated operational occurrences, and assist the Engineered Safety Features Actuation System in mitigating the consequences of accidents. The Allowable Value is the least conservative value of the as-found setpoint that the channel can have when tested, such that a channel is OPERABLE if the as-found setpoint is within the as-found tolerance and is conservative with respect to the Allowable Value during a CHANNEL CALIBRATION or CHANNEL OPERATIONAL TEST. As such, the Allowable Value differs from the NTSP by an amount greater than or equal to the expected instrument channel uncertainties, such as drift, during the surveillance interval. In this manner, the actual NTSP setting ensures that a Safety Limit is not exceeded at any given point of time as long as the channel has not drifted beyond expected tolerances during the surveillance interval. Although the channel is OPERABLE under these circumstances, the trip setpoint must be left adjusted to a value within the as-left tolerance band, in accordance with uncertainty assumptions stated in the setpoint methodology (as-left criteria), and confirmed to be operating within the statistical allowances of the uncertainty terms assigned (as-found criteria).

Field setting is the term used for the actual setpoint implemented in the plant surveillance procedures, where margin has been added to the calculated field setting. The as-found and as-left tolerances apply to the field settings implemented in the surveillance procedures to confirm SHEARON HARRIS - UNIT 1 B 3/4 3-2a Amendment No. 146