RA-22-0081, License Amendment Request to Address Administrative Changes to the Operating License and Technical Specifications

License Amendment Request to Address Administrative Changes to the Operating License and Technical Specifications

ML23038A186
Person / Time
Site:	Harris
Issue date:	02/07/2023
From:	Haaf T Duke Energy Progress
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
RA-22-0081
Download: ML23038A186 (1)

Text

Thomas P. Haaf

( ~ DUKE Site Vice President Harris Nuclear Plant ENERGY 5413 Shearon Harris Rd New Hill, NC 27562-9300

984-229-2512

10 CFR 50.90

February 7, 2023 Serial: RA-22-0081

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 Address Administrative Changes to the Operating License and Technical Specifications

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 Shearon Harris Nuclear Power Plant, Unit 1 (HNP), Operating License and Technical Specifications (TS). The proposed amendment to the HNP TS would remove the reference to Duke Energy procedure EGR-NGGC-0153, Engineering Instrument Setpoints, which was superseded by procedure AD-EG-ALL-1153, Engineering Instrument Setpoint/Uncertainty Calculations, a document incorporated by reference in the HNP Final Safety Analysis Report (FSAR). This amendment w ill also remove the reference to Attachment 1, TDI Diesel Engine Requirements, of the Renewed Facility Operating License in accordance with License Amendment No. 53, issued by letter dated January 12, 1995 (ADAMS Accession No. ML020570303).

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 concluded that the proposed changes involve no significant hazards consideration.

of this license amendment request provides Duke Energys evaluation of the proposed changes. Enclosure 2 provides a copy of the proposed TS changes (mark-up). provides a copy of the proposed Operating License changes (mark-up). Enclosure 4 provides a copy of the proposed TS Bases changes (mark-up) for information only, as they will be implemented in accordance with the TS Bases Control Program upon implementation of the amendment.

Approval of the proposed license amendment is requested within twelve months of acceptance.

The amendment shall be implemented within 90 days from approval.

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

This letter contains no regulatory commitments.

U.S. Nuclear Regulatory Commission Page 2 of 2 Serial: RA-22-0081

Please refer any questions regarding this submittal to Ryan Treadway, Director - Nuclear Fleet Licensing, at 980-373-5873.

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

Executed on February 7, 2023.

Sincerely,

~ TITomas-P:-Haa -

Site Vice President Harris Nuclear Plant

Enclosures:

1. Evaluation of the Proposed Changes

2. Proposed Technical Specification Changes (Mark-up)

3. Proposed Operating License Changes (Mark-up)

4. Proposed Technical Specification Bases Changes (Mark-up)

cc: P. Boguszewski, Senior NRC Resident Inspector, HNP D. Crowley, Radioactive Materials Branch Manager, N.C. DHSR M. Mahoney, NRC Project Manager, HNP L. Dudes, NRC Regional Administrator, Region II U.S. Nuclear Regulatory Commission Serial: RA-22-0081

ENCLOSURE 1

EVALUATION OF THE PROPOSED CHANGES

SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1

DOCKET NO. 50-400

RENEWED LICENSE NUMBER NPF-63

5 PAGES PLUS THE COVER

U.S. Nuclear Regulatory Commission Page 1 of 5 Serial: RA-22-0081

Evaluation of the Proposed Changes

License Amendment Request to Address Administrative Changes to the Operating License and Technical Specifications 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 Shearon Harris Nuclear Power Plant, Unit 1 (HNP), Operating License and Technical Specifications (TS). The proposed amendment to the HNP TS would remove the reference to Duke Energy procedure EGR-NGGC-0153, Engineering Instrument Setpoints, which was superseded by procedure AD-EG-ALL-1153, Engineering Instrument Setpoint/Uncertainty Calculations, a document incorporated by reference in the HNP Final Safety Analysis Report (FSAR). This amendment w ill also remove the reference to Attachment 1, TDI Diesel Engine Requirements, of the Renewed Facility Operating License in accordance with License Amendment No. 53, issued by letter dated January 12, 1995 (ADAMS Accession No. ML020570303).

2.0 DETAILED DESCRIPTION

2.1 Current Operating License and Technical Specifications 2.1.1 EGR-NGGC-0153

Prior to being superseded, EGR-NGGC-0153 implemented, in part, the HNP commitment to Regulatory Guide 1.105, "Instrument Setpoints," Revision 1, as described in Section 1.8 of the HNP FSAR. Specifically, EGR-NGGC-0153 impl emented the program requirements for both methodology and scope concerning instrument uncertainty and scaling calculations for HNP.

The methodology requirements provided a description of the detailed rules and plant specific criteria involved in instrument loop uncertainty analysis and setpoint determination. The scoping criteria defined those instrument loops that, as a minimum, shall have documented instrument uncertainty and scaling calculations completed to ensure these vital systems were operating within established safety limits. This setpoint methodology was retained in Duke Energy fleet procedure AD-EG-ALL-1153 when EGR-NGGC-0153 was superseded.

EGR-NGGC-0153 is referenced in Note 8 of HNP TS Table 2.2-1, Reactor Trip System Instrumentation Trip Setpoints, and in Note 2 of TS Table 3.3-4, Engineered Safety Features Actuation System Instrumentation Trip Setpoints.

2.1.2 Operating License Attachment 1

The Renewed Facility Operating License was amended per License Amendment No. 53 to delete License Condition 2.C.(8), Transamerica Delaval, Inc., (TDI) Diesel Generators, and, TDI Diesel Engine Requirements. Reference to this Attachment 1 remains listed on page 12 of the Renewed Facility Operating License.

U.S. Nuclear Regulatory Commission Page 2 of 5 Serial: RA-22-0081

2.2 Reason for the Proposed Change

2.2.1 EGR-NGGC-0153

Procedure EGR-NGGC-0153 was superseded by Du ke Energy fleet procedure AD-EG-ALL-1153. This amendment would replace the reference to procedure EGR-NGGC-0153 in Note 8 of HNP TS Table 2.2-1 and Note 2 of TS Table 3.3-4 to instead reference the FSAR, which includes procedure AD-EG-ALL-1153 as a document incorporated by reference.

The proposed change is not a change to the methodology utilized in HNP TS Table 2.2-1 Note 8 and TS Table 3.3-4 Note 2, but rather an administrative change to the TS notes to reference the updated location of the associated methodology.

2.2.2 Operating License Attachment 1

Renewed Facility Operating License NPF-63 was amended by License Amendment No. 53 to delete License Condition 2.C.(8) and Attachment 1. Although License Condition 2.C.(8) was deleted, Attachment 1 to the Renewed Facility Operating License unintentionally remains listed on page 12 of the license and appears in the authority copy of the Renewed Facility Operating License in ADAMS (ADAMS Accession No. ML052860283).

2.3 Description of the Proposed Change

2.3.1 EGR-NGGC-0153

Note 8 in HNP TS Table 2.2-1 and Note 2 in TS Table 3.3-4 will be revised to replace the reference to EGR-NGGC-0153 with a reference to the HNP FSAR, which includes procedure AD-EG-ALL-1153 as a document incorporated by reference.

2.3.2 Operating License Attachment 1 of the Renewed Facility Operating License, and the reference to Attachment 1 on page 12 of the license, will be deleted as previously approved per License Amendment No. 53.

3.0 TECHNICAL EVALUATION

3.1 EGR-NGGC-0153

The methodology utilized in HNP TS Table 2.2-1 Note 8 and TS Table 3.3-4 Note 2 was relocated to fleet procedure AD-EG-ALL-1153 when procedure EGR-NGGC-0153 was superseded. No changes were made to the methodology. The proposed change to the HNP TS to reference the HNP FSAR in place of proc edure EGR-NGGC-0153 does not adversely alter the current TS or introduce any new TS requirements. The relocated methodology previously captured in EGR-NGGC-0153 is maintained in fleet procedure AD-EG-ALL-1153, which is a document incorporated by reference in the HNP FSAR. As a document incorporated by reference in the HNP FSAR, it is subject to the update and reporting requirements of 10 CFR 50.71(e) and change controls of 10 CFR 50.59. There is no impact on plant operations or systems as a result of the proposed change.

U.S. Nuclear Regulatory Commission Page 3 of 5 Serial: RA-22-0081

3.2 Operating License Attachment 1

Deletion of Attachment 1 to the Renewed Facility Operating License was previously approved per License Amendment No. 53. The change proposed in this submittal would reflect the approved removal of this attachment and is strictly administrative.

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements and Guidance

10 CFR 50.36, Technical specifications

The NRC's regulatory requirements related to the content of the TS are set forth in Title 10 of the Code of Federal Regulations (10 CFR) Section 50.36, "Technical specifications." This regulation requires that the TS include items in the following five specific categories: (1) safety limits, limiting safety system settings, and limitin g control settings, (2) limiting conditions for operation (LCO), (3) surveillance requirements, (4) design features, and (5) administrative controls.

Per 10 CFR 50.36(c)(2)(ii), a TS LCO must be established for each item meeting one or more of the following criteria:

Criterion 1: Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

Criterion 2: A process variable, design feature, or operating restriction that is an initial condition of a Design Basis Accident or Transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

Criterion 3: A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a Design Basis Accident or Transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

Criterion 4: A structure, system, or component which operating experience or probabilistic safety assessment has shown to be si gnificant to public health and safety.

Conclusion

Duke Energy has evaluated the proposed changes against the applicable regulatory requirements described above. Based on this evaluation, there is reasonable assurance that the health and safety of the public will remain unaffected following the approval of these proposed changes.

4.2 Significant Hazards Consideration

Pursuant to 10 CFR 50.90, Duke Energy Progress, LLC (Duke Energy), hereby requests a revision to the Shearon Harris Nuclear Power Plant, Unit 1 (HNP), Operating License and U.S. Nuclear Regulatory Commission Page 4 of 5 Serial: RA-22-0081

Technical Specifications (TS). The proposed amendment to the HNP TS would remove the reference to Duke Energy procedure EGR-NGGC-0153, Engineering Instrument Setpoints, which was superseded by procedure AD-EG-ALL-1153, Engineering Instrument Setpoint/Uncertainty Calculations, a document incorporated by reference in the HNP Final Safety Analysis Report (FSAR). This amendment w ill also remove the reference to Attachment 1, TDI Diesel Engine Requirements, of the Renewed Facility Operating License in accordance with License Amendment No. 53, issued by letter dated January 12, 1995 (ADAMS Accession No. ML020570303).

Duke Energy has evaluated whether 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:

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

Response: No.

The impacts of the proposed changes are administrative and do not affect how plant equipment is operated or maintained, nor do the changes impact the intent of the Operating License or TS.

No changes are proposed that will make changes to the physical plant or analytical methods.

The proposed changes do not adversely affect accident initiators or precursors nor alter the design assumptions, conditions, and configuration of the facility or the manner in which the plant is operated and maintained. The proposed changes do not alter or prevent the ability of structures, systems, or components from performing their intended function to mitigate the consequences on an initiating event with the assumed acceptance limits. The proposed changes do not affect the source term, containment isolation, or radiological release assumptions used in evaluating the radiological consequences of an accident previously evaluated. Further, the proposed changes do not increase the types and amounts of radioactive effluent that may be released offsite, nor significantly increase individual or cumulative occupational or public radiation exposure.

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

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

Response: No.

The proposed changes neither install or remove any plant equipment, nor alter the design, physical configuration, or mode of operation of any plant structure, system, or component.

The proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated in the Updated FSAR. No new accident scenarios, failure mechanisms, or limiting single failures are introduced as a result of the proposed changes.

Specifically, no new hardware is being added to the plant as part of the proposed changes, no U.S. Nuclear Regulatory Commission Page 5 of 5 Serial: RA-22-0081

existing equipment design or function is being modified, and no significant changes in operations are being introduced. No new equipment performance burdens are imposed.

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 amendment involve a significant reduction in a margin of safety?

Response: No.

The proposed changes in this license amendment request do not alter the design, configuration, operation, or function of any plant system, structure, or component. The ability of any operable structure, system, or component to perform its designated safety function is unaffected by these changes. The proposed changes do not alter the manner in which safety limits, limiting safety system settings, or limiting conditions for operation are determined. They do not alter any safety analysis assumptions, initial conditions, or results of any accident analyses. The proposed changes will not result in plant operation in a configuration outside the design basis.

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

Based upon the above evaluation, Duke Energy concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c) and, accordingly, a finding of "no significant hazards consideration" is justified.

4.3 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 Commission's 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

S

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 changes do 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 offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure.

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

U.S. Nuclear Regulatory Commission Serial: RA-22-0081

ENCLOSURE 2

PROPOSED TECHNICAL SPECIFICATION CHANGES (MARK-UP)

SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1

DOCKET NO. 50-400

RENEWED LICENSE NUMBER NPF-63

2 PAGES PLUS THE COVER

TABLE 3.3-4 (Continued)

TABLE NOTATIONS

• Time constants utilized in the lead-lag controller for Steam Line Pressure--Low are 50 seconds and Tz s 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.

1. 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 NGGG 015a, "Engineering Instrument Setpoints.'1 __ -.-~ the FSAR.

The as-found and as-left tolerances are specified in the Technical Requirements Manual.

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

U.S. Nuclear Regulatory Commission Serial: RA-22-0081

ENCLOSURE 3

PROPOSED OPERATING LICENSE CHANGES (MARK-UP)

SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1

DOCKET NO. 50-400

RENEWED LICENSE NUMBER NPF-63

5 PAGES PLUS THE COVER L. This license is effective as of the date of issuance and shall expire at midnight on October 24, 2046.

FOR THE NUCLEAR REGULATORY COMMISSION

/RA/

Eric J. Leeds, Director Office of Nuclear Reactor Regulation

Attachments/Appendices:

1. Attashment 1 TOI Diesel Engine Requirements 2. Appendix A-Technical Specifications

3. Appendix B - Environmental Protection Plan Renumber items 1-4.

4. Appendix C - Antitrust Conditions

5. Appendix D - Additional Conditions

Date of Issuance: December 17, 2008

Renewed License No. NPF-63 Amendment No. 174 llCENSE AUTHORITY. FILE COPl. DO NOT REMOVl January 12, 1987

1. Changes to the maintenance and surveillance proarams enqines, s identified in Shearon Harris ~SF.R No. 4, the provis ns of 10 C~R SO.SQ.

The frequencv f the ma.for enaine overhauls referred to *n the 7 icense conditions belo shall be consistent with Section JV.I "Overhaul Frequencv" in Re *sion? o~ Apoendix II o~ the Desion eview/0ua1itv

~evalidation repor which was transmitted hv letter ated May 1, JqR~;

from J. George, Own s r-rouo, to H. nenton, NPC.

~. Connectino rod assembl s shall be subiected followinq insoections at each major engine ove haul:

a. The ~urfaces of the ra fretting. If fretting eth shal insoected for siqns of shall be subject to an engineering evaluation corrective action.

b, All connecting-rod bolts sha 1 e lubricated in accordance with the engine manufacturer's ins uctions ~rid torqued to the specifi-above the crankpin shall b meas ed ultrasonicallv pre-and cations of the manufacturer. e lenqths of the two pairs of bolts post-tensioning.

c. The lengths of the tw pairs of bolts bove the crankpin shall he measured ultrasonica v prior to deten *onina and disassemblv of the bolts. If bolt te ion 1s less than q3 ~ the value at installation, the cause shall b determined, aoprocriat corrective action shall be re-evaluate. be taken, and t interval between checks o bolt tension s~all

d. All connec *ncr-rod damage (e., oallinn', and the two oairs o~ conn cting rod bolts bolts shall be visually insoe ted for thread ahove t crankpin shall be inspected by magnetic rticle testin. (MT) to verify the continued absence of crac inq. All wash s used with the bolts shall be examined visua11 for signs of lling or crackina, and repfaced if damaged.

e. sual inspection shall be performed of all external surfa s of the link rod box to verifv the absence of anv sians of serv, e induced distress. * ~

All of the bolt holes in the link rod box shall be inspected for thread damage 'e.g., gallina' or other signs o~ abnormalities.

In addition, the bolt holes subject to the highest stresses (i.e.,

- ? -

the pair immediately above the cran~pin' shall be examined with an appropriate nondestructive method to verify the continued absence of crackino. Anv indications shall be recorded ~or engineering evaluat~on an~ appropriate corrective action.

3. vlinder hloc~s shall be suhiected to the followina i terval specified in the inspections: at

a. inspecterl for 11 1 crac and "stud-to-end" cracks as defined - in a repo ~

• bv Failur~ iaament 11 crar s, 11 stud-to-stud" Ana1vs*s Associates, tnc. 1 FaA4) entitled, "Design eview of TDl R-4 and V-4 Series Emeraencv niesel Generator Cv rider r:nocks" (J:aAA re rt no. FaAA-84-Q-1.1. ll, dated December QA4. (Note that the FaAA r port specifies additional inspectio to be oerformed for blocks with "known" or 11 assumed 11 liaament era sl. The inspection using the cumu ative damage index model i the subiect FaAA report. intervals Ii.., frenuency) shall not excee the intervals calculated In addition, in ection method shall be nsistent with or equivalent to those identif, din the subiect J:aAA reoort.

b. In addition to insp tions specifie the aforementioned FaAA reoort, bloc~s with I nown 11 or "as med ligament cracks 11 fas defined

• n the J

  aAA report) sh 11 be ins cted at each refueling outage exposed by the removal o. two r more cylinder heads. This process to determine whether or ot era shave initiated on the top sur~ace shall be repeated over sev refuelinq outages until the entire block top has been inspect Liquid-oenetrant testing or a similarly sensitive nonde tru tive testing technique shall be used to detect cracking. and ddy c rent shall he used as appropriate to determine the dept of anv er cks discovered.

c. Jf inspection reve s cracks in the cvlinder blocks between stud ho 1 es of ad.iacent cylinders ( 11 stud-t -stud 11 cracks' or "stud-to-end" and the affect a engine shall he consi red inooerable. The enaine cracks, this co ition shall be repor d orornptlv to the NQC staff shall not be estored to "operable" stats until the proposed

• disposition and/or corrective actions hav been aoproved hy the NRC staff

4. The followi.* air roll test shall be oerfonned as when the Q ant is a1readv in an Action Statement ecified below, exceot echnical Specification 3/4.8.1, Electric Power Svstems, A.C. Sources":

Own rs ~roup, by letter dated December 11, 19R4.,s eoort was transmitted to H. nenton, ~RC, from The enqines shall be rolled over with the airstart system anrl with the cylinder stopcocks open prior to each planned start, unless that start occurs within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of a shutdown. The engines shall also he rolled ver with the airstart svstem and with the cvlinder stopcocks open a er hours, but no more than 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after engine shutdown and then r led ov once again approximately ?4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after each shutdown. (Tn e event over ocedure prior to expiration of the R-hour or ?4-hour pe iods an e gine is removed from service for anv reason other than the o11ino noted a ove, that enqine need not be relied over while it is ut of service. The licensees shall air roll the enaine over with the stoococks ooen at tH time it is returned to service'. The oriqin anv water detected in t he cylinder must be determined and anv cili der h~ad which leaks due to crack shall be replaced. The ahove air oll test mav he discontinued 1lowino the first refuelino outaae suh ect to the following condi 'ons: *

a. All cylinder 1980). are Group III heads after September

b. Quality revalidati insoections, as i fltified in the Design Review/Quality Reva "dation reoort, ve been completed for all cylinder heads.

c. r,roup III heads contin strate iea~ - free per+onnance. This shall he confime prior to deletinq air roll tests.

5. the fo 11 owi nq:

a. The turbocharoer thrus ll be visuallv insoected for excessive wear after. starts since* the orevious visual inspection.

b. Turbocharger rotor axial clearance sh ]1 be measured at each refueling outaae o verify compliance "th TOI/Elliott specifications. In aqdition, thrust bea inq measurements shall b~ compared with easurements taken previous to determine a need for further ins ~ ction or corrective action.

c. ~pectroo r. phic and ferrographic engine oil ana vsis shall be oerfonn ~ quarterly to provide early evidence o bearing and articulate si~e which could sianify thrust be ino deqradation. degra tion. Particular attention shall be paid copper level

6. Prior o restart following the first refueling~

The engine base shall be inspected for degenerate micros ucture '1*!idmanstaetten qraphite1 and the results submitted to Vie N~C for evaluation.

b. The exhaust manifold caos~rew tor~ues fwithout be checked/corrected for hoth engines.

- 4 -

c. A visual inspection, liquid penetrant test, and dimensional check o-F diesel aenerator lA governor shaft shall be performed.

A liquid penetrant test of diesel generator lA governor drive aear and shaft shall be performed to check for fatigue checks

e. Install an acceotable jacket water standoipe level n both diesel generators.

f. esser coupling shall be added on to the engine dr' en lube mp suction line to mitigate the thennal exoan on loadina st esses on the pump,inlet nozzle.

a. Replace t e ?.1/2 inch Oresser coupling located be een the turbocharg and lube oil sumo tank for both ain lines with a?1/2 in h 150 lb. S.O. flange with A307 olts.

h. The four starti a air manifold flarae hor suoport modifications specified in the R/0~ shall be imoieme ed.

i. The ~acket water pi _ shall be reinforced a and fittina (1 qe bore) suoport mernhers in the l)P/()R,

J. The two-directional rest ai~ts a each fuel oil drio header

(?. per engine) shall be m ifi. to a thr~e-directional restraint in order to provide axial r raint of the header and to minimize the effects on all associat tubing.

k. An anchor (six-way s added on the fuel-oil-. to-day-tank return pipi g ftwo lies per enaine' in order to reduce the unsupoorte span length nd to minimize the effects of the off engine o ing.

l. On the

1) i five ad : ustment potentiometers on the inted circuit de of the ad.iustment for each of the board f the voltaqe regulator with glyp ol lacquer. If a iustments to the potentiometer are n ded, procedures sh l specify that the glyptol lacquer shal be removed d then reapplied after the ad.iustments hav been oerfonned.

The lua arranaement for the heatsink connection and the oower ~ircuit. reactor shall be modified so that t ere ~re no more than two lugs on each bolt.

1) For the bridqe rectifier assemblv, the diodes shall b mounted on the heatsinks with drilled holes, nuts, and lockwashers and tightened to the prooer torque.

U.S. Nuclear Regulatory Commission Serial: RA-22-0081

ENCLOSURE 4

PROPOSED TECHNICAL SPECIFICATION BASES CHANGES (MARK-UP)

(FOR INFORMATION ONLY)

SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1

DOCKET NO. 50-400

RENEWED LICENSE NUMBER NPF-63

2 PAGES PLUS THE COVER

AD-EG-ALL-1153, "Engineering LIMITING SAFETY SYSTEM SETTINGS Instrument Setpoint/Uncertainty Calculations,"

BASES,..___ ____ ___.7

REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS (Continued)

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 t Setpoint or more conservative actual field setpoint are specified in EGR-NGGC-0153 ------

(superseded by i\\D EG ALL 1153), " 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 the Technical Requirements Manual, 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 (superseded by AD-EG-ALL-1153).

HNP uses the plant-specific term Nominal Trip Setpoint (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 8.

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

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

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

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

SHEARON HARRIS - UNIT 1 B 2-3b Amendment No. 189 AD-EG-ALL-1153, "Engineering 3/4.3 INSTRUMENTATION Instrument Setpoint/Uncertainty Calculations,"

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 the Technical Requirements Manual 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 OPE RABLE 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 (superseded by /\\D EG /\\LL 1153), " 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 the Technical Requirements Manual, 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 (superseded by AD-EG-ALL-1153).

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

SHEARON HARRIS - UNIT 1 B 3/4 3-2a Amendment No. 89