RA-21-0112, License Amendment Request Regarding Administrative Change to Reflect Development of a Technical Requirements Manual

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License Amendment Request Regarding Administrative Change to Reflect Development of a Technical Requirements Manual
ML21158A131
Person / Time
Site: Harris Duke Energy icon.png
Issue date: 06/07/2021
From: Maza K
Duke Energy Progress
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
RA-21-0112
Download: ML21158A131 (32)
v • d • e


Text

Kim E. Maza Site Vice President Harris Nuclear Plant 5413 Shearon Harris Rd New Hill, NC 27562-9300 984-229-2512 10 CFR 50.90 June 7, 2021 Serial: RA-21-0112 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 Administrative Change to Reflect Development of a Technical Requirements Manual 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), Technical Specifications (TS). The proposed amendment will revise HNP TS to reflect the transition of the licensee-controlled plant procedure PLP-106, Technical Specification Equipment List Program, to a licensee-controlled Technical Requirements Manual (TRM). The proposed change is an administrative change to the HNP TS with no impact on technical content.

The Enclosure provides a description and assessment of the proposed change. Attachments 1 and 2 provide copies of the proposed changes to the TS and TS Bases, respectively. is provided for information only, as changes to the HNP TS Bases will be processed in accordance with HNP TS 6.8.4.n, Technical Specifications (TS) Bases Control Program, upon implementation of the amendment.

The proposed change has 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 change involves no significant hazards consideration. Additionally, this letter contains no regulatory commitments.

Approval of the proposed license amendment is requested within twelve months of acceptance for review by the NRC staff. The amendment shall be implemented within 90 days from approval.

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

Please refer any questions regarding this submittal to Art Zaremba, Manager - Nuclear Fleet Licensing, at (980) 373-2062.

U.S. Nuclear Regulatory Commission Page 2 of 2 Serial: RA-21-0112 I declare under penalty of perjury that the foregoing is true and correct.

Executed on June 7, 2021.

Sincerely, Kim E. Maza Site Vice President Harris Nuclear Plant

Enclosure:

Description and Assessment of the Proposed Change : Proposed Technical Specification Changes (Mark-up) : Proposed Technical Specification Bases Changes (Mark-up) cc: J. Zeiler, NRC Sr. Resident Inspector, HNP W. L. Cox, Ill, Section Chief, N.C. DHSR M. Mahoney, NRC Project Manager, HNP L. Dudes, NRC Regional Administrator, Region II

U.S. Nuclear Regulatory Commission Serial: RA-21-0112 Enclosure ENCLOSURE DESCRIPTION AND ASSESSMENT OF THE PROPOSED CHANGE SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 DOCKET NO. 50-400 RENEWED LICENSE NUMBER NPF-63 4 PAGES PLUS THE COVER

U.S. Nuclear Regulatory Commission Page 1 of 4 Serial: RA-21-0112 Enclosure Description and Assessment of the Proposed Change License Amendment Request Regarding Administrative Change to Reflect Development of a Technical Requirements Manual 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), Technical Specifications (TS). The proposed amendment will revise HNP TS to reflect the transition of the licensee-controlled plant procedure PLP-106, Technical Specification Equipment List Program, to a licensee-controlled Technical Requirements Manual (TRM). The proposed change is an administrative change to the HNP TS with no impact on technical content.

2.0 DETAILED DESCRIPTION 2.1 Current Technical Specification The following HNP TS refer to plant procedure PLP-106 for content that was previously relocated from the TS:

  • TS Table 2.2-1, Reactor Trip System Instrumentation Trip Setpoints, Note 8
  • TS Table 3.3-2, Reactor Trip System Instrumentation Response Times
  • TS 3/4.3.2, Engineered Safety Features Actuation System Instrumentation
  • TS Table 3.3-4, Engineered Safety Features Actuation System Instrumentation Trip Setpoints, Note 2
  • TS Table 3.3-5, Engineered Safety Features Response Times
  • TS 3/4.4.9, Pressure/Temperature Limits
  • TS Table 4.4-5, Reactor Vessel Material Surveillance Program
  • TS Figure 3.4-4, Maximum Allowed PORV Setpoint for the Low Temperature Overpressure Protection System
  • TS 3/4.6.3, Containment Isolation Valves
  • TS Table 3.6-1, Containment Isolation Valves
  • TS Figure 4.7-1, Sample Plan (2) for Snubber Functional Test
  • TS 3/4.8.4, Electrical Equipment Protective Devices 2.2 Reason for the Proposed Change The HNP TS are based upon the format and content of the NUREG-0452, "Standard Technical Specifications for Westinghouse Pressurized Water Reactors," series. As a result, the HNP TS numbers and associated Bases numbers differ from those contained in NUREG-1431,

U.S. Nuclear Regulatory Commission Page 2 of 4 Serial: RA-21-0112 Enclosure Standard Technical Specifications - Westinghouse Plants (Revision 4, ADAMS Accession No. ML12100A222).

HNP has not converted to the Improved Standard TS structure and has continued to maintain a licensee-controlled document in the form of plant procedure PLP-106, titled Technical Specification Equipment List Program, to capture relocated TS content approved by the NRC, including equipment lists, figures, and surveillance programs. PLP-106 is a document incorporated by reference into the HNP Final Safety Analysis Report (FSAR) and is subject to the update and reporting requirements of 10 CFR 50.71(e) and change controls of 10 CFR 50.59.

To standardize the process of updating the licensee-controlled document with that of the rest of the Duke Energy fleet, PLP-106 is being converted into a TRM. As a TRM, the content previously contained in PLP-106 will continue to be incorporated by reference in the FSAR and subject to the update and reporting requirements of 10 CFR 50.71(e), with changes processed in accordance with 10 CFR 50.59, without the required additional resources associated with processing a change to a plant procedure.

2.3 Description of the Proposed Change The proposed administrative change will replace all TS references to plant procedure PLP-106, as listed in Section 2.1 above, with references to the TRM.

3.0 TECHNICAL EVALUATION

The proposed change to reference a TRM in place of plant procedure PLP-106 does not adversely alter the current TS or introduce any new TS requirements. The relocated TS content currently captured in PLP-106 will be maintained in the TRM, with changes to the licensee-controlled content processed in accordance with 10 CFR 50.59. When implemented, this change will provide continuity and consistency throughout the Duke Energy fleet in the processing of changes to relocated TS content in licensee-controlled documents. There is no impact on plant operations or systems as a result of the proposed change.

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 10 CFR 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 limiting control settings, (2) limiting conditions for operation, (3) surveillance requirements, (4) design features, and (5) administrative controls.

Conclusion Duke Energy has evaluated the proposed change against the applicable regulatory requirements described above. Based on this evaluation, there is reasonable assurance

U.S. Nuclear Regulatory Commission Page 3 of 4 Serial: RA-21-0112 Enclosure that the health and safety of the public will remain unaffected following the approval of the proposed change.

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), Technical Specifications (TS). The proposed amendment will revise HNP TS to reflect the transition of the licensee-controlled plant procedure PLP-106, Technical Specification Equipment List Program, to a licensee-controlled Technical Requirements Manual (TRM). The proposed change is an administrative change to the HNP TS with no impact on technical content.

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

Response: No.

The proposed change to reflect a TRM in place of the currently referenced plant procedure PLP-106 is administrative in nature and does not change the technical content of the TS. The proposed change does not adversely affect accident initiators or precursors nor alter the design assumptions, conditions or configurations of the facility.

The proposed change does not alter or prevent the capability of structures, systems and components (SSCs) to perform their intended function to mitigate the consequences of any initiating events within the assumed acceptance limits.

Therefore, the proposed change does 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 accident previously evaluated?

Response: No.

The proposed change to reflect a TRM in place of the currently referenced plant procedure PLP-106 is administrative in nature and does not change the technical content of the TS. The proposed change does not alter the design requirements of any SSC or its function during accident conditions. The proposed change does not involve a physical alteration to the plant or any changes in methods governing normal plant operation. The proposed change does not alter any assumptions made in the safety analysis.

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

U.S. Nuclear Regulatory Commission Page 4 of 4 Serial: RA-21-0112 Enclosure (3) Does the proposed change involve a significant reduction in a margin of safety?

Response: No.

The proposed change to reflect a TRM in place of the currently referenced plant procedure PLP-106 is administrative in nature and does not change the technical content of the TS. The proposed change does not alter the way safety limits, limiting safety system settings or limiting conditions for operation are determined. The safety analysis acceptance criteria are not affected by the proposed change. The proposed change will not result in plant operation in a configuration outside the design basis and does not adversely affect systems that respond to safely shutdown the plant and maintain the plant in a safety shutdown condition.

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

Based upon the above evaluation, Duke Energy concludes that the proposed change 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 The proposed amendment is for administrative, non-technical changes only and would not change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20. 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 offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure.

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

U.S. Nuclear Regulatory Commission Serial: RA-21-0112 ATTACHMENT 1 PROPOSED TECHNICAL SPECIFICATION CHANGES (MARK-UP)

SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 DOCKET NO. 50-400 RENEWED LICENSE NUMBER NPF-63 17 PAGES PLUS THE COVER

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3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.1 As a minimum, the Reactor Trip System instrumentation channels and interlocks of Table 3.3-1 shall be OPERABLE.

APPLICABILITY: As shown in Table 3.3-1.

ACTION: As shown in Table 3.3-1.

SURVEILLANCE REQUIREMENTS 4.3.1.1 Each Reactor Trip System instrumentation channels and interlock and the automatic trip logic shall be demonstrated OPERABLE by the performance of the Reactor Trip System Instrumentation Surveillance Requirements specified in Table 4.3-1.

4.3.1.2 The REACTOR TRIP SYSTEM RESPONSE TIME of each Reactor trip function shall be verified to be within its limit, specified in the Technical Specification Equipment List Program, plant procedure PLP-106, at the frequency specified in the Surveillance Frequency Control Program.

Technical Requirements Manual SHEARON HARRIS - UNIT 1 3/4 3-1 Amendment No. 154

Technical Requirements Manual INSTRUMENTATION ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS 4.3.2.1 Each ESFAS instrumentation channel and interlock and the automatic actuation logic and relays shall be demonstrated OPERABLE by performance of the ESFAS Instrumentation Surveillance Requirements specified in Table 4.3-2.

4.3.2.2 The ENGINEERED SAFETY FEATURES RESPONSE TIME of each ESFAS function shall be verified to be within its limit specified in the Technical Specification Equipment List Program, plant procedure PLP-106, at the frequency specified in the Surveillance Frequency Control Program.

Technical Requirements Manual SHEARON HARRIS - UNIT 1 3/4 3-17 Amendment No. 154

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.

the Technical Requirements Manual SHEARON HARRIS - UNIT 1 3/4 3-36 Amendment No. 161

Technical Requirements Manual REACTOR COOLANT SYSTEM 3/4.4.9 PRESSURE/TEMPERATURE LIMITS LIMITING CONDITION FOR OPERATION 3.4.9.2 The Reactor Coolant System (except the pressurizer) temperature and pressure shall be limited in accordance with the limit lines shown on Figures 3.4-2 and 3.4-3 during heatup, cooldown, and inservice leak and hydrostatic testing with:

a. A maximum heatup rate as shown on Table 4.4-6.
b. A maximum cooldown rate as shown on Table 4.4-6.
c. A maximum temperature change of less than or equal to 10°F in any 1-hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves.

APPLICABILITY: MODES 4, 5, and 6 with reactor vessel head on.

ACTION:

With any of the pressure limits exceeded, restore the temperature and/or pressure to within the limit within 30 minutes; if the pressure and temperature limit lines shown on Figure 3.4-2 and 3.4-3 were exceeded, perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the Reactor Coolant System; determine that the Reactor Coolant System remains acceptable for continued operation or maintain the RCS Tavg and pressure at less than 200°F and 500 psig, respectively.

SURVEILLANCE REQUIREMENTS 4.4.9.2.1 The Reactor Coolant System temperature and pressure shall be determined to be within the limits at the frequency specified in the Surveillance Frequency Control Program during system heatup, cooldown, and inservice leak and hydrostatic testing operations.

4.4.9.2.2 Deleted from Technical Specifications. Refer to the Technical Specification Equipment List Program, plant procedure PLP-106.

Technical Requirements Manual SHEARON HARRIS - UNIT 1 3/4 4-34 Amendment No. 154

Technical Requirements Manual FIGURE 3.4-2 REACTOR COOLANT SYSTEM COOLDOWN LIMITATIONS - APPLICABLE UP TO 55 EFPY SHEARON HARRIS - UNIT 1 3/4 4-35 Amendment No.

Technical Requirements Manual FIGURE 3.4-3 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS - APPLICABLE UP TO 55 EFPY SHEARON HARRIS - UNIT 1 3/4 4-36 Amendment No.

Technical Requirements Manual 500 LOW PORV HIGH PORV PORV SETPOINT (PSIG) 400 300 0 100 200 300 400 MEASURED RCS TEMPERATURE (°F)

Technical Requirements Manual RCS TEMP (°F) LOW PORV* (psig) HIGH PORV* (psig) 90 400 410 250 400 410 325 440 450

  • VALUES BASED ON 55 EFPY REACTOR VESSEL DATA INSTRUMENT ERRORS ARE CONTROLLED BY THE TECHNICAL SPECIFICATION EQUIPMENT LIST PROGRAM, PLANT PROCEDURE PLP-106.

FIGURE 3.4-4 MAXIMUM ALLOWED PORV SETPOINT FOR THE LOW TEMPERATURE OVERPRESSURE PROTECTION SYSTEM SHEARON HARRIS - UNIT 1 3/4 4-41 Amendment No.

CONTAINMENT SYSTEMS 3/4.6.3 CONTAINMENT ISOLATION VALVES Technical Requirements Manual LIMITING CONDITION FOR OPERATION 3.6.3 Each containment isolation valve specified in the Technical Specification Equipment List Program, plant procedure PLP-106, shall be OPERABLE with isolation times less than or equal to required isolation times.

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

ACTION:

With one or more of the containment isolation valve(s) inoperable, maintain at least one isolation valve OPERABLE in each affected penetration that is open and:

a. Restore the inoperable valve(s) to OPERABLE status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or in accordance with the Risk-Informed Completion Time Program, or
b. Isolate each affected penetration within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or in accordance with the Risk-Informed Completion Time Program by use of at least one deactivated automatic valve secured in the isolation position, or
c. Isolate each affected penetration within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or in accordance with the Risk-Informed Completion Time Program by use of at least one closed manual valve or blind flange, or
d. Be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREMENTS 4.6.3.1 Each isolation valve shall be demonstrated OPERABLE prior to returning the valve to service after maintenance, repair or replacement work is performed on the valve or its associated actuator, control or power circuit by performance of a cycling test, and verification of isolation time.

SHEARON HARRIS - UNIT 1 3/4 6-14 Amendment No. 

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Technical Requirements Manual Technical Requirements Manual Technical Requirements Manual ELECTRICAL POWER SYSTEMS 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES Specifications 3/4.8.4.1 and 3/4.8.4.2 have been deleted from Technical Specifications and relocated to plant procedure PLP-106.

the Technical Requirements Manual.

PAGES 3/4 8-20 THROUGH 3/4 8-43 HAVE BEEN DELETED.

Pages 3/4 8-20, 3/4 8-21, 3/4 8-39, and 3/4 8-40 by Amendment No. .

Pages 3/4 8-22 through 3/4 8-38B and 3/4 8-41 through 3/4 8-43 by Amendment No. 13.

SHEARON HARRIS - UNIT 1 3/4 8-19 Amendment No

U.S. Nuclear Regulatory Commission Serial: RA-21-0112 ATTACHMENT 2 PROPOSED TECHNICAL SPECIFICATION BASES CHANGES (MARK-UP)

SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 DOCKET NO. 50-400 RENEWED LICENSE NUMBER NPF-63 6 PAGES PLUS THE COVER

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3/4.3 INSTRUMENTATION the Technical Requirements Manual 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 Technical Requirements Manual 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, 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. 161

REACTOR COOLANT SYSTEM BASES PRESSURE/TEMPERATURE LIMITS (Continued) heatup and the time (or coolant temperature) along the heatup ramp. Furthermore, since the thermal stresses at the outside are tensile and increase with increasing heatup rate, a lower bound curve cannot be defined. Rather, each heatup rate of interest must be analyzed on an individual basis.

Following the generation of pressure-temperature curves for both the steady-state and finite heatup rate situations, the final limit curves are produced as follows. A composite curve is constructed based on a point-by-point comparison of the steady-state and finite heatup rate data.

At any given temperature, the allowable pressure is taken to be the lesser of the three values taken from the curves under consideration.

The use of the composite curve is necessary to set conservative heatup limitations because it is possible for conditions to exist such that over the course of the heatup ramp the controlling condition switches from the inside to the outside and the pressure limit must at all times be based on analysis of the most critical criterion.

The composite curves for the heatup rate data and the cooldown rate data in Figures 3.4-2 and 3.4-3 have not been adjusted for possible errors in the pressure and temperature sensing instruments. However, the heatup and cooldown curves in plant operating procedures have been adjusted for these instrument errors. The instrument errors are controlled by the Technical Specification Equipment List Program, Plant Procedure PLP-106.the Technical Requirements Manual.

"ISLH" pressure-temperature (P-T) curves may be used for inservice leak and hydrostatic tests with fuel in the reactor vessel. However, ISLH tests required by the ASME code must be completed before the core is critical.

Although the pressurizer operates in temperature ranges above those for which there is reason for concern of nonductile failure, operating limits are provided to assure compatibility of operation with the fatigue analysis performed in accordance with the ASME Code requirements.

LOW TEMPERATURE OVERPRESSURE PROTECTION A Note prohibits application of LCO 3.0.4.b to an inoperable overpressure protection system.

There is an increased risk associated with entering MODE 4 from MODE 5 with the overpressure protection system inoperable and the provisions of LCO 3.0.4.b, which allow entry into a MODE or other specified condition in the Applicability with the LCO not met after performance of a risk assessment addressing inoperable systems and components, cannot be applied to this circumstance.

The OPERABILITY of two PORVs or an RCS vent opening of at least 2.9 square inches ensures that the RCS will be protected from pressure transients which could exceed the limits of Appendix G to 10 CFR Part 50 when one or more of the RCS cold legs are less than or equal to 325°F. Either PORV has adequate relieving capability to protect the RCS from overpressurization when the transient is limited to either: (1) the start of an idle RCP with the secondary water temperature of the steam generator less than 50°F above the RCS cold leg temperatures, or (2) the start of a charging/safety injection pump and its injection into a water-solid RCS.

The maximum allowed PORV setpoint for the Low Temperature Overpressure Protection System (LTOPS) is derived by analysis which models the performance SHEARON HARRIS - UNIT 1 B 3/4 4-14 Amendment No. 179

the Technical Requirements Manual.

3/4.7 PLANT SYSTEMS BASES 3/4.7.8 SNUBBERS All snubbers are required OPERABLE to ensure that the structural integrity of the Reactor Coolant System and all other safety-related systems is maintained during and following a seismic or other event initiating dynamic loads.

Surveillance to demonstrate OPERABILITY is by performance of an augmented inservice inspection program specified in the Technical Specification Equipment List Program, plant procedure PLP-106. The program is in accordance with the ASME OM Code as required by 10 CFR 50.55a.

the Technical Requirements Manual.

Revision 1 SHEARON HARRIS - UNIT 1 B 3/4 7-4 Amendment No. 102

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