View source for HNP-07-005, Request for License Amendment to Relocate and Revise Technical Specifications Surveillance Requirement 4.0.5

{{Adams
| number = ML071800018
| issue date = 06/15/2007
| title = Request for License Amendment to Relocate and Revise Technical Specifications Surveillance Requirement 4.0.5
| author name = Duncan R
| author affiliation = Progress Energy Carolinas, Inc
| addressee name = 
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000400
| license number = NPF-063
| contact person = 
| case reference number = HNP-07-005
| document type = Letter type:HNP, License-Application for Facility Operating License (Amend/Renewal) DKT 50, Technical Specifications
| page count = 51
| project = 
| stage = Request
}}

=Text=
{{#Wiki_filter:&?  Progress Energy                                                  ViceRer JPresidentD~cnl Robert J. Duncan, 11 Harris Nuclear Plant Progress Energy Carolinas, Inc.
JUN 1 5 2007                                     Serial: HNP-07-005 10 CFR 50.90 U.S. Nuclear Regulatory Commission ATTENTION: Document Control Desk Washington, DC 20555 Shearon Harris Nuclear Power Plant, Unit No. 1 Docket No. 50-400/License No. NPF-63 Request for License Amendment to Relocate and Revise Technical Specifications Surveillance Requirement 4.0.5 Ladies and Gentlemen:
In accordance with the Code of Federal Regulations, Title 10, Part 50.90, Carolina Power and Light Company (CP&L) doing business as Progress Energy Carolinas, Inc., requests a license amendment for the Harris Nuclear Plant (HNP) Technical Specifications (TS). The proposed amendment is administrative in nature, and it will relocate the Inservice Testing (IST) requirements to the administrative section of TS, remove the Inservice Inspection (ISI) requirements, and establish a TS Bases Control Program. This amendment supports the third ten-year intervals of the ISI and IST Programs.
Attachment I provides the description, proposed change, background, and technical analysis for the proposed amendment.
Attachment 2 details, in accordance with 10 CFR 50.91 (a), the basis for HNP's determination that the proposed change does not involve a significant hazards consideration.
Attachment 3 provides the proposed TS changes.
Attachment 4 provides the revised TS changes.
Attachment 5 provides the proposed TS Bases changes (for information only).
P.O.
Box 165 New Hill, NC 27562 T> 919.362.2502 F> 919.362.2095

HNP-07-005 Page 2 With respect to this proposed amendment, there is no/significant change in the types or significant increase in the amounts of any effluents/that may be released offsite and there is no significant increase in individual or cumulative occupational radiation exposure. The proposed change to the Technical Specifications meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental assessment or environmental impact statement is required for approval of this application.
In accordance with 10 CFR 50.91(b), HNP is providing the State of North Carolina with a copy of the proposed license amendment.
HNP requests approval of the proposed amendment by May of 2008, with the amendment being implemented within 60 days.
This document contains no new Regulatory Commitment.
Please refer any question regarding this submittal to Mr. Dave Corlett at (919) 362-3137.
I declare, under penalty of perjury, that the attached information is true and correct (Executed on      JUN 15 2007       ).
Si   erely, R. J. Duncan II Vice President Harris Nuclear Plant RJD/jpy Attachments:
: 1. Description, Proposed Change, Background, and Technical Analysis
: 2. 10 CFR 50.92 No Significant Hazards Evaluation
: 3. Proposed Technical Specifications (TS) Changes
: 4. Revised Technical Specifications (TS) Pages
: 5. Proposed Technical Specifications (TS) Bases Changes (For Information Only) c:
Mr. P. B. O'Bryan, NRC Senior Resident Inspector Ms. B. 0. Hall, N.C. DENR Section Chief Ms. L. M. Regner, NRC Project Manager Dr. W. D. Travers, NRC Regional Administrator

Attach ent 1 to SERIAL: HNP-07-005 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT NO. 1 DOCKET NO. 50-400/LICENSE NO. NPF-63 REQUEST FOR LICENSE AMENDMENT DESCRIPTION, PROPOSED CHANGE, BACKGROUND, AND TECHNICAL ANALYSIS DESCRIPTION, PROPOSED CHANGE, BACKGROUND, AND TECHNICAL ANALYSIS Page Al-1 of 8

Attachnent 1 to SERIAL: HNP-07-005 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT NO. 1 DOCKET NO. 50-400/LICENSE NO. NPF-63 REQUEST FOR LICENSE AMENDMENT DESCRIPTION, PROPOSED CHANGE, BACKGROUND, AND TECHNICAL ANALYSIS Description In accordance with the Code of Federal Regulations, Title 10; Part 50.90, "Application for amendment of license or construction permit," Carolina Power and Light Company (CP&L) doing business as Progress Energy Carolinas, Inc., requests a license amendment for the Harris Nuclear Plant (HNP) Technical Specifications (TS). The proposed amendment will: 1) relocate the Inservice Testing (IST)
Program of TS Surveillance Requirement (SR) 4.0.5 to the administrative section of the TS under TS 6.8.4 and revise the wording to be consistent with NUREG-1 431, Revision 3, Standard Technical Specifications Westinghouse Plants, Specifications, published in June 2004; 2) remove the Inservice Inspection (ISI) activities currently contained in TS SR 4.0.5 and move them to an owner-controlled program consistent with NUREG-1431 since these requirements are redundant with the regulations under 10 CFR 50.55a; and 3) establish a TS Bases Control Program consistent with NUREG-1 431. To incorporate these changes, several other sections of TSthat reference TS SR 4.0.5 will be revised to reference the IST Program. The proposed changes are administrative in nature, and they do not eliminate any tests (IST) or inspections (ISI). This proposed amendment will resolve the ASME Code inconsistency between the IST Program and the TS as required by 10 CFR 50.55a(f)(5)(ii), and will support implementation of the third ten-year intervals of the ISI and IST Programs. These administrative type changes to the TS have no adverse impact on the public health and safety.
Proposed Change The following summarizes the specific changes proposed in this amendment:
Inservice Inspection and Testing Progirams TS SR 4.0.5 will be relocated to new TS 6.8.4.1, "Inservice Testing Program," as an IST Program consistent with NUREG-1 431, Section 5.5.8, "Inservice Testing Program," and it will be revised to include the appropriate ASME Code reference. In addition, The TS Bases associated with TS SR 4.0.5 will be deleted.
TS SR 4.0.5.a will be deleted.
TS SR 4.0.5.b will be moved to new TS 6.8.4.1.1 and will be revised to be consistent with NUREG-1 431, Section 5.5.8.a (e.g., the proposed TS will not refer to ISI activities as currently contained in TS SR 4.0.5.b).
Page A1-2 of 8

I I/
Attachment 1 to SERIAL: HNP-07-005 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT NO. 1 DOCKET NO. 50-400/LICENSE NO. NPF-63 REQUEST FOR LICENSE AMENDMENT DESCRIPTION, PROPOSED CHANGE, BACKGROUND, AND TECHNICAL ANALYSIS Proposed Change (continued)
TS SR 4.0.5.c will be moved to new TS 6.8.4.1.2 and will be revised to be consistent with NUREG-1431, Section 5.5.8.b (e.g., the proposed TS will not refer to ISI activities as currently contained in TS SR 4.0.5.c).
TS SR 4.0.5.d will be deleted.
New TS 6.8.4.1.3 will be added consistent with NUREG-1 431, Section 5.5.8.c.
TS SR 4.0.5.e will be moved to new TS 6.8.4.1.4. The existing wording is consistent with NUREG-1431, Section 5.5.8.d, and it will be revised to include the appropriate ASME Code reference.
The following TS SRs will be revised to reflect the relocation of TS SR 4.0.5. by replacing "Specification 4.0.5" with "the Inservice Testing Program."
SR   4.1.2.3.1   Charging Pump - Shutdown SR   4.1.2.4     Charging Pump - Operating SR   4.4.2.1     Pressurizer Code Safety Valves - Shutdown SR   4.4.2.2     Pressurizer Code Safety Valves - Operating SR   4.4.4.1     Reactor Coolant System Power-Operated Relief Valves SR   4.5.2.f     Emergency Core Cooling Systems Subsystems SR   4.6.2.1.b   Containment Spray System SR   4.6.3.3     Containment Isolation Valves SR   4.6.5       Containment Vacuum Relief System SR   4.7.1.1     Main Steam Line Code Safety Valves SR   4.7.1.5     Main Steam Line Isolation Valves SR   4.7.13.a    Essential Services Chilled Water System Technical Specification Bases Control Program New TS 6.8.4.m will be added to be consistent with NUREG-1431, Section 5.5.14, "Technical Specifications (TS) Bases Control Program."
Page A1-3 of 8

i a Attachment 1 to SERIAL: HNP-07-005 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT NO. 1 DOCKET NO. 50-400/LICENSE NO. NPF-63 REQUEST FOR LICENSE AMENDMENT DESCRIPTION, PROPOSED CHANGE, BACKGROUND, AND TECHNICAL ANALYSIS

===Background===
The following regulatory bases and guidance and industry guidance are applicable to the proposed amendment:
Title 10 of the Code of FederalRegulations (10 CFR), Section 50.55a(f)(5)(ii),
requires that if a revised inservice test program for a facility conflicts with the technical specification for the facility, the licensee shall apply to the Commission for amendment of the technical specifications to conform the technical specification to the revised program.
10 CFR 50.36, Technical specifications, Paragraph (c)(3) requires that the TS include surveillance requirements. This paragraph states that surveillance requirements are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met.
10 CFR 50.92, Issuance of amendment, requires that a proposed license amendment for a facility licensed under 10 CFR 50.22 shall not: (i) involve a significant increase in the probability or consequences of an accident previously evaluated; or (ii) create the possibility of a new or different kind of accident from any accident previously evaluated; or (iii) involve a significant reduction in a margin of safety.
10 CFR 51.22, Criterionfor categoricalexclusion; identification of licensing and regulatoryactions eligible for categoricalexclusion or otherwise not requiring environmental review, states that an environmental assessment or environment impact statement for a license amendment is not required if: (i) the amendment involves no significant hazards consideration, (ii) there is no significant change in the types or significant increase in the amounts of any effluents that may be released offsite, and (iii) there is no significant increase in individual or cumulative occupational radiation exposure.
Page A1-4 of 8

Attachment 1 to SERIAL: HNP-07-005 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT NO. 1 DOCKET NO. 50-400/LICENSE NO. NPF-63 REQUEST FOR LICENSE AMENDMENT DESCRIPTION, PROPOSED CHANGE, BACKGROUND, AND TECHNICAL ANALYSIS Background (continued)
NUREG-1 431, Revision 3, Standard Technical Specifications Westinghouse Plants, Specifications, published in June 2004, contains the improved Standard Technical Specifications (STS) for Westinghouse plants. Revision 3 incorporated the cumulative changes to Revision 1 and 2, and resulted from the experience gained from license amendment applications to convert to these improved STS or to adopt partial improvements to existing TS. Licensees are encouraged to upgrade their TS consistent with those criteria and conforming, to the practical extent, to 'Revision 3 to the improved STS. This proposed amendment to relocate TS SR 4.0.5 to the administrative section is consistent with the guidance contained in NUREG-1431, Revision 3.
Page Al-5 of 8

Attachment 1 to SERIAL: HNP-07-005 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT NO. 1 DOCKET NO. 50-400/LICENSE NO. NPF-63 REQUEST FOR LICENSE AMENDMENT DESCRIPTION, PROPOSED CHANGE, BACKGROUND, AND TECHNICAL ANALYSIS Technical Analysis The following analyzes the specific changes proposed in this amendment:
Inservice Testing Program TS SR 4.0.5 establishes the surveillance requirements for inservice inspection and testing of ASME Class 1, 2, and 3 components for the Harris Nuclear Plant. TS SR 4.0.5.a states:
    "Inservice inspection of ASME Code Class 1, 2, and 3 components (including supports and attachments) and inservice testing of ASME Code Class 1, 2, and 3 pumps and valves shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50.55a(b)(3)(v), except where specific written relief has been granted by the Commission pursuant to 10 CFR Part 50, Section 50.55a(f)(6)(i) or Section 50.55a(g)(6)(i)."
The regulations in 10 CFR 50.55a(f)(4) establish the effective Code edition and addenda to be used by licensees for performing IST of pumps and valves. Pursuant to 10 CFR 50.55a(f)(4)(ii), HNP submitted the third 10-year interval of the IST Program to the NRC in a letter (HNP-06-141) dated December 18, 2006. The IST Program for the third 10-year interval was updated to comply with the appropriate revisions of the ASME Code for Operation and Maintenance (OM) of Nuclear Power Plants and included the 2001 Edition through the 2003 Addenda subject to the limitations and modifications in 10 CFR 50.55a(b)(3). As a result, TS SR 4.0.5 references the ASME Section XI Code rather than the ASME OM Code for the IST Program.
According to 10 CFR 50.55a(f)(5)(ii), "If a revised inservice test program for a facility conflicts with the technical specification for the facility, the licensee shall apply to the Commission for amendment of the technical specifications to conform the technical specification to the revised program." Therefore, since the revised IST Program conflicts with the existing TS SR 4.0.5, then the proposed amendment will correct this condition.
Page AI-6 of 8

o ,l Attachment 1 to SERIAL: HNP-07-005 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT NO. 1 DOCKET NO. 50-400/LICENSE NO. NPF-63 REQUEST FOR LICENSE AMENDMENT DESCRIPTION, PROPOSED CHANGE, BACKGROUND, AND TECHNICAL ANALYSIS Technical Analysis (continued)
For the IST Program, the proposed amendment will delete TS SR 4.0.5 and its associated Bases, relocate the IST Program to Section 6 of the TS, and delete the reference to TS SR 4.0.5 in the associated individual TS. These changes do not eliminate any inservice tests and do not relinquish HNP of its responsibility to seek relief from Code test requirements when they are impractical. These changes will eliminate the ASME Code inconsistency between the IST Program and the TS as required by 10 CFR 50.55a(f)(5)(ii).
Inservice Inspection Program TS SR 4.0.5 establishes the surveillance requirements for inservice inspection and testing of ASME Class 1, 2, and 3 components for the Harris Nuclear Plant.
TS SR 4.0.5.a states:
         "Inservice inspection of ASME Code Class 1, 2, and 3 components (including supports and attachments) and inservice testing of ASME Code Class 1, 2, and 3 pumps and valves shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50.55a(b)(3)(v), except where specific written relief has been granted by the Commission pursuant to 10 CFR Part 50, Section 50.55a(f)(6)(i) or Section 50.55a(g)(6)(i)."
When NUREG-1431 was developed, the ISI requirements were removed from the improved STS since they were identical to the requirements of 10 CFR 50.55a and were, thus, redundant. Therefore, the proposed amendment will remove the ISI requirements of TS SR 4.0.5 and move them to an owner-controlled program consistent with NUREG-1 431.
The proposed amendment will delete TS SR 4.0.5, which refers to the ISI Program, ASME Section XI, and 10 CFR 50.55a. These changes do not eliminate any inservice inspections and do not relinquish HNP of its responsibility to seek relief from Code inspection requirements when they are impractical. These changes will eliminate the redundancy of ISI requirements from the TS as these requirements are already covered under 10 CFR 50.55a.
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4 Attachment 1 to SERIAL: HNP-07-005 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT NO. 1 DOCKET NO. 50-400/LICENSE NO. NPF-63 REQUEST FOR LICENSE AMENDMENT DESCRIPTION, PROPOSED CHANGE, BACKGROUND, AND TECHNICAL ANALYSIS Technical Analysis (continued)
Technical Specifications Bases Control Proqram The proposed amendment will add new TS Section 6.8.4.m to establish a TS Bases Control Program. The TS changes that are approved by the NRC Consolidated Line Item Improvement Program (CLIIP) often include the requirement for a TS Bases Control Program as a condition for implementing a particular Technical Specification Task Force (TSTF) traveler (e.g., TSTF-358 associated with missed surveillances, TSTF-359 associated with mode change limitations, and TSTF-372 associated with the addition of LCO 3.0.8 on the inoperability of snubbers). This change is requested because it is anticipated that future CLIIP TS changes will invoke a similar requirement for a TS Bases Control Program. Therefore, this change will allow HNP to be consistent with NUREG-1431, and it does not affect the other changes to the ISI and IST Programs requested by this proposed amendment.
Summary The proposed amendment is administrative in nature, and it will relocate the IST requirements to the administrative section of TS, remove the ISI requirements from TS and move them to an owner-controlled program, and establish a TS Bases Control Program. The changes to the IST Program will eliminate the ASME Code inconsistency between the IST Program and the TS as required by 10 CFR 50.55a(f)(5)(ii). The changes to the ISI Program will eliminate the redundancy of ISI requirements from the TS as these requirements are already covered under 10 CFR 50.55a. The addition of a TS Bases Control Program will allow HNP to be consistent with improved STS. The proposed changes are consistent with NUREG-1 431.
Conclusion HNP has concluded, based on the considerations discussed above, that: (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 &#xfd;health and safety of the public.
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Attachment 2 to SERIAL: HNP-07-005 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT NO. 1 DOCKET NO. 50-400/LICENSE NO. NPF-63 REQUEST FOR LICENSE AMENDMENT 10 CFR 50.92 NO SIGNFICANT HAZARDS EVALUATION 10 CFR 50.92 NO SIGNIFICANT HAZARDS EVALUATION Page A2-1 of 5

l Attachment 2 to SERIAL: HNP-07-005 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT NO. 1 DOCKET NO. 50-400/LICENSE NO. NPF-63 REQUEST FOR LICENSE AMENDMENT 10 CFR 50.92 NO SIGNFICANT HAZARDS EVALUATION A written evaluation of the significant hazards consideration of a proposed license amendment is required by 10 CFR 50.92. Harris Nuclear Plant (HNP) has evaluated the proposed amendment and determined that it involves no significant hazards consideration. According to 10 CFR 50.92, a proposed amendment to an operating license involves no significant hazards consideration if operation of the facility in accordance with the proposed amendment would not:
: 1. Involve a significant increase in the probability or consequences of an accident previously evaluated; or
: 2. Create the possibility of a new or different kind of accident from any accident previously evaluated; or
: 3. Involve a significant reduction in a margin of safety The basis for this determination is presented below.
Proposed Change In accordance with the Code of Federal Regulations, Title 10, Part 50.90, "Application for amendment of license or construction permit," Carolina Power and Light Company (CP&L) doing business as Progress Energy Carolinas, Inc., requests a license amendment for the Harris Nuclear Plant (HNP) Technical Specifications (TS). The proposed amendment will: 1) relocate the Inservice Testing (IST)
Program of TS Surveillance Requirement (SR) 4.0.5 to the administrative section of the TS under TS 6.8.4 and revise the wording to be consistent with NUREG-1431, Revision 3, Standard Technical Specifications Westinghouse Plants, Specifications, published in June 2004; 2) remove the Inservice Inspection (ISI) activities currently contained in TS SR 4.0.5 and move them to an owner-controlled program consistent with NUREG-1431 since these requirements are redundant with the regulations under 10 CFR 50.55a; and 3) establish a TS Bases Control Program consistent with NUREG-1431. To incorporate these changes, several other sections of TS that reference TS SR 4.0.5 will be revised to reference the IST Program. The proposed changes are administrative in nature, and they do not eliminate any tests (IST) or inspections (ISI). This proposed amendment will resolve the ASME Code inconsistency between the IST Program and the TS as required by 10 CFR 50.55a(f)(5)(ii), and will support implementation of the third ten-year intervals of the ISI and IST Programs. These administrative type changes to the TS have no adverse impact on the public health and safety.
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Attachment 2 to SERIAL: HNP-07-005 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT NO. 1 DOCKET NO. 50-400/LICENSE NO. NPF-63 REQUEST FOR LICENSE AMENDMENT 10 CFR 50.92 NO SIGNFICANT HAZARDS EVALUATION Basis This amendment does not involve a significant hazards consideration for the following reasons:
: 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No.
The proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated, and it does not change an accident previously evaluated in the Final Safety Analysis Report (FSAR). The proposed change is administrative in nature, and it will continue to ensure that the inspection and testing requirements required by regulations are met. The American Society of Mechanical Engineers (ASME) Code requirements are established, reviewed and approved by ASME, the industry, and ultimately endorsed by the NRC for inclusion into 10 CFR 50.55a. Updates to the ASME Code reflect advances in technology and consider information obtained from plant operating experience to provide enhanced inspection and testing. Thus, the proposed change will revise TS to appropriately reference the ASME Code required by 10 CFR 50.55a for performing inservice testing, specifically referencing the ASME Code for Operation and Maintenance of Nuclear Power Plants, rather than the ASME Section XI Code. The proposed change does not affect operations, and the inspection and testing required is not an accident initiator.
Therefore, this amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.
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Attachment 2 to SERIAL: HNP-07-005 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT NO. 1 DOCKET NO. 50-400/LICENSE NO. NPF-63 REQUEST FOR LICENSE AMENDMENT 10 CFR 50.92 NO SIGNFICANT HAZARDS EVALUATION Basis (Continued)
: 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 amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated, and it does not change an accident previously evaluated in the Final Safety Analysis Report (FSAR). As noted above, the proposed change is administrative in nature, the inspection and testing required is not an accident initiator, and no new accident precursors are being introduced. The proposed change will revise TS to appropriately reference the ASME Code required by 10 CFR 50.55a for performing inservice testing, which will continue to ensure that the inspection and testing requirements required by regulations are met. Since inservice testing will continue to be performed in accordance with regulations, adequate assurance is provided to ensure that the safety-related pumps and valves will continue to operate as required. No new testing is required that could create a new or different type of accident.
Therefore, this amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.
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Attachment 2 to SERIAL: HNP-07-005 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT NO. 1 DOCKET NO. 50-400/LICENSE NO. NPF-63 REQUEST FOR LICENSE AMENDMENT 10 CFR 50.92 NO SIGNFICANT HAZARDS EVALUATION Basis (Continued)
: 3. Does the proposed change involve a significant reduction in a margin of safety?
Response: No.
The proposed amendment does not involve a significant reduction in a margin of safety. The proposed amendment does not adversely affect a plant safety limit or a limiting safety system setting, and does not alter a design basis limit for a parameter evaluated in the FSAR. The proposed change is administrative in nature, and it will continue to ensure that the inspection and testing requirements required by regulations are met. Since inservice testing will continue to be performed in accordance with regulations, adequate assurance is provided to ensure that the safety-related pumps and valves will continue to operate as required and perform their intended safety function.
Therefore, this amendment does not involve a significant reduction in a margin of safety.
Based on the above, HNP 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.
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Attachment 3 to SERIAL: HNP-07-005 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT NO. 1 DOCKET NO. 50-400/LICENSE NO. NPF-63 REQUEST FOR LICENSE AMENDMENT PROPOSED TECHNICAL SPECIFICATIONS (TS) CHANGES PROPOSED TECHNICAL SPECIFICATIONS (TS) CHANGES Page A3-1 of 17

I APPLICABILITY SURVEILLANCE REQUIREMENTS 4.0.1 Surveillance Requirements shall be met during the OPERATIONAL MODES or other conditions specified for individual Limiting Conditions for Operation unless otherwise stated in an individual Surveillance Requirement.
4.0.2. Each Surveillance Requirement shall be performed within the specified surveillance interval with a maximum allowable extension not to exceed 25% of the specified surveillance interval.
4.0.3 If it is discovered that a surveillance was not performed within its specified surveillance interval, then compliance with the requirement to declare the LCO not met may be delayed, from the time of discovery, up to 24 hours or up to the limit of the specified surveillance interval, whichever is less. This delay period is permitted to allow performance of the surveillance.
If the surveillance is not performed within the delay period, the LCO must immediately be declared not met, and the applicable ACTION requirements must be met.
When the surveillance is performed within the delay period and the surveillance criteria are not met, the LCO must immediately be declared not met, and the applicable ACTION requirements must be met.
Surveillance Requirements do not have to be performed on inoperable equipment.
4.0.4 Entry into an OPERATIONAL MODE or Other specified condition shall not be made unless the Surveillance Requirement(s) associated with the Limiting Condition for Operation has been performed within the stated surveillance interval or as otherwise specified. This provision shall not prevent passage through or to OPERATIONAL MODES as required to comply with ACTION requi rements.
4.0.5A   urveillance Requirements for inservice inspection and testing of ASME o      ass 1, 2, and 3 components shall be applicable as follows:
: a.      Inservice inspection of ASME Code Class 1, 2, and 3 components (including supports and attachments) and inservice testing of ASME Code Class 1, 2, and 3 pumps and valves shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR Part 50, Section 50.55a(f) or (g)as modified by 10 CFR 50.55a(b)(3)(v). except where specific written relief has been granted by the Commission pursuant to 10 CFR Part 50. Section 50.55a(f)(6)(i) or Section 50.55a(g)(6)(i);
SHEARON HARRIS    - UNIT 1             3/4 0-2                  Amendment No. 1e

APPLICABILITY SURVEILLANCE REQUIREMENTS (Continued) adPesrVeslCode and applicable Addendape          forming    inservice A YeaBole  r andnrsurel  Vese       Reuieds   frqunciespe     forday
: b. Theuroveilneiontervalspecifiedtion4..2aet-ionlIcabl to he aBoiler inspection and testing activities       actii    ties Qatermnloy d.
Presurfr Vse e for ev Code sery 3imntsstpeastionc pice     inspectionce Ant               pners92day ape 7siny      s iteaMonhllAils                           ncied pervel3     ayse Semanualyoreverly trmqinolog frqece         6mnhinservice At least o eformn            inspection       18 days once perandtein Every 9 months                At least once per 276 days e  NohninteASME Yearly Boiler  and Pressure Vessel or annually             Reqir       dfrqeeshl At least once     per 366 for days
: c. The provisions of Specification 4.0.2 are applicable to the above required frequencies for performing inservice inspection and testing acti vi ties;
: d. Performance of the above inservice inspection and testing activities shall be in addition to other specified Surveillance Requirements: and construed to supersede the requirements of any Technical
       ?>/A- 0-3 ha4s beeri Jc i 4                     4,mn                4o.
SHEARO)N HARRIS  -  UNIT 1               3/4 0-3I\A(~                  Y'4C        Ao

REACTIVITY CONTROL SYSTEMS CHARGING PUMP - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.2.3 One charging/safety injection pump in the boron injection flow path required by Specification 3.1.2.1 shall be OPERABLE and capable of being powered from an OPERABLE emergency power source.
APPLICABILITY:     MODES 4*, 5", and 6'#.
ACTION:
With no charging/safety injection pump OPERABLE or capable of being powered from an OPERABLE emergency power source, suspend all operations involving CORE ALTERATIONS or positive reactivity changes.
4.1.2.3.1     The above required charging/sa   ty injection pump shall be demonstrated OPERABLE by verifying, on rec rculation flow or in service supplying flow to the reactor coolant syste and reactor coolant pump seals, that a differential pressure across the pump f eat          than r e u to 2446 psid is developed when tested pursuant to S eci ica io 4.0.5 4.1.2.3.2 All charging/safety injection pumps, excluding the above required OPERABLE pump, shall be demonstrated inoperable* by verifying that each pump's motor circuit breaker is secured in the open position prior to the.
temperature of one or more of the RCS cold legs decreasing below 3257F and at least once per 31 days thereafter, except when the reactor vessel head is removed.
A maximum of one charging/safety injection pump shall be OPERABLE whenever the temperature of one or more of the RCS cold legs is less than or equal to 325 0 F and the reactor vessel head is in place.
An inoperable pump may be energized for testing provided the discharge of the pump has been isolated from the RCS by a closed isolation valve with power removed from the valve operator or by a manual isolation valve secured in the closed position.
For periods of no more than 1 hour, when swapping pumps, it is permitted that there be no OPERABLE charging/safety injection pump. No CORE ALTERATIONS or positive reactivity changes are permitted during this time.
SHEARON HARRIS - UNIT 1               3/4 1-9                    Amendment No..

REACTIVITY CONTROL SYSTEMS CHARGING PUMPS - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.4   At least two charging/safety injection pumps shall be OPERABLE.
APPLICABILITY:    MODES 1, 2, and 3.
ACTION:
With only one charging/safety injection pump OPERABLE, restore at least two charging/safety injection pumps to OPERABLE status within 72 hours or be in at least HOT STANDBY and borated to a SHUTDOWN MARGIN as specified in the CORE OPERATING LIMITS REPORT (COLR), plant procedure PLP-106 at 200&deg;F within the LE&
next 6 hours: restore at least two charging/safety injection pumps to OPERABLE, status within the next 7 days or be in HOT SHUTDOWN within the next 6 hours.
SURVEILLANCE REQUIREMENTS 4.1.2.,4 At least two charging/safety injection pumps shall be demonstrated OPERABLE by verifying, on recirculation flow or in service supplying flow to the Reactor Coolant System and reactor coolant pump seals, that a differential pressure across each pum               t an or equal to 2446 psid is developed when tested pursuant to pecification 4.0 SHEARON HARRIS  -  UNIT 1            3/4 1-10                   Amendment No.
* r REACTOR COOLANT SYSTEM 3/4.4.2  SAFETY VALVES SHUTDOWN LIMITING CONDITION FOR OPERATION 3.4.2.1 A minimum of one pressurizer Code safety valve shall be OPERABLE with a lift setting of 2485 psig +/- 1%.
APPLICABILITY:    MODES 4 and 5.
ACTION:
With no pressurizer Code safety valve OPERABLE, immediately suspend all operations involving positive reactivity changes and place an OPERABLE RHR loop into operation in the shutdown cooling mode.
SURVEILLANCE REQUIREMENTS
      .4.2-1  No add'*   na  requirements other than those required by Specification    .0.
The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.
SHEARON HARRIS - UNIT 1               3/4 4-8

REACTOR COOLANT SYSTEM OPERATING LIMITING CONDITION FOR OPERATION 3.4.2.2 All pressurizer Code safety valves shall be OPERABLE with a lift setting of 2485 psig +/- I%.*
APPLICABILITY:       MODES 1, 2, and 3.
ACTION:
With one pressurizer Code safety valve inoperable, either restore the inoperable valve to OPERABLE status within 15 minutes or be in at least HOT STANDBY within 6 hours and in at least HOT SHUTDOWN within the following 6 hours.
SURVEILLANCE REQUIREMENTS 4.4.22tN        dditional requirements other than those required by (Specification      .0.
The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.
 &#xfd;'-)MLAKUN MAKKI&#xfd;) -  UNI I I           614 4-j ArpoA.        00o.

REACTOR COOLANT SYSTEM RELIEF VALVES SURVEILLANCE REQUIREMENTS 4.4.4.1     In addition to the requirements of(LSpecification 4.0. each PORV shall be demonstrated OPERABLE at least once per 18-months-by:
: a.      Performing a CHANNEL CALIBRATION of the actuation instrumentation, and
: b.      Operating the valve through one complete cycle of full travel during MODES 3 or 4, prior to going to 325 0 F.
4.4.4.2 Each block valve shall be demonstrated OPERABLE at least once per 92 days t)y operating the valve through one complete cycle of full travel unless thie block valve is closed with power removed in order to meet the requi reme ents of ACTION b. or c. in Specification 3.4.4.
4.4.4.3 The accumulator for the safety-related PORVs shall be demonstrated OPERABLE at least once per 18 months by isolating the normal air and nitrogen supplies and operating the valves through a complete cycle of full travel.
SHEARON HARRIS - UNIT 1                3/4 4-12                    Amendment No. 276

EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
: d. At least once per 18 months by:
: 1. Verifying automatic interlock action of the RHR system from the Reactor Coolant System by ensuring that with a simulated or actual Reactor Coolant System pressure signal greater than or equal to 425 psig the interlocks prevent the valves from being opened.
: 2. A visual inspection of the containment sump and verifying that the subsystem suction inlets are not restricted by debris and that the sump components (trash racks, screens, etc.) show no evidence of structural distress or abnormal corrosion.
: e. At least once per 18 months by:
1.'    Verifying that each automatic valve   in the flow path actuates to its correct position on  safety injection actuation test signal and on safety   injection switchover to containment sump from an RWST Lo-Lo   level test signal, and
: 2. Verifying that each of the following pumps start automatically upon receipt of a safety injection actuation test signal:
a)     Charging/safety injection pump, b)     RHR pump.
: f. By verifying that each of the following pumps develops the e     e di       ntia pressure when tested pursuant to
: 1. Charging/safety injection pump (Refer to Specificati( pn
: 2. RHR pump &#xfd;u!100 psid at a flow rate of at least 3663 gjpm.
: g. By verifying that the locking mechanism is in place and lo( ked for the following High Head ECCS throttle valves:
: 1. Within 4 hours following completion of each valve strroking operation or maintenance on the valve when the ECCS subsystems are required to be OPERABLE, and SHEARON HARRIS - UNIT 1               3/4 5-5                    Amendment No. 77

CONTAINMENT SYSTEMS 3/4.6.2   DEPRESSURIZATION AND COOLING SYSTEMS CONTAINMENT SPRAY SYSTEM LIMITING CONDITION FOR OPERATION 3.6.2.1 Two independent Containment Spray Systems shall be OPERABLE with each Spray System capable of taking suction from the RWST and transferring suction to the containment sump.
APPLICABILITY:    MODES 1, 2, 3, and 4.
ACTION:
With one Containment Spray System inoperable, restore the inoperable Spray System to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours; restore the inoperable Spray System to OPERABLE status within the next 48 hours or be in COLD SHUTDOWN within the following 30 hours.
Refer also to Specification 3.6.2.3 Action.
SURVEILLA      REQUIREMENTS      ---     T~ v          e~r        h~-~~
4.6. .1 Each Containment Spray System shall be demonstrated OPERABLE:
: a. At least once per 31 days by verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked I sealed, or otherwise secured in position, is in its correct position;
: b. By verifying that, on an indicated recirculation flow of at least 1832 gpm, each pump develops a differential pressure of greater t hn orequal to 186 si when tested pursuant to S ecifica ion .0. :
: c. At least once per 18 months by:                                     (2
: 1. Verifying that each automatic valve in the flow path actuates to its correct position on a containment spray actuation test signal and
: 2. Verifying that each spray pump starts automatically on a containment spray actuation test signal.
: 3. Verifying that, coincident with an indication of containment spray pump running, each automatic valve from the sump and RWST actuates to its appropriate position following an RWST Lo-Lo test signal.
: d. At least once per 10 years by performing an air or smoke flow test through each spray header and verifying each spray nozzle is unobstructed.
SHEARON HARRIS - UNIT 1                3/4 6-11                  Amendment No.    "

CONTAINMENT SYSTEMS CONTAINMENT ISOLATION VALVES SURVEILLANCE REQUIREMENTS (Continued) 4.6.3.2 Each isolation valve shall be demonstrated OPERABLE at least once per~'*
18 months by:
: a. Verifying that on a Phase "A" Isolation test signal, each Phase "A" isolation valve actuates to its isolation position;
: b. Verifying that on a Phase "B" Isolation test signal, each Phase "B" isolation valve actuates to its isolation position; and
: c. Verifying that on a Containment Ventilation Isolation test signal, each normal, preentry purge makeup and exhaust, and containment vacuum relief valve actuates to its isolation position, and
: d. Verifying that, on a Safety Injection "S" test signal, each containment isolation valve receiving an "S" signal actuates to its isolation position, and
: e. Verifying that, on a Main Steam Isolation test signal, each main steam isolation valve actuates to its isolation position, and
: f. Verifying that, on a Main Feedwater Isolation test signal, each feedwater isolation valve actuates to its isolation position.
4.6.3.3 The isolation time of each power-operated or automatic valve shall be determined to be within its limit specified in the Technical Specification E        n      Pr r      plant procedure PLP-106, when tested pursuant to SHEARON HARRIS - UNIT 1              3/4 6-15                   Amendment No.

CONTAINMENT SYSTEMS 3/4.6.5  VACUUM RELIEF SYSTEM LIMITING CONDITION FOR OPERATION 3.6.5 The containment vacuum relief system shall be OPERABLE with an Actuation Setpoint of equal to or less negative than -2.5 inches water gauge differential pressure (containment pressure less atmospheric pressure)
APPLICABILITY:    MODES 1, 2, 3, and 4.
ACTION:
With one containment vacuum relief system inoperable, restore the system to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
SURVEILLANCE REQUIREMENTS required by SHEARON HARRIS  -  UNIT 1            3/4 6-32                 Amendment No. (e

3/4.7    PLANT SYSTEMS 3/4.7.1    TURBINE CYCLE SAFETY VALVES LIMITING CONDITION FOR OPERATION 3.7.1.1 All main steam line Code safety valves associated with each steam generator shall be OPERABLE with lift settings as specified in Table 3.7-2.
APPLICABILITY:    MODES 1, 2, and 3.
ACTION:
: a. With one or more main steam line Code safety valves inoperable, operation may proceed provided, that within 4 hours, either the inoperable valve is restored to OPERABLE status or the Power Range Neutron Flux High Trip Setpoint is reduced per Table 3.7-1; otherwise, be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours.
SURVEILLANCE REQUIREMENTS 4.No          additional SHEARON HARRIS - UNIT 1               3/4 7-1                 Amendment No. d

PLANT SYSTEMS MAIN STEAM LINE ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.7.1.5    Each main steam line isolation valve (MSIV)  shall be OPERABLE.
APPLICABILITY:    MODES 1, 2, 3, and 4.
ACTION:
MODE 1:
With one MSIV inoperable but open, POWER OPERATION may continue provided the inoperable valve is restored to OPERABLE status within 4 hours; otherwise be in HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours.
MODES 2, 3, and 4:
With one MSIV inoperable, subsequent operation in MODE 2, 3, or 4 may proceed provided the isolation valve is maintained closed. Otherwise, be in HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours.
SURVEILLANCE REQUIREMENTS 4.7.1.5 Each MSIV shall be demonstrated                   rf n .       full- closure within 5 seconds when tested pursuant to               &#xfd;no0    tS The provisions of Specification 4.0.4 are not applica                    no     ~
ESor    4.
SHEARON HARRIS - UNIT 1               3/4 7-9                  Amendment No.    &e

PLANT SYSTEMS 3/4.7.13   ESSENTIAL SERVICES CHILLED WATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.13 At least two independent Essential Services Chilled Water System loops shall be OPERABLE.
APPLICABILITY:   MODES 1, 2, 3, and 4.
ACTION:
With only one Essential Services Chilled Water System loop OPERABLE, restore at least two loops to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
SURVEILLANCE REQUIREMENTS 4.7.13 The Essential Services Chilled Water System shall be demonstrated OPERABLE by:
a.
and
: b. At least once per 18 months by demonst   ting that:
: 1. Non-essential portions of the ystem are automatically isolated upon receipt of a Sa ety Injection actuation signal, and
: 2. The system starts automatica ly on a Safety Injection actuation signal.
kY&#xfd;6GMeVt &#xfd;0.
SHEARON HARRIS - UNIT 1               3/4 7-30

INSERT A (Please add Insert A on new page 6-19d)
: 1. Inservice Testina Proaram This program provides controls for inservice testing of ASME Code Class 1, 2, and 3 components. The program shall include the following:
: 1. Testing frequencies specified in the ASME Code for Operation and Maintenance of Nuclear Power Plants and applicable Addenda as follows:
ASME Code for Operation and Maintenance of Nuclear Power Plants and applicable Addenda            Required Frequencies for terminology for inservice testing        performing inservice testing activities                           activities Weekly                    At least once per 7 days Monthly                   At least once per 31 days Quarterly or every 3 months           At least once per 92 days Semiannually or every 6 months           At least once per 184 days Every 9 months                At least once per 276 days Yearly or annually              At least once per 366 days Biennially or every 2 years          At least once per 731 days
: 2. The provisions of Specification 4.0.2 are applicable to the above required Frequencies for performing inservice testing activities,
: 3. The provisions of Specification 4.0.3 are applicable to inservice testing activities, and
: 4. Nothing in the ASME Code for Operation and Maintenance of Nuclear Power Plants shall be construed to supersede the requirements of any Technical Specification.
6-19d

INSERT A (continued)
: m. Technical Specifications (TS) Bases Control Proqram This program provides a means for processing changes to the Bases of these Technical Specifications.
: 1. Changes to the Bases of the TS shall be made under appropriate administrative controls and reviews.
: 2. Licensees may make changes to Bases without prior NRC approval provided the changes do not require either of the following:
: a. A change in the TS incorporated in the license, or
: b. A change to the FSAR or Bases that requires NRC approval pursuant to 10 CFR 50.59.
: 3. The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the FSAR.
: 4. Proposed changes that meet the criteria of Specification 6.8.4.m.2 above shall be reviewed and approved by the NRC prior to implementation.
Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71(e).
6-19e

Attachment 4 to SERIAL: HNP-07-005 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT NO. 1 DOCKET NO. 50-400/LICENSE NO. NPF-63 REQUEST FOR LICENSE AMENDMENT REVISED TECHNICAL SPECIFICATION (TS) PAGES REVISED TECHNICAL SPECIFICATIONS (TS) PAGES Page A4-1 of 17

APPLICABILITY SURVEILLANCE REQUIREMENTS 4.0.1 Surveillance Requirements shall be met during the OPERATIONAL MODES or other conditions specified for individual Limiting Conditions for Operation unless otherwise stated in an individual Surveillance Requirement.
4.0.2 Each Surveillance Requirement shall be performed within the specified surveillance interval with a maximum allowable extension not to exceed 25% of the specified surveillance interval.
4.0.3 If it is discovered that a surveillance was not performed within its specified surveillance interval, then compliance with the requirement to declare the LCO not met may be delayed, from the time of discovery, up to 24 hours or up to the limit of the specified surveillance interval, whichever is less. This delay period is permitted to allow performance of the surveillance.
If the surveillance is not performed within the delay period, the LCO must immediately be declared not met, and the applicable ACTION requirements must be met.
When the surveillance is performed within the delay period and the surveillance criteria are not met, the LCO must immediately be declared not met, and the applicable ACTION requirements must be met.
Surveillance Requirements do not have to be performed on inoperable equipment.
4.0.4 Entry into an OPERATIONAL MODE or other specified condition shall not be made unless the Surveillance Requirement(s) associated with the Limiting Condition for Operation has been performed within the stated surveillance interval or as otherwise specified. This provision shall not prevent passage through or to OPERATIONAL MODES as required to comply with ACTION requirements.
4.0.5   DELETED SHEARON HARRIS   - UNIT 1            3/4 0-2                   Amendment No.

APPLICABILITY SURVEILLANCE REQUIREMENTS (Continued)
Page 3/4 0-3 has been deleted by Amendment No.
SHEARON HARRIS - UNIT 1             3/4 0-3                  Amendment No.

REACTIVITY CONTROL SYSTEMS CHARGING PUMP - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.2.3 One charging/safety injection pump in the boron injection flow path required by Specification 3.1.2.1 shall be OPERABLE and capable of being powered from an OPERABLE emergency power source.
APPLICABILITY:   MODES 4", 5*#, and 6'#.
ACTION:
With no charging/safety injection pump OPERABLE or capable of being powered from an OPERABLE emergency power source, suspend all operations involving CORE ALTERATIONS or positive reactivity changes.
SURVEILLANCE REQUIREMENTS 4.1.2.3.1 The above required charging/safety injection pump shall be demonstrated OPERABLE by verifying, on recirculation flow or in service supplying flow to the reactor coolant system and reactor coolant pump seals.,
that a differential pressure across the pump of greater than or equal to 2446 psid is developed when tested pursuant to the Inservice Testing Program.
4.1.2.3.2 All charging/safety injection pumps, excluding the above required OPERABLE pump, shall be demonstrated inoperable" by verifying that each pump",s motor circuit breaker is secured in the open position prior to the temperature of one or more of the RCS cold legs decreasing below 325&deg;F and at least once per 31 days thereafter, except when the reactor vessel head is removed.
A maximum of one charging/safety injection pump shall be OPERABLE whenever the temperature of one or more of the RCS cold legs is less than or equal to 325&deg;F and the reactor vessel head is in place.
An inoperable pump may be energized for testing provided the discharge of the pump has been isolated from the RCS by a closed isolation valve with power removed from the valve operator or by a manual isolation valve secured in the closed position.
For periods of no more than 1 hour, when swapping pumps, it is permitted that there be no OPERABLE charging/safety injection pump. No CORE ALTERATIONS or positive reactivity changes are permitted during this time.
SHEARON HARRIS - UNIT 1               3/4 1-9                  Amendment No.

REACTIVITY CONTROL SYSTEMS CHARGING PUMPS - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.4  At least two charging/safety injection pumps shall be OPERABLE.
APPLICABILITY:   MODES 1, 2, and 3.
ACTION:
With only one charging/safety injection pump OPERABLE, restore at least two charging/safety injection pumps to OPERABLE status within 72 hours or be in at least HOT STANDBY and borated to a SHUTDOWN MARGIN as specified in the CORE OPERATING LIMITS REPORT (COLR), plant procedure PLP-106 at 200&deg;F within the next 6 hours; restore at least two charging/safety injection pumps to OPERABLE status within the next 7 days or be in HOT SHUTDOWN within the next 6 hours.
SURVEILLANCE REQUIREMENTS 4.1.2.4 At least two charging/safety injection pumps shall be demonstrated OPERABLE by verifying, on recirculation flow or in service supplying flow to the Reactor Coolant System and reactor coolant pump seals, that a differential pressure across each pump of greater than or equal to 2446 psid is developed when tested pursuant to the Inservice Testing Program.                         I SHEARON HARRIS - UNIT 1             3/4 1-10                  Amendment No.

REACTOR COOLANT SYSTEM 3/4.4.2  SAFETY VALVES SHUTDOWN LIMITING CONDITION FOR OPERATION 3.4.2.1 A minimum of one pressurizer Code safety valve shall be OPERABLE with a lift setting of 2485 psig +/- 1%.
APPLICABILITY:   MODES 4 and 5.
ACTION:
With no pressurizer Code safety valve OPERABLE, immediately suspend all operations involving positive reactivity changes and place an OPERABLE RHR loop into operation in the shutdown cooling mode.
SURVEILLANCE REQUIREMENTS 4.4.2.1 No additional requirements other than those required by the Inservice I Testing Program.                                                               I The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.
SHEARON HARRIS - UNIT 1             3/4 4-8                    Amendment No.

REACTOR COOLANT SYSTEM OPERATING LIMITING CONDITION FOR OPERATION 3.4.2.2 All pressurizer Code safety valves shall be OPERABLE with a lift setting of 2485 psig +/- 1%."
APPLICABILITY:   MODES 1, 2, and 3.
ACTION:
With one pressurizer Code safety valve inoperable, either restore the inoperable valve to OPERABLE status within 15 minutes or be in at least HOT STANDBY within 6 hours and in at least HOT SHUTDOWN within the following 6 hours.
SURVEILLANCE REQUIREMENTS 4.4.2.2 No additional requirements other than those required by the Inservice I Testing Program.                                                               I The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.
SHEARON HARRIS - UNIT 1              3/4 4-9                  Amendment No.

REACTOR COOLANT SYSTEM RELIEF VALVES SURVEILLANCE REQUIREMENTS 4.4.4.1 In addition to the requirements of the Inservice Testing Program, each PORV shall be demonstrated OPERABLE at least once per 18 months by:
: a. Performing a CHANNEL CALIBRATION of the actuation instrumentation, and
: b. Operating the valve through one complete cycle of full travel during MODES 3 or 4, prior to going to 325 0 F.
4.4.4.2 Each block valve shall be demonstrated OPERABLE at least once per 92 days by operating the valve through one complete cycle of full travel unless the block valve is closed with power removed in order to meet the requirements of ACTION b. or c. in Specification 3.4.4.
4.4.4.3 The accumulator for the safety-related PORVs shall be demonstrated OPERABLE at least once per 18 months by isolating the normal air and nitrogen supplies and operating the valves through a complete cycle of full travel.
SHEARON HARRIS  - UNIT 1            3/4 4-12                 Amendment No.

EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
: d. At least once per 18 months by:
: 1. Verifying automatic interlock action of the RHR system from the Reactor Coolant System by ensuring that with a simulated or actual Reactor Coolant System pressure signal greater than or equal to 425 psig the interlocks prevent the valves from being opened.
: 2. A visual inspection of the containment sump and verifying that the subsystem suction inlets are not restricted by debris and that the sump components (trash racks, screens, etc.) show no evidence of structural distress or abnormal corrosion.
: e. At least once per 18 months by:
: 1. Verifying that each automatic valve   in the flow path actuates to its correct position on  safety injection actuation test signal and on safety   injection switchover to containment sump from an RWST Lo-Lo   level test signal, and
: 2. Verifying that each of the following pumps start automatically upon receipt of a safety injection actuation test signal:
a)     Charging/safety injection pump, b)     RHR pump.
: f. By verifying that each of the following pumps develops the required differential pressure when tested pursuant to the Inservice Testing Program:
: 1. Charging/safety injection pump (Refer to Specification 4.1.2.4)
: 2. RHR pump &#x17d; 100 psid at a flow rate of at least 3663 gpm.
: g. By verifying that the locking mechanism is in place and locked for the following High Head ECCS throttle valves:
: 1. Within 4 hours following completion of each valve stroking operation or maintenance on the valve when the ECCS subsystems are required to be OPERABLE, and
: 2. At least once per 18 months.
SHEARON HARRIS  - UNIT 1              3/4 5-5                    Amendment No.

Iio
* 31" CONTAINMENT SYSTEMS 3/4.6.2    DEPRESSURIZATION AND COOLING SYSTEMS CONTAINMENT SPRAY SYSTEM LIMITING CONDITION FOR OPERATION 3.6.2.1 Two independent Containment Spray Systems shall be OPERABLE with each Spray System capable of taking suction from the RWST and transferring suction to the containment sump.
APPLICABILITY:    MODES 1, 2, 3, and 4.
ACTION:
With one Containment Spray System inoperable, restore the inoperable Spray System to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours; restore the inoperable Spray System to OPERABLE status within the next 48 hours or be in COLD SHUTDOWN within the following 30 hours.
Refer also to Specification 3.6.2.3 Action.
SURVEILLANCE REQUIREMENTS 4.6.2.1    Each Containment Spray System shall be demonstrated OPERABLE:
: a. At least once per 31 days by verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position;
: b. By verifying that, on an indicated recirculation flow of at least 1832 gpm, each pump develops a differential pressure of greater than or equal to 186 psi when tested pursuant to the Inservice Testing Program;
: c. At least once per 18 months by:
: 1. Verifying that each automatic valve in the flow path actuates to its correct position on a containment spray actuation test signal and
: 2. Verifying that each spray pump starts automatically on a containment spray actuation test signal.
: 3. Verifying that, coincident with an indication of containment spray pump running, each automatic valve from the sump and RWST actuates to its appropriate position following an RWST Lo-Lo test signal.
: d. At least once per 10 years by performing an air or smoke flow test through each spray header and verifying each spray nozzle is unobstructed.
SHEARON HARRIS - UNIT 1                 3/4 6-11                 Amendment No.

CONTAINMENT SYSTEMS CONTAINMENT ISOLATION VALVES SURVEILLANCE REQUIREMENTS (Continued) 4.6.3.2 Each isolation valve shall be demonstrated OPERABLE at least once per 18 months by:
: a. Verifying that on a Phase "A" Isolation test signal, each Phase "A" isolation valve actuates to its isolation position;
: b. Verifying that on a Phase "B" Isolation test signal, each Phase "B" isolation valve actuates to its isolation position; and
: c. Verifying that on a Containment Ventilation Isolation test signal, each normal, preentry purge makeup and exhaust, and containment vacuum relief valve actuates to its isolation position, and
: d. Verifying that, on a Safety Injection "S" test signal, each containment isolation valve receiving an "S" signal actuates to its isolation position, and
: e. Verifying that, on a Main Steam Isolation test signal, each main steam isolation valve actuates to its isolation position, and
: f. Verifying that, on a Main Feedwater Isolation test signal, each feedwater isolation valve actuates to its isolation position.
4.6.3.3 The isolation time of each power-operated or automatic valve shall be determined to be within its limit specified in the Technical Specification Equipment List Program, plant procedure PLP-106, when tested pursuant to the Inservice Testing Program.
SHEARON HARRIS - UNIT 1              3/4 6-15                 Amendment No.

4 CONTAINMENT SYSTEMS 3/4.6.5  VACUUM RELIEF SYSTEM LIMITING CONDITION FOR OPERATION 3.6.5 The containment vacuum relief system shall be OPERABLE with an Actuation Setpoint of equal to or less negative than -2.5 inches water gauge differential pressure (containment pressure less atmospheric pressure)
APPLICABILITY:   MODES 1, 2, 3, and 4.
ACTION:
With one containment vacuum relief system inoperable, restore the system to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
SURVEILLANCE REQUIREMENTS 4.6.5 No additional requirements other than those required by the Inservice Testing Program.                                                              I SHEARON HARRIS - UNIT 1             3/4 6-32                  Amendment No.

3/4.7   PLANT SYSTEMS 3/4.7.1   TURBINE CYCLE SAFETY VALVES LIMITING CONDITION FOR OPERATION 3.7.1.1 All main steam line Code safety valves associated with each steam generator shall be OPERABLE with lift settings as specified in Table 3.7-2.
APPLICABILITY:   MODES 1, 2, and 3.
ACTION:
: a. With one or more main steam line Code safety valves inoperable, operation may proceed provided, that within 4 hours, either the inoperable valve is restored to OPERABLE status or the Power Range Neutron Flux High Trip Setpoint is reduced per Table 3.7-1; otherwise, be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours.
SURVEILLANCE REQUIREMENTS 4.7.1.1 No additional requirements other than those required by the Inservice I Testing Program.                                                                I SHEARON HARRIS - UNIT 1               3/4 7-1                 Amendment No.

0 -*, 4 a' PLANT SYSTEMS MAIN STEAM LINE ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.7.1.5     Each main steam line isolation valve (MSIV) shall be OPERABLE.
APPLICABILITY:     MODES 1, 2, 3, and 4.
ACTION:
MODE 1:
With one MSIV inoperable but open, POWER OPERATION may continue provided the inoperable valve is restored to OPERABLE status within 4 hours; otherwise be in HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours.
MODES 2, 3, and 4:
With one MSIV inoperable, subsequent operation in MODE 2, 3, or 4 may proceed provided the isolation valve is maintained closed. Otherwise, be in HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours.
SURVEILLANCE REQUIREMENTS 4.7.1.5 Each MSIV shall be demonstrated OPERABLE by verifying full closure within 5 seconds when tested pursuant to the Inservice Testing Program. The provisions of Specification 4.0.4 are not applicable for entry into MODES 3 or 4.
SHEARON HARRIS - UNIT 1                3/4 7-9                  Amendment No.

PLANT SYSTEMS 3/4.7.13   ESSENTIAL SERVICES CHILLED WATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.13 At least two independent Essential Services Chilled Water System loops shall be OPERABLE.
APPLICABILITY:   MODES 1, 2, 3, and 4.
ACTION:
With only one Essential Services Chilled Water System loop OPERABLE, restore at least two loops to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
SURVEILLANCE REQUIREMENTS 4.7.13 The Essential Services Chilled Water System shall be demonstrated OPERABLE by:
: a. Performance of surveillances as required by the Inservice Testing Program, and
: b. At least once per 18 months by demonstrating that:
: 1. Non-essential portions of the system are automatically isolated upon receipt of a Safety Injection actuation signal, and
: 2. The system starts automatically on a Safety Injection actuation signal.
SHEARON HARRIS - UNIT 1               3/4 7-30                Amendment No.

ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued)
: 1. Inservice Testing Program This program provides controls for inservice testing of ASME Code Class 1, 2, and 3 components. The program shall include the following:
: 1)   Testing frequencies specified in the ASME Code for Operation and Maintenance of Nuclear Power Plants and applicable Addenda as follows:
ASME Code for Operation and Maintenance of Nuclear             Required Frequencies for Power Plants and applicable            performing inservice testing Addenda terminology for inservice      activities testing activities Weekly                                 At  least once  per 7 days Monthly                                At  least once  per 31 days Quarterly or every 3 months            At  least once  per 92 days Semiannually or every 6 months         At  least once  per 184 days Every 9 months                         At  least once  per 276 days Yearly or annually                     At  least once  per 366 days Biennially or every 2 years            At  least once  per 731 days
: 2)   The provisions of Specification 4.0.2 are applicable to the above required Frequencies for performing inservice testing activities,
: 3)   The provisions of Specification 4.0.3 are applicable to inservice testing activities, and
: 4)   Nothing in the ASME Code for Operation and Maintenance of Nuclear Power Plants shall be construed to supersede the requirements of any Technical Specification.
SHEARON HARRIS - UNIT 1               6-19d                     Amendment No.

a ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued)
: m. Technical Specifications (TS) Bases Control Program This program provides a means for processing changes to the Bases of these Technical Specifications.
: 1)   Changes to the Bases of the TS shall be made under appropriate administrative controls and reviews.
: 2)   Licensees may make changes to Bases without prior NRC approval provided the changes do not require either of the following:
: a. A change in the TS incorporated in the license, or
: b. A change to the FSAR or Bases that requires NRC approval pursuant to 10 CFR 50.59.
: 3)   The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the FSAR.
: 4)   Proposed changes that meet the criteria of Specification 6.8.4.m.2 above shall be reviewed and approved by the NRC prior to implementation. Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71(e).
SHEARON HARRIS    - UNIT 1              6-19e                   Amendment No.

Attachment 5 to SERIAL: HNP-07-005 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT NO. 1 DOCKET NO. 50-400/LICENSE NO. NPF-63 REQUEST FOR LICENSE AMENDMENT PROPOSED TECHNICAL SPECIFICATIONS (TS) BASES CHANGES (FOR INFORMATION ONLY)
PROPOSED TECHNICAL SPECIFICATIONS (TS) BASES CHANGES (FOR INFORMATION ONLY)
Page A5-1 of 2

APPLICABILITY BASES (g)EasBodifierad ASEBoiler       and Pressure   Vessel byess10 CFRse    Code and 5Code)((          apialRedaplicbef fromenya ofTheiaov Addenda.               0 Thisirdb F 055~)o clreqirementso hasbeprovided              in wnsrecnistingcyteCmissuvionlande isnoteralpar relat clreqificantiohis specification iasbein      esto aensrecnistency provided     c rfha     t                       inservice teCminssuvillance inspection f isnoteralsar perfori thof     gouthese   Technical   Specifications.adt                    eoean thougou therin      these ce Technical testingSpecifications.adt inspecioand  activities.                       r eoeaneSectionmiute miute XI of the rhsseliiative toldsa                   the lrfcaino frequencies for performingthreuedisvcenpcio th nevc npcinand        testing activities.rqie yScto                                      Io h Under the terms of this specification, the more restrictive requirements of the Technical Specifications take precedence over the ASME Boiler and Pressure Vessel Code and applicable Addenda. For example, the requirements of Specification 4.0.4 to perform surveillance activities prior to entry into an OPERATIONAL MODE or other specified applicability condition takes precedence over the ASME Boiler and Pressure Vessel Code provision which allows pumps to be tested up to 1 week after return to normal operation. And for example, the Technical Specification definition of OPERABLE does not grant a grace period before a device that is not capable of performing its specified function is declared inoperable and takes precedence over the ASME Boiler and Pressure Vessel Code provision which allows a valve to be incapable of performing its                          ,
spcfedfntinfrup to 24 hours before being declared inoperable.
SHEARON HARRIS - UNIT 1                       B 3/4 0-3                           Amendment No. U11i7}}