ML11172A071

Application to Modify Technical Specifications for Use of Areva Advanced W17 Htp Fuel (TS-SQN-2011-07)

ML11172A071
Person / Time
Site:	Sequoyah
Issue date:	06/17/2011
From:	Krich R Tennessee Valley Authority
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
TS-SQN-2011-07
Download: ML11172A071 (79)

Text

Tennessee Valley Authority 1101 Market Street, LP 3R Chattanooga, Tennessee 37402-2801 R. M. Krich Vice President Nuclear Licensing June 17, 2011 10 CFR 50.90 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 Sequoyah Nuclear Plant, Units 1 and 2 Facility Operating License Nos. DPR-77 and DPR-79 NRC Docket Nos. 50-327 and 50-328

Subject:

Application to Modify Technical Specifications for Use of AREVA Advanced W17 HTP Fuel (TS-SQN-2011-07)

In accordance with the provisions of 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," the Tennessee Valley Authority (TVA) is submitting a request for an amendment to Facility Operating License Nos.

DPR-77 and DPR-79 for the Sequoyah Nuclear Plant (SQN), Units 1 and 2, respectively.

SQN currently uses AREVA Mark BW 17x17 fuel assemblies in both SQN, Units 1 and 2 (with four Advanced Mark-BW(A) assemblies in SQN, Unit 1, scheduled to be discharged in the spring 2012). To address fuel assembly distortion and its resultant fuel handling issues, a more robust fuel (AREVA Advanced W17 High Thermal Performance (HTP) fuel) has been selected for use at SQN, Units 1 and 2.

This license amendment request seeks to amend the licensing basis and the Technical Specifications (TS) to permit the use of AREVA Advanced W1 7 HTP fuel at SQN. The AREVA Advanced W1 7 HTP fuel assembly design consists of standard uranium dioxide (U0 2) fuel pellets with gadolinium oxide (Gd 20 3) burnable poison and M5TM cladding. The NRC has previously approved the use of similar fuel at other plants.

The enclosure provides a description of the proposed changes, technical evaluation of the proposed changes, regulatory evaluation, and a discussion of environmental considerations. Attachments 1 and 2 to the enclosure provide the existing TS and Bases pages marked-up to show the proposed changes. Attachments 3 and 4 to the enclosure provide the existing TS and Bases pages retyped to show the proposed changes. Attachment 5 contains the technical basis to support the requested fuel printed on recycled paper

U.S. Nuclear Regulatory Commission Page 2 June 17, 2011 change and the associated TS changes. The AREVA Small Break Loss of Coolant Accident (LOCA) and Realistic Large Break LOCA evaluations are provided as Attachments 6 and 7, respectively.

Attachments 5, 6, and 7 contain information that AREVA NP considers to be proprietary in nature and subsequently, pursuant to 10 CFR 2.390, Public inspections, exemptions, requests for withholding," paragraph (a)(4), TVA requests that such information be withheld from public disclosure. Attachment 11 contains the affidavits from AREVA NP supporting this request. Attachments 8, 9, and 10 contain the redacted versions of the proprietary attachments with the proprietary material removed, which are suitable for public disclosure.

SQN plans to refuel and operate with AREVA Advanced W1 7 HTP fuel beginning with the cycles following the refueling outages in the fall of 2012 for SQN, Unit 2, and in the fall of 2013 for SQN, Unit 1. The transition is planned to occur over two refueling cycles on each unit. TVA requests review and approval for the use of AREVA Advanced W17 HTP fuel in SQN, Units 1 and 2, including use in a mixed core by June 2012 with implementation to occur for SQN, Unit 2, no later than the startup from the SQN, Unit 2, fall 2012 refueling outage. SQN, Unit 1, is scheduled to begin loading the AREVA Advanced W1 7 HTP fuel during the SQN, Unit 1, fall 2013 refueling outage.

TVA has determined that there are no significant hazards considerations associated with the proposed change and that the change qualifies for a categorical exclusion from environmental review pursuant to the provisions of 10 CFR 51.22(c)(9).

The SQN Plant Operations Review Committee and the SQN Nuclear Safety Review Board have reviewed this proposed change and determined that operation of SQN in accordance with the proposed change will not endanger the health and safety of the public.

Additionally, in accordance with 10 CFR 50.91(b)(1), TVA is sending a copy of this letter, the enclosure, and the non-proprietary attachments to the Tennessee Department of Environment and Conservation.

There are no regulatory commitments associated with this submittal. Please address any questions regarding this request to Dan Green at 423-751-8423.

U.S. Nuclear Regulatory Commission Page 3 June 17, 2011 I declare under penalty of perjury that the foregoing is true and correct. Executed on this 17th day of June 2011.

Respectfully, R. M. Krich

Enclosure:

Evaluation of Proposed Change cc (Enclosure):

NRC Regional Administrator - Region II NRC Resident Inspector - Sequoyah Nuclear Plant Director, Division of Radiological Health - Tennessee State Department of Environment and Conservation

ENCLOSURE TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT UNITS 1 AND 2 EVALUATION OF PROPOSED CHANGE

Subject:

Application to Modify Technical Specifications for Use of AREVA Advanced W17 HTP Fuel (TS-SQN-2011-07)

1.

SUMMARY

DESCRIPTION

2.

DETAILED DESCRIPTION

3.

TECHNICAL EVALUATION

4.

REGULATORY EVALUATION 4.1 Applicable Regulatory Requirements/Criteria 4.2 Precedent 4.3 Significant Hazards Consideration 4.4 Conclusions

5.

ENVIRONMENTAL CONSIDERATION

6.

REFERENCES ATTACHMENTS

1.

Proposed TS Changes (Mark-Ups) for SQN, Units 1 and 2

2.

Proposed TS Bases Changes (Mark-Ups) for SQN, Units 1 and 2

3.

Proposed TS Changes (Final Typed) for SQN, Units 1 and 2

4.

Proposed TS Bases Changes (Final Typed) for SQN, Units 1 and 2

5.

ANP-2986(P), Revision 2, Sequoyah HTP Fuel Transition, June 2011

6.

ANP-2971(P), Revision 1, Sequoyah Units 1 and 2 HTP Fuel S-RELAP5 Small Break LOCA Analysis, May 2011

7.

ANP-2970(P), Revision 0, Sequoyah Units 1 and 2 HTP Fuel Realistic Large Break LOCA Analysis, April 2011 E-1

8.

ANP-2986(NP), Revision 2, Sequoyah HTP Fuel Transition, June 2011

9.

ANP-2971 (NP), Revision 1, Sequoyah Units 1 and 2 HTP Fuel S-RELAP5 Small Break LOCA Analysis, May 2011

10.

ANP-2970(NP), Revision 0, Sequoyah Units 1 and 2 HTP Fuel Realistic Large Break LOCA Analysis, April 2011

11.

AREVA NP Affidavits E-2

1.0

SUMMARY

DESCRIPTION This evaluation supports a request to amend Operating Licenses DPR-77 and DPR-79 for the Sequoyah Nuclear Plant (SQN), Units 1 and 2, by revising the licensing basis and the Technical Specifications (TS) for SQN, Units 1 and 2 to allow the use of AREVA Advanced W17 High Thermal Performance (HTP) fuel in the SQN, Units 1 and 2, reactors. SQN plans to refuel and operate with AREVA Advanced W1 7 HTP fuel beginning with the refueling outage in the fall of 2012 for Unit 2 and the fall of 2013 for Unit 1. The transition is planned to occur over two refueling cycles for each SQN unit.

The AREVA Advanced W17 HTP fuel design consists of standard uranium dioxide (U0 2) fuel pellets with Gd 20 3 burnable poison and M5TM cladding.

2.0 DETAILED DESCRIPTION 2.1 Change to Fuel Design SQN, Units 1 and 2, currently use AREVA Mark-BW fuel assemblies, approved by Reference 2, with each assembly consisting of 264 rods (pins) and 25 guide tubes/instrument tubes. The pins may contain fuel or a fuel/neutron poison mixture. The assembly is held together by eight intermediate spacer grids, two end spacers and is closed at the top and bottom by end fittings. Lateral support and positioning of the fuel rods within an assembly is provided by the intermediate spacer grids secured in position by swaged, deflection limiting ferrules with an initial gap at eight guide tube locations.

The guide tubes provide channels which guide the rod cluster control assemblies (RCCAs) over their entire length of travel and form the longitudinal structure of the assembly. In selected fuel assemblies, the central guide tube houses incore instrumentation. The fuel is low enrichment U0 2 in the form of ceramic pellets clad in M5TM tubes.

SQN, Units 1 and 2, have experienced fuel distortion that has affected efficient fuel handling activities and has a potential for slow or incomplete RCCA insertion and grid spacer damage. In the SQN, Unit 1 Cycle 16 refueling outage, four lead assemblies were placed in the core. The lead test assemblies utilized a MONOBLOCTM guide tube and spacer grid attachment design that has a more robust lateral stiffness and is similar to the AREVA Advanced W17 HTP fuel proposed to be used at SQN. These assemblies are operating in their third cycle with an estimated end of cycle assembly burnup of 50,000 MWD/MTU.

Performance of the lead test assemblies has lead to the decision to introduce the AREVA Advanced W1 7 HTP fuel (with improvements in materials and the incorporation of intermediate flow mixers) into SQN, Units 1 and 2. Additional design details and evaluations of the AREVA Advanced W1 7 HTP fuel are in Attachment 4.

Along with the physical fuel change, a change from AREVA Mark-BW fuel design and evaluation methods to AREVA Advanced W17 HTP fuel design and evaluation methods is also required. These design and evaluation methods and their acceptance criteria are described in Attachment 4. To support the change in fuel from AREVA Mark-BW fuel to AREVA Advanced W1 7 HTP fuel (and transition cores with both fuel types) certain TS require changes. These changes are described below and are shown on the marked up pages in Attachment 1.

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2.2 Requested Technical Specification Changes Safety Limit (SL) 2.1.1, Reactor Core Safety Limits (SL)

Figure 2.1-1, Reactor Core Safety Limit - Four Loops in Operation The restrictions of this SL prevent overheating of the fuel and possible cladding perforation that would result in the release of fission products to the reactor coolant. The change to the SL will add limits for the departure from nucleate boiling ratio (DNBR) for each of the resident fuel types during fuel transition (i.e., Advanced W17 HTP fuel and Mark-BW fuel). The DNBR limits for each fuel type are as follows.

For the Advanced W1 7 HTP fuel design

> 1.132 for the BHTP correlation (Reference 3, pg. 4-1)

> 1.21 for the BWU-N correlation (Reference 6, pg. v)

For the Mark-BW fuel design

> 1.21 for the BWCMV correlation (Reference 4, pg. xviii, and Reference 5, pg. iv)

> 1.21 for the BWU-N correlation (Reference 6, pg. v)

The DNBR limits will be reflected in SL 2.1.1.1 as follows.

SL 2.1.1.1 The departure from nucleate boiling ratio (DNBR) shall be maintained > 1.132 for the BHTP correlation, > 1.21 for the BWU-N correlation, and > 1.21 for the BWCMV correlation.

Limitations for local fuel pin centerline temperature will be added as follows.

SL 2.1.1.2 The maximum local fuel pin centerline temperature shall be maintained < 4901'F, decreasing by 13.7 0F per 10,000 MWD/MTU of burnup for COPERNIC applications, and !5 46420F, decreasing by 580F per 10,000 MWD/MTU of burnup for TACO3 applications.

The COPERNIC and TACO3 applications for these limits are presented in References 7 and 8.

The current SL uses a combination of thermal power, pressurizer pressure, and the highest coolant loop operating temperature to restrict fuel operation in the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature. These parameters are indirect measures but are relatable to the departure from nucleate boiling (DNB) parameter which cannot be directly measured.

Core Safety Limit (CSL) lines are being modified as a result of the transition to the Advanced W17 HTP design and the implementation of the BHTP DNB correlation.

Section 7.3 of BAW-10220P (Reference 1) provides a description of the processes used to develop or validate CSL lines. For the transition to Advanced W1 7 HTP fuel at SQN, the revised CSL lines are developed using the LYNXT thermal-hydraulic analysis code, E-4

the Statistical Core Design method, and the BHTP and BWU-N CHF correlations. A comparison of the current and revised CSL lines is provided in Figure 1.

An evaluation of the existing Overtemperature AT and Overpower AT trip functions showed that with the reduced CSL lines adequate protection is being provided by the existing trip functions, so no change to the trip functions definitions was required.

Table 2.2-1, Reactor Trip System Instrumentation Setpoints Figure 3.2-1, Flow vs. Power for 4 Loops in Operation The thermal hydraulic analysis indicates that the transition from a full core of Mark-BW fuel to a full core of Advanced W1 7 HTP fuel will result in a small increase in bypass flow and a small decrease in the RCS loop flow due to the higher pressure drop of the Advanced W17 HTP fuel. However, coincident with the fuel transition, steam generators will be replaced at SQN, Unit 2. Steam generators were previously replaced in SQN, Unit 1, in 2003 prior to startup from the Unit 1 Cycle 13 refueling outage. The combined effect of the fuel transition and the steam generator replacement is a small net increase in RCS loop flow. Given this beneficial increase in RCS flow, the new replacement steam generators with minimal tube plugging, and favorable historical measured flow, the TS design flow rate requirement (Table 2.2-1) is being increased from 90,045 (87,000 x 1.035) gallons per minute (gpm) per loop to 94,600 (91,400 x 1.035) gpm per loop. In the same vein, Figure 3.2-1 reflects an increase in total RCS flow rate at 100%

Thermal Power Fraction from 360,100 gpm to 378,400 gpm, and for 90% Thermal Power Fraction from 342,095 gpm to 359,480 gpm.

a TS 6.9.1.14.a, Core Operating Limits Report The following table summarizes the changes to the Core Operating Limits Report (COLR) reference listing:

TS SECTION 6.9.1.14.a - CORE OPERATING LIMITS REPORT (COLR) REFERENCE LIST Current SQN TS Proposed Proposed SQN TS Basis for Basis for TS 6.9.1.14.a COLR Prop 6.9.1.14.a COLR Reference COLR Reference list Action List Limits change

1. BAW-10180P-A, keep

1.

BAW-10180-A, Revision MTC Document NEMO - Nodal 1, NEMO - Nodal remains Expansion Method Expansion Method applicable Optimized Optimized, March 1993

2. BAW-10169P-A, keep

2. BAW-10169P-A, Peaking Limits Document RSG Plant Safety Revision 0, RSG Plant remains Analysis - B&W Safety Analysis - B&W applicable Safety Analysis Safety Analysis Methodology for Methodology for RSG Plants Recirculating Steam Generator Plants, October 1989

3. BAW-10163P-A, keep

3. BAW-10163P-A, AFD, fl, f2, Document Core Operating Revision 0, Core RIL, Monitor remains Limit Methodology Operating Limit Factors applicable for Westinghouse-Methodology for Designed Reactors Westinghouse-Designed PWR, June 1989

4. BAW-10168P-A

delete,

4. EMF-2328(P)(A), PWR LOCA limits, New RSG LOCA-B&W replace Small Break LOCA FQ methodology LOCA Evaluation with Evaluation Model, March (Note 1)

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TS SECTION 6.9.1.14.a - CORE OPERATING LIMITS REPORT (COLR) REFERENCE LIST Current SQN TS Proposed Proposed SON TS Basis for 6.9.1.14.a COLR 6.9.1.14.a COLR Reference COLR Bange Reference list Action List Limits change Model for RSG 2001 Plants

5. WCAP-1 0054-P-A delete

]

Historical; No Westinghouse longer Small Break ECCS aplcal to Evaluation Model current fuel Using theintecr NOTRUMP Code A

6. WCAP-10266-P-A delete Historical; No The 1981 Revision longer of Westinghouse applicable to Evaluation Model fuel in the Using BASH Code core

7. BAW-10227P-A keep

5. BAW-10227P-A, CFM and TCS Document Evaluation of Revision 1, Evaluation of Limits [f2(AI)]

remains Advanced Cladding Advanced Cladding and applicable and Structural Structural Material (M5)

Material (M5) in in PWR Reactor Fuel, PWR Reactor Fuel June 2003

8. BAW-10186-A keep

6.

BAW-10186P-A Burnup Limits Document Extended Burnup Revision 2, Extended remains Evaluation Burnup Evaluation, June applicable 2003

9. EMF-2103P-A Realistic Large Break LOCA Methodology for Pressurized Water keep

7. EMF-2103P-A, Revision 0, Realistic Large Break LOCA Methodology for Pressurized Water Reactors, April 2003 LOCA Limits Document remains applicable add

8. BAW-10241P-A, IC-DNB MAP Document Revision 1, BHTP DNB limits, AFD supports Correlation Applied with Limits, f1(AI)

COLR Limits LYNXT, July 2005.

Limits (Note 2) add

9. BAW-10199P-A, IC-DNB MAP Document Revision 0, The BWU limits, AFD supports Critical Heat Flux Limits, f1(AI)

COLR Limits Correlations, August Limits (Note 2) 1996 add

10. BAW-10189P-A, CHF IC-DNB MAP Document Testing and Analysis of limits, AFD supports the Mark-BW Fuel Limits, f1(AI)

COLR Limits Assembly Design, Limits (Note 2)

January 1996 add

11. BAW-10159P-A, IC-DNB MAP Document BWCMV Correlation of limits, AFD supports Critical Heat Flux in Limits, f1(AI)

COLR Limits Mixing Vane Grid Fuel Limits (Note 2)

Assemblies, August 1990 add

12. BAW-10231(P)(A),

CFM and TCS Use of most Revision 1, COPERNIC Limits, f2(AI) up-to-date Fuel Rod Design fuel rod Computer Code, performance January 2004 code (Note 3)

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Notes to COLR Reference List Table:

Note 1:

The added topical report replaces COLR reference 4. (BAW-10168P-A Rev. 3, RSG LOCA - BWNT LOCA EM for Recirculating Water Steam Generator Plants) with EMF-2328(P)(A), PWR Small Break LOCA Evaluation Mode. The S-RELAP5 implementation is consistent with code strategy for non-Babcock & Wilcox plants, and continues the use of approved methodology including RODEX2A.

Note 2:

The applicable and previously approved Critical Heat Flux reports are added to provide a more comprehensive COLR reference list.

Note 3:

The added topical report presents the approved fuel performance method that models degradation of thermal conductivity vs. burnup for U0 2 fuel.

Mark-ups of the affected TS pages are provided in Attachment 1.

Corresponding changes will also be made to the TS Bases. Mark-ups of these proposed changes are provided in Attachment 2.

3.0 TECHNICAL EVALUATION

Analyses and evaluations of the change to AREVA Advanced W1 7 HTP fuel are described in Attachment 5. These analyses and evaluations address reactor core designs with AREVA W1 7 HTP fuel only, and mixed cores with AREVA Mark-BW fuel and AREVA Advanced W17 HTP fuel. The discussions in Attachment 5 include an overview of mechanical design features, neutronics, thermal hydraulics, and accident analyses. A summary of the small break loss-of-coolant accident (LOCA) evaluation is provided in Attachment 6 and a summary of the realistic large break LOCA is provided in.

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements As discussed in the SQN Updated Safety Analysis Report (UFSAR), SQN, Units 1 and 2, were designed and constructed to the Proposed General Design Criteria (GDC) for Nuclear Power Plant Construction Permits published in July 1967. The SQN, Units 1 and 2 construction permits were issued in May 1970. However, the SQN UFSAR addresses the NRC GDC published as Appendix A to 10 CFR 50 in July 1971. The NRC used the "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition" (NUREG 75-087, later re-designated NUREG-0800) in the original license review for SQN. The following GDCs provide guidance for assessing fuel transition.

GDC 10, which requires that the reactor core and associated coolant, control, and protection systems be designed with appropriate margin to assure that specified acceptable fuel design limits are not to be exceeded during any condition of normal operation, including the effects of anticipated operational occurrences.

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GDC 11, which requires that the reactor core be designed so that the net effect of the prompt inherent nuclear feedback characteristics tends to compensate for a rapid increase in reactivity.

GDC 12, which requires that the reactor core be designed to assure that power oscillations, which can result in conditions exceeding specified acceptable fuel design limits, are not possible or can be reliably and readily detected and suppressed.

GDC 20, which requires that the reactor core be designed to initiate the reactivity control systems automatically to assure that acceptable fuel design limits are not exceeded as a result of anticipated operational occurrences and to sense accident conditions and to initiate operation of systems and components important to safety.

GDC 25, which requires that the protection system be designed to assure that specified acceptable fuel design limits are not exceeded for any single malfunction of the reactivity control systems.

GDC 26, which requires that two independent reactivity control systems be provided.

One system shall be capable of reliable controlling reactivity changes to assure that under conditions of normal operation, including anticipated operational occurrences, and with appropriate margin for malfunctions, specified acceptable fuel design limits are not exceeded. The other system shall be capable of reliably controlling the rate of reactivity changes from planned, normal power changes to assure that acceptable fuel design limits are not exceeded. One of the systems shall be capable of holding the reactor core subcritical under cold conditions.

GDC 27, which requires that the reactivity control systems be designed to have a combined capability, in conjunction with poison addition by the Emergency Core Cooling System, of reliably controlling reactivity changes under postulated accident conditions, with appropriate margin for stuck rods, to assure the capability to cool the core is maintained.

GDC 28, which requires that the reactivity control systems be designed to assure that the effects of postulated reactivity accidents can neither result in damage to the reactor coolant pressure boundary greater than limited local yielding, nor sufficiently disturb the core, its support structures, or other reactor vessel internals so as to significantly impair the capability to cool the core.

GDC 35, which requires that a system be provided to provide abundant emergency core cooling. The system safety function shall be to transfer heat from the reactor core following any loss of reactor coolant at a rate such that fuel and clad damage that could interfere with continued core cooling is prevented and clad metal-water reaction is limit to negligible amounts.

The following NUREG-0800 Standard Review Plan (SRP) sections have been applied by NRC in assessing similar fuel design and utilization changes.

SRP 4.2, "Fuel System Design" SRP 4.3, "Nuclear Design" SRP 4.4, "Thermal and Hydraulic Design" E-8

SRP 6.3, "Emergency Core Cooling Systems" SRP 15, "Accident Analyses," various sections, as applicable provides specific evaluations which address these specific GDCs and SRP sections.

4.2 Precedent Although similar AREVA fuel designs are licensed for other Westinghouse plants, there is no precedent that covers all aspects of the fuel design proposed for this request.

4.3 Significant Hazards Determination The proposed amendment for Sequoyah Nuclear Plant (SQN), Units 1 and 2 requests changes to the Technical Specifications which support a change in fuel type from AREVA Mark-BW fuel to AREVA Advanced W17 High Thermal Performance (HTP) fuel.

The design criteria for the new fuel (AREVA Advanced W1 7 HTP) are consistent with those for the existing fuel and ensure that the new fuel is compatible with the SQN, Units 1 and 2, reactors and the existing fuel on the basis of coolant flow and neutronic characteristics as well as departure from nucleate boiling (DNB) and peak cladding temperature requirements. The new fuel design ensures mechanical compatibility with the existing fuel, reactor core, control rods, steam supply system, and fuel handling system. The following Technical Specification changes are requested to support the fuel transition. Technical Specification 2.1.1 (Safety Limits) is revised to include Departure from Nuclear Boiling Ratio (DNBR) limits, maximum local fuel centerline temperature limits, and revised Core Safety Limit (CSL) lines for the AREVA Advanced W17 HTP fuel. Technical Specification Table 2.2-1 and Technical Specification 3.2.2, Table 3.2-1 are revised to reflect increase in per loop and full core design limit flow values. Technical Specification 6.9.1.14.a is revised to update the Core Operating Limits Report (Core Operating Limits Report) reference list.

AREVA Advanced W17 HTP fuel uses M5TM cladding material. This cladding material is the same material used for the AREVA Mark-BW fuel, and does not require any change to the Technical Specifications. The list of methods used in the development of the core operating limits is updated to add methods to support the AREVA fuel analysis for the Advanced W17 HTP fuel.

The Tennessee Valley Authority (TVA) has concluded that the change to the SQN, Units 1 and 2 Operating Licenses transition to AREVA Advanced W17 HTP fuel in accordance with this proposed change does not involve a significant hazards consideration. TVA's conclusion is based on its evaluation in accordance with 10 CFR 50.91 (a)(1) of 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 reactor fuel and the analyses associated with it are not accident initiators. The response of the fuel to an accident is analyzed using conservative techniques and E-9

the results are compared to approved acceptance criteria. These evaluation results will show that the fuel response to an accident is within approved acceptance criteria for cores loaded with the new AREVA Advanced W17 HTP fuel and cores loaded with both AREVA Advanced W17 HTP and AREVA Mark-BW fuel.

Therefore, the change in fuel design does not affect accident or transient initiation or consequences.

The addition of limits on DNBR and maximum local fuel pin centerline temperature to Safety Limit Technical Specification 2.1.1 or the proposed change to the Safety Limit Technical Specification Figure 2.1-1 does not require any physical change to any plant system, structure, or component. Specifying DNBR and maximum local fuel pin centerline temperature and the change to the CSL lines are consistent with the Standard Review Plan (SRP) for ensuring that the fuel design limits are met.

Operations and analysis will continue to be in compliance with Nuclear Regulatory Commission (NRC) regulations. The new CSL limits will ensure DNBR and the peak fuel centerline temperature is maintained for protecting the fuel. The addition of DNBR limits or fuel pin centerline temperature limits, or changes to the CSL lines do not impact the initiation or the mitigation of an accident.

The proposed change Technical Specification Table 2.2-1 and Figure 3.2-1 are revised to present a new loop flow and total core flow design limit based on the new AREVA Advanced W17 HTP fuel and the new steam generators (now installed for SQN Unit 1 and that will be installed concurrently with the introduction of the new Advanced W1 7 HTP fuel for SQN Unit 2). Core flow is not an accident initiator and does not play a role in accident mitigation.

The core operating limits to be developed using the new methodologies will be established in accordance with the applicable limitations as documented in the appropriate NRC Safety Evaluation reports. The proposed change to add and remove various topical reports cited in Technical Specification 6.9.1.14.a (including adding revision numbers and revision dates to current cited topical reports) enables the use of appropriate methodologies to re-analyze certain events. The proposed methodologies will ensure that the plant continues to meet applicable design criteria and safety analysis acceptance criteria. The proposed change to the list of NRC-approved methodologies listed in Technical Specification 6.9.1.14.a is administrative in nature and has no impact on any plant configuration or system performance relied upon to mitigate the consequences of an accident. The proposed change will update the listing of NRC-approved methodologies consistent with the transition to AREVA Advanced W17 HTP fuel. Changes to the calculated core operating limits may only be made using NRC-approved methods, must be consistent with all applicable safety analysis limits and are controlled by the 10 CFR 50.59 process. The list of methodologies in the Technical Specifications does not impact either the initiation of an accident or the mitigation of its consequences.

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

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2.

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

Response: No.

Use of AREVA Advanced W1 7 HTP fuel in the SQN, Units 1 and 2, reactor cores does not adversely affect any fission product barrier, nor does it alter the safety function of safety systems, structures, or components, or their roles in accident prevention or mitigation. The operational characteristics of AREVA Advanced W1 7 HTP fuel are bounded by the safety analyses. The AREVA Advanced W1 7 HTP fuel design performs within fuel design limits and does not create the possibility of a new or different type of accident.

The addition of limits on DNBR and maximum local fuel pin centerline temperature to Safety Limit Technical Specification 2.1.1 or the proposed change to the Safety Limit Technical Specification Figure 2.1-1 does not require any physical change to any plant system, structure, or component. Specifying DNBR and maximum local fuel pin centerline temperature and the change to the CSL lines are consistent with the SRP for ensuring that the fuel design limits are met. Operations and analysis will continue to be in compliance with NRC regulations. The new CSL limits will ensure DNBR and the peak fuel centerline temperature is maintained for protecting the fuel. The addition of DNBR limits or fuel pin centerline temperature limits, or changes to the CSL lines do not affect any accident initiators that would create a new accident.

The proposed change Technical Specification Table 2.2-1 and Figure 3.2-1 are revised to present a new loop flow and total core flow design limit based on the new AREVA Advanced W17 HTP fuel and the new steam generators (now installed for SQN, Unit 1, and that will be installed concurrently with the introduction of the new Advanced W17 HTP fuel for SQN, Unit 2). Core flow is not an accident initiator and does not play a role in accident mitigation and cannot create the possibility of a new or different kind of accident.

The proposed change to the list of topical reports used to determine the core operating limits is administrative in nature and has no impact on any plant configuration or on system performance. It updates the list of NRC-approved topical reports used to develop the core operating limits. There is no change to the parameters within which the plant is normally operated. The possibility of a new or different accident is not created.

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

3.

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

Response: No.

Use of AREVA Advanced W1 7 HTP fuel does not adversely affect any fission product barrier, nor does it alter the safety function of safety systems, structures, or components, or their roles in accident prevention or mitigation. The operational characteristics of AREVA Advanced W17 HTP fuel are bounded by the safety E-11

analyses. The AREVA Advanced W17 HTP fuel design performs within fuel design limits. The proposed changes do not result in exceeding design basis limits.

Therefore, the licensed safety margins are maintained.

The addition of limits on DNBR and maximum local fuel pin centerline temperature to Safety Limit Technical Specification 2.1.1 or the proposed change to the Safety Limit Technical Specification Figure 2.1-1 does not require any physical change to any plant system, structure, or component. Specifying DNBR and maximum local fuel pin centerline temperature and the change to the CSL lines are consistent with the SRP for ensuring that the fuel design limits are met. Operations and analysis will continue to be in compliance with NRC regulations. The new CSL limits will ensure DNBR and the peak fuel centerline temperature is maintained for protecting the fuel. The addition of DNBR limits or fuel pin centerline temperature limits, or changes to the CSL lines do not impact licensed safety margins.

The proposed change Technical Specification Table 2.2-1 and Figure 3.2-1 are revised to present a new loop flow and total core flow design limit based on the new AREVA Advanced W17 HTP fuel and the new steam generators (now installed for SQN Unit 1 and that will be installed concurrently with the introduction of the new Advanced W17 HTP fuel for SQN Unit 2). The proposed changes to core flow are provided to ensure licensed safety margins are maintained.

The proposed change to the list of topical reports in Technical Specification 6.9.1.14.a does not amend the cycle specific parameters presently required by the Technical Specifications. The individual Technical Specifications continue to require operation of the plant within the bounds of the limits specified in the COLR.

The proposed change to the list of analytical methods referenced in the COLR is administrative in nature and does not impact the margin of safety.

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

Based on the above, TVA concludes that the proposed amendments do not involve a 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.

4.4 Conclusion 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) issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

E-12

5.0 ENVIRONMENTAL CONSIDERATION

A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR Part 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or a significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or curiulative 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 assessment need be prepared in connection with the proposed amendment.

6.0 REFERENCES

1. BAW-1 0220P, Mark-BW Fuel Assembly Application for Sequoyah Nuclear Units 1 & 2

2. NRC Letter to TVA, Issuance of Technical Specification Amendments for the Sequoyah Nuclear Plant, Units 1 and 2 (TAC Nos. M95144 and M95145) (TS 96-01),

dated April 21, 1997

3. BAW-10241P-A, Revision 1, BHTP DNB Correlation Applied with LYNXT, July 2005

4. BAW-10159P-A, BWCMV Correlation of Critical Heat Flux in Mixing Vane Grid Fuel Assemblies, July 1990

5. BAW-10189P-A, CHF Testing and Analysis of the Mark-BW Fuel Assembly Design, January 1996

6. BAW-10199P-A, The BWU Critical Heat Flux Correlations, August 1996

7. BAW-10231P-A, Revision 1, COPERNIC Fuel Rod Design Computer Code, January 2004

8. BAW-10162-A, TACO3, Fuel Pin Thermal Analysis Computer Code, November 1989 E-13

Figure 1 SQN Units 1 and 2 Core Safety Limit Line Comparison Current Mark-- BW versus Proposed Advanced W17 HTP Fuel 680 660 2400 psia UNACCEPTABLE OPERATION 640-,

I.

620-I--

600 580 560-'

ACCEPTABLE OPERATION

-R, Revised Core Safety Limit (Adv W17 HTP)

Existing Core Safety Limit (TS Figure 2.1-1) 540 0.0 0.2 0.4 0.6 0.8 1.0 1.2 FRACTION OF RATED THERMAL POWER E-14

ATTACHMENT 1 Proposed TS Changes (Mark-Ups) for SQN, Units 1 and 2

Proposed TS Changes (Mark-Ups) for SQN, Unit I

2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS REACTOR CORE 2.1.1 The combination of THERMAL POWER, pressurizer pressure, and the highest operating loop coolant temperature (Tavg) shall not exceed the limits shown in Figure 2.1-1.

-L

ýand the following SLs shall not be exceeded:

APPLICABILITY: MODES 1 an ACTION:

lnsert SL 2.1.1.1

""nnnnAnrAA 4R-p nolni nonnon R1" Inný rnmuntion o1 Inne ninrnm; ennnaReIln l900 aV6FaQ@ !@FmR1Qr9FaWF@ 90 J 1-r1EXPAR WuWtRME Rag-8Mcooaa !R8 aBIRFERoeRIE) I3roccurIZ8r BFrocc Fe "Re, b in HOT STANDBY within 7\\

1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

I If SL 2.1.1 is violated, restore compliance and REACTOR COOLANT SYSTEM PRESSURE 2.1.2 The Reactor Coolant System pressure shall not exceed 2735 psig.

APPLICABILITY: MODES 1, 2, 3,4 and 5.

ACTION:

MODES 1 and 2 Whenever the Reactor Coolant System pressure has exceeded 2735 psig, be in HOT STANDBY with the Reactor Coolant System pressure within its limit within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

MODES 3, 4 and 5 Whenever the Reactor Coolant System pressure has exceeded 2735 psig, reduce the Reactor Coolant System pressure to within its limit within 5 minutes.

Septmbcr 3, 18* 8 SEQUOYAH - UNIT 1 2-1 Amendment No. 41

Insert SL 2.1.1.1 2.1.1.1 The departure from nucleate boiling ratio (DNBR) shall be maintained > 1.132 for the BHTP correlation, > 1.21 for the BWU-N correlation, and > 1.21 for the BWCMV correlation.

2.1.1.2 The maximum local fuel pin centerline temperature shall be maintained < 4901'F, decreasing by 13.7°F per 10,000 MWD/MTU of burnup for COPERNIC applications, and < 4642°F, decreasing by 58°F per 10,000 MWD/MTU of burnup for TACO3 applications.

.1 I

t+E SEQUOYAH UNIT 1

+

Replace Figure with Insert SQN U1 Figure 2.1-1

=U

F; U

N Aendmber 23, 1982 Ameudment 1.9 M

~2-2

• I * * * +.1 II' m

m*

Insert SQN U1 Figure 2.1-1 68OA

~~2400 psia-640 1C485 psi.a 50ACCEPTABLE_____

OPERA7ION 0.o 0.0 0-2 0.4 0.8 0-8 1.2 FRACTION OF RATED THERMAL POWER

TABLE 2.2-1 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT

1. Manual Reactor Trip

2. Power Range Neutron Flux

3. Power Range Neutron Flux High Positive Rate

4. Power Range Neutron Flux, High Negative Rate

5.

Intermediate Range, Neutron Flux

6.

Source Range Neutron Flux

7. Overtemperature AT

8. Overpower AT

9. Pressurizer Pressure--Low

10. Pressurizer Pressure--High

11. Pressurizer Water Level-High

12. Loss of Flow NOMINAL TRIP SETPOINT Not Applicable Low Setpoint - 25% of RATED THERMAL POWER High Setpoint - 109% of RATED THERMAL POWER 5% of RATED THERMAL POWER with a time constant _> 2 second 5% of RATED THERMAL POWER with a time constant > 2 second 25% of RATED THERMAL POWER 105 counts per second See Note 1 See Note 2 1970 psig 2385 psig 92% of instrument span 90% of design flow per loop*

ALLOWABLE VALUES Not Applicable Low Setpoint - < 27.4% of RATED THERMAL POWER High Setpoint - < 111.4% of RATED I THERMAL POWER

< 6.3% of RATED THERMAL POWER with a time constant _2 second

• < 6.3% of RATED THERMAL POWER with a time constant _2 second

• 45.20% of RATED THERMAL POWER

< 1.45 x 105 counts per second See Note 3 See Note 4

_ 1964.8 psig

< 2390.2 psig

< 92.7% of instrument span I

_> 89.6% of design flow per loop*

I

• Design flow 9i0s945(8, X 1.035) gpm per loop.

94,60091,400 Sept4mbc1 13, 2006 Amendment No. 44, 141, 185, 221, 223, 310 SEQUOYAH - UNIT 1 2-5

Figure 3.2-1 Flow vs. Power for 4 Loops in Operation 1380000 1 1378000

_]3740001!

'L_

368000 F362000 1360000 Azze ~

~-.e :

/

F-378400

/

/

C.7 94 Z--F-98 G

Th-,,ermal Power Fra&t;cn (In of Ri?)

April e

?1 1997 A~endmer,- No. 223 STEQ"IY.AZ -

UNIT 3/4 2-1-7

ADMINISTRATIVE CONTROLS MONTHLY REACTOR OPERATING REPORT 6.9.1.10 DELETED.

CORE OPERATING LIMITS REPORT 6.9.1.14 Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT before each reload cycle or any remaining part of a reload cycle for the following:

1.

f1(AI) limits for Overtemperature Delta T Trip Setpoints and f2(AI) limits for Overpower Delta T Trip Setpoints for Specification 2.2.1.

2.

Moderator Temperature Coefficient BOL and EOL limits and 300 ppm surveillance limit for Specification 3/4.1.1.3,

3.

Shutdown Bank Insertion Limit for Specification 3/4.1.3.5,

4.

Control Bank Insertion Limits for Specification 3/4.1.3.6,

5.

AXIAL FLUX DIFFERENCE Limits for Specification 3/4.2.1,

6.

Heat Flux Hot Channel Factor and K(z) for Specification 3/4.2.2, and

7.

Nuclear Enthalpy Rise Hot Channel Factor for Specification 3/4.2.3. __

6.9.1.14.a The analytical methods u Revision 1, rmine the core operating limits s all be those previously reviewed and approved by NRC, pe se descr lowing d uments:

O o 18 o n u Ments I

The COLR will contain the com ete identification for ach of the TS reference topical reports used to prepare the COLR (i.e., report um er Revision 0, ion, date, and any suppleme ts).

1.

BAW-1 01 80P-A, MO - Nodal Expa sion Method Optimized June

2.

BAW-10169P-A, SG Plant Safety lysis - B&W Safety Analysis Methodology for Revision 0, Recirculating Steam Generator Plants

3.

BAW-1 01631

, "Core Operating Limit Methodology for Westinghouse-Designed PWRs'

4.

2AW-! 01SOP.A, "RSG LOC.A. -_R&W Less of Coolant Accident Evaluatin M for

\\rRecirculating Steam Generater P!antc" EMF-2328 (P)(A), "PWR Small Break LOCA Evaluation Model," March 2001 SEQUOYAH - UNIT 1 6-13 Amendment No. 52, 58, 72, 74, 117, 152, 155, 156, 171,216, 223, 281,300, 314

ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT (continued)

8.

IWCAI 10541 P A, "Weotingh*u*

9 Small Broak EQCS REalutio Modoll UingI I

th 6-ICAPGA 1 026 P A, "-Tho 1081 Rovicion Of Wertinghousc; EvAluAtion Modol Ucing BASH Q---

Revision 1, June 2003

[

7-.

BAW-10227P-A v

e vanced Cladding and Structural Material (M5) in PWR R e a cto r F u e l" R e. visio n.2,J une2 0 3 8-7 BAW-10186-A, "Extonri

,rnup Evaluation" S,-:Revisio 0....

r pI20 IM 9.

EMF-21_03P-A,PReaNi ea LOCA Methodology for Pressurized Water Reactors"Reactors"--

Insert 6.9.1.14.a II 6.9.1.14.b The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met.

6.9.1.14.c THE CORE OPERATING LIMITS REPORT shall be provided within 30 days after cycle start-up (Mode 2) for each reload cycle or within 30 days of issuance of any midcycle revision of the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.

REACTOR COOLANT SYSTEM (RCS) PRESSURE AND TEMPERATURE LIMITS (PTLR) REPORT 6.9.1.15 RCS pressure and temperature limits for heatup, cooldown, low temperature operation, criticality, and hydrostatic testing, LTOP arming, and PORV lift settings as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:

Specification 3.4.9.1, "RCS Pressure and Temperature (P/T) Limits" Specification 3.4.12, "Low Temperature Over Pressure Protection (LTOP) System" 6.9.1.15.a The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

1.

Westinghouse Topical Report WCAP-14040-NP-A, "Methodology used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves."

2.

Westinghouse Topical Report WCAP-1 5293, "Sequoyah Unit 1 Heatup and Cooldown Limit Curves for Normal Operation and PTLR Support Documentation."

3.

Westinghouse Topical Report WCAP-1 5984, "Reactor Vessel Closure HeadNessel Flange Requirements Evaluation for Sequoyah Units 1 and 2."

6.9.1.15.b The PTLR shall be provided to the NRC within 30 days of issuance of any revision or supplement thereto.

STEAM GENERATOR TUBE INSPECTION REPORT 6.9.1.16 A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with the Specification 6.8.4.k, Steam Generator (SG) Program. The report shall include:

geptcmber 24, 2008 SEQUOYAH - UNIT 1 6-13a Amendment No. 52, 58, 72, 74, 117,155,223,241,258,294,297, 306, 314, 320

Insert 6.9.1.14.a

8.

BAW-1 0241 P-A, Revision 1, "BHTP DNB Correlation Applied with LYNXT," July 2005

9.

BAW-10199P-A, Revision 0, "The BWU Critical Heat Flux Correlations," August 1996

10.

BAW-10189P-A, "CHF Testing and Analysis of the Mark-BW Fuel Assembly Design," January 1996

11.

BAW-10159P-A, "BWCMV Correlation of Critical Heat Flux in Mixing Vane Grid Fuel Assemblies," August 1990

12.

BAW-10231(P)(A), Revision 1, "COPERNIC Fuel Rod Design Computer Code,"

January 2004

Proposed TS Changes (Mark-Ups) for SQN, Unit 2

2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS REACTOR CORE 2.1.1 The combination of THERMAL POWER, pressurizer pressure, and the highest operating loop coolant temperature (Tavg) shall not exceed the limits shown in Figure 2.1-1.

Aand the following SLs shall not be

~exceeded:

APPLICABILITY: MODES 1 and 2. "

• • _*S ACTION:

F.........

Whone;8R@V trio ROWn d(Rieir hV tfle Gcombn-aion o t~f th hnioflort o~ntr;at0o loop ;averaga ;toMter-atur.A and dod the appro-Drte presuri-'r pressre line, be in HOT STANDBY within r-1tr~M;~L r~r~

nat~ ~ixcou 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

[I SL 21.1 is violated, restore compliance and L/

REACTOR COOLANT SYSTEM PRESSURE 2.1.2 The Reactor Coolant System pressure shall not exceed 2735 psig.

APPLICABILITY: MODES 1, 2, 3, 4 and 5.

ACTION:

MODES 1 and 2 Whenever the Reactor Coolant System pressure has exceeded 2735 psig, be in HOT STANDBY with the Reactor Coolant System pressure within its limit within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

MODES 3, 4 and 5 Whenever the Reactor Coolant System pressure has exceeded 2735 psig, reduce the Reactor Coolant System pressure to within its limit within 5 minutes.

SEQUOYAH -UNIT 2 2-1 Septcmber 3, 1985 Amendment No. 33

Insert SL 2.1.1.1 2.1.1.1 The departure from nucleate boiling ratio (DNBR) shall be maintained > 1.132 for the BHTP correlation, > 1.21 for the BWU-N correlation, and > 1.21 for the BWCMV correlation.

2.1.1.2 The maximum local fuel pin centerline temperature shall be maintained < 4901'F, decreasing by 13.7°F per 10,000 MWD/MTU of burnup for COPERNIC applications, and < 4642°F, decreasing by 58°F per 10,000 MWD/MTU of burnup for TACO3 applications.

Azuencimexit No. 21.

Insert SQN U2 Figure 2.1-1 680-AC L

U_

_UNACCEPTABLE OOPERATION 2400 ;}sia1 0340 2.1n3 psi a I

O21D 8

pi f1ol 1.

d. --..

4 0

.00 0.4 0.8 0A*

i.

1.2 FRACTION OF RATED THERMAL POWER

TABLE 2.2-1 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT

1. Manual Reactor Trip

2.

Power Range, Neutron Flux

3.

Power Range, Neutron Flux, High Positive Rate

4.

Power Range, Neutron Flux, High Negative Rate

5.

Intermediate Range, Neutron Flux

6.

Source Range, Neutron Flux

7.

Overtemperature AT

8. Overpower AT

9. Pressurizer Pressure--Low

10. Pressurizer Pressure--High

11. Pressurizer Water Level--

High

12. Loss of Flow NOMINAL TRIP SETPOINT Not Applicable Low Setpoint - 25% of RATED THERMAL POWER High Setpoint - 109% of RATED THERMAL POWER 5% of RATED THERMAL POWER with a time constant

>_2 seconds 5% of RATED THERMAL POWER with a time constant

>2 seconds 25% of RATED THERMAL POWER 105 counts per second See Note 1 See Note 2 1970 psig 2385 psig 92% of instrument span 90% of design flow per loop*

ALLOWABLE VALUES Not Applicable Low Setpoint - *27.4% of RATED THERMAL POWER High Setpoint - _< 111.4% of RATED THERMAL POWER

• 6.3% of RATED THERMAL POWER with a time constant

Ž_2 seconds

< 6.3% of RATED THERMAL POWER with a time constant

>_2 seconds

• 45.20% of RATED THERMAL I POWER

< 1.45 x 105 counts per second I See Note 3 See Note 4

Ž1964.8 psig

< 2390.2 psig

< 92.7% of instrument span I

> 89.6% of design flow per I

loop*

9o*Design i

l ow 4 (9045 x 1.035) gpm per loop.

SEQUOYAH - UNIT 2 2-5 Septomber 13, 2006 Amendment No. 36, 132, 177, 203, 212, 214, 299

Figure 3.2-1 Flow vs. Power for 4 Loops in Operation 1380000oo R214 92 94 96 98 100 90 102 Thermal Power Fraction (% of RTP)

Amendment No.

214 SEQUOYAH - UNIT 2 3/4 2-15 I

ADMINISTRATIVE CONTROLS MONTHLY REACTOR OPERATING REPORT 6.9.1.10 DELETED CORE OPERATING LIMITS REPORT 6.9.1.14 Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT before each reload cycle or any remaining part of a reload cycle for the following:

1.

f1(AI) limits for Overtemperature Delta T Trip Setpoints and f2(AI) limits for Overpower Delta T Trip Setpoints for Specification 2.2.1.

2.

Moderator Temperature Coefficient BOL and EOL limits and 300 ppm surveillance limit for Specification 3/4.1.1.3,

3.

Shutdown Bank Insertion Limit for Specification 3/4.1.3.5,

4.

Control Bank Insertion Limits for Specification 3/4.1.3.6,

5.

AXIAL FLUX DIFFERENCE Limits for Specification 3/4.2.1,

6.

Heat Flux Hot Channel Factor and K(z) for Specification 3/4.2.2, and

7.

Nuclear Enthalpy Rise Hot Channel Factor for Specification 3/4.2.3.

6.9.1.14.a The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by NRC, specificall those described in the following documents:

Revision 1 1993 The COLR will contain the complet identification for each of the TS refereLa popIcaireports used to prepare the COLR (i.e., report nu ber ion, date, and an suplements Revision 0!

October 1989

1.

BAW-1 01 80P-A, MO - Nodal Expansi Method ptimized June 1ý

2.

BAW-10169P-A, RSG Plant Safety A sis - B&W Safety Analysis Methodology for Recirculating Steam Generator Plants"

• Revision 0, I"/

3.

BAW-10163P-A, `Core perating Limit Methodology for Westinghouse-Designed PWRs"

4.

BAW 10168P A, "RSG LOCA R&W Lcc of Coolant Accident E-alua-tin.Mode! for Recirculating Steam Genorator Plant&"

EMF-2328 (P)(A), "PWR Small Break LOCA Evaluation Model,' March 2001 Novomber 16, 2006 SEQUOYAH - UNIT 2 6-13 Amendment No. 44, 50, 64, 66, 107, 134,142, 146, 161,206, 214, 223, 272, 289, 303

ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT (continued)

WUAiQAP 10954 P A, -wectngloucow_ R-Mall Bireak EGGSE Evalu ation modol Using the NMOTRUM'A)P Codo" P

I dI m m I I I I I

• A AI I vvu'riu~tt r noiuzi -~vi~on01wocinnoco vau~lonMOOIucia 11

- 7T 1n,,-

June 2003 BAW-1 0227P-A, "Ev on of Advanced Cladding and Structural Material (M5) in PWR Reactor Fuel" BAW-10186-Axtended Burnup Evaluation""-"

EMF-2Ai*Al 2003 EMFA21O3PýAýreakLEOCA Methodýology for Pressurized Water n79.

Reactors Insert 6.9 6.9.1.14.b The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met.

6.9.1.14.c THE CORE OPERATING LIMITS REPORT shall be provided within 30 days after cycle start-up (Mode 2) for each reload cycle or within 30 days of issuance of any midcycle revision of the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.

REACTOR COOLANT SYSTEM (RCS) PRESSURE AND TEMPERATURE LIMITS (PTLR) REPORT 6.9.1.15 RCS pressure and temperature limits for heatup, cooldown, low temperature operation, criticality, and hydrostatic testing, LTOP arming, and PORV lift settings as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:

Specification 3.4.9.1, "RCS Pressure and Temperature (PIT) Limits" Specification 3.4.12, "Low Temperature Over Pressure Protection (LTOP) System" 6.9.1.15.a The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

1.

Westinghouse Topical Report WCAP-14040-NP-A, "Methodology used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves."

2.

Westinghouse Topical Report WCAP-15321, "Sequoyah Unit 2 Heatup and Cooldown Limit Curves for Normal Operation and PTLR Support Documentation."

3.

Westinghouse Topical Report WCAP-1 5984, "Reactor Vessel Closure HeadNessel Flange Requirements Evaluation for Sequoyah Units 1 and 2."

6.9.1.15.b The PTLR shall be provided to the NRC within 30 days of issuance of any revision or supplement thereto.

STEAM GENERATOR (SG) TUBE INSPECTION REPORT 6.9.1.16.1 A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with Specification 6.8.4.k, "Steam Generator (SG) Program." The report shall include:

SEQUOYAH - UNIT 2 6-14 Amendment No. 44, 50, 64, 66,107, 134, 146, 206, 214, 231,249, 284, 303, 305, 311

Insert 6.9.1.14.a

8.

BAW-1 0241 P-A, Revision 1, "BHTP DNB Correlation Applied with LYNXT," July 2005

9.

BAW-10199P-A, Revision 0, "The BWU Critical Heat Flux Correlations," August 1996

10.

BAW-1 01 89P-A, "CHF Testing and Analysis of the Mark-BW Fuel Assembly Design," January 1996

11.

BAW-10159P-A, "BWCMV Correlation of Critical Heat Flux in Mixing Vane Grid Fuel Assemblies," August 1990

12.

BAW-1 0231 (P)(A), Revision 1, "COPERNIC Fuel Rod Design Computer Code,"

January 2004

ATTACHMENT 2 Proposed TS Bases Changes (Mark-Ups) for SQN, Units I and 2

Proposed TS Bases Changes (Mark-Ups) for SQN, Unit 1

cladding (due to departure from nucleate boiling) and overheating of the fuel pellet (centerline fuel melt), either of which could result in 2.1 SAFETY LIMITS BASES 2.1.1 REACTOR CORE J

I 1The restrictions of this safety limit prevent overheating of the fuel ape..

sbibIe cladding perforation W4 would result in the release of fission products to the reactor coolant. Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coeffici i& Inma qnri tha rlnrinn citrfnra tamnrnhira 4c _Hnhtlv nhnv, th, nnlint cnfiirntinn temperatu Overheating of the fuel is prevented by maintaining the steady state peak linear

" 'I' heat rate (LHR) below the level at which fuel centerline melting occurs.I Operation above the upper boundary of the nucleate boiling regime could result in excessive eeddi4g temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient DNB is not a directly measurable parameter during operation and therefore THERMAL POWER and R ctor Coolant Temperature and Pressure have been related to DNB. The DNB correlations have been deeloped to predict the DNB flux and the location of DNB for axially uniform and non-uniform heat flux d stributions. The local DNB heat flux ratio, DNBR, defined as the ratio of the heat flux that would cause NE at a particular core location to the local heat flux, indicative of the margin to DNB.

I-from the outer surface of the cladding to the reactor coolant water the statistical DNBR design limit Insert B 2.1 as THERMi POWER The DNB desion iasis. is that there mus[ L i at Measi a Fi Marceni 1ronam}ilry wa~n 4r,,.erpen confidence that DNB %To meet the DNB Design Basis, a statistical core design (SCD) process has been used to develop an appropriate statistical DNBR design limit. U 44

!-.meetki@ thk.r, 3_063.0,1` H-a-8., oneplnlsinpant operating parameters, nu16A-cla an nermal parameters, and fuel fabrication parameters are considered statistically such that there is at least a 95 percent probability at a 95 percent confidence level that the minimum DNBR for the limiting rod is greater than or equal to the DNBR limit. The...

A.n.. o in6 th. 21'4049 plant pa"ameter. are used to determine the plant uno:."tainty. This DNBR uncertaint combined with the correlation QNIR limit, establishes a doesgn D.NB9R w'.'e which must be met in pl safety analysis u values of input parameters without SCD analysis ftical heat flux a

The cu he loci of points of THERMAL %-vvFuru, rmvutu,, uulfit for

/ System pressure and average temperature for which the minimum DNBR is no less than the safety uncertainty

/ anatlsis DNBR limit, or the average enthalpy at the vessel exit is equal to the enthalpy of saturated liquid.

/4Th curves of Figure 2.1-1 are based on an enthalpy rise hot channel factor, FA'H, nominal values have been reduced to include a 4% total rod power uncertainty factor), and a rofronccc.in, "ith, a,peak* of 1.55 for fFaxial poWor chap.

An allowance is included for an increase in FH at reduced pewer based on the expression:

design axial

?

power FN -FRTPr 1 3

(-

1 poe FN =FRT [1+.3 (l--P)]

distribution, I

THERMAL POWER Fz with a where P POWER <- Fand reference R

T M

Ocosine shape is at a peak of 1.70 - A46-9 ie 1.55 MRTP

= N om in al V alu es

1.

We s-i-e P ool AN

--~.6

- W~tinhou~ FuI April 21, 1997 Amendment No. 19, 114, 138, 155, 223 SEQUOYAH - UNIT 1 B 2-4

Insert B 2.1 Operation above the boundary of the nucleate boiling regime could result in excessive cladding temperature because of the onset of DNB and the resultant sharp reduction in heat transfer coefficient. Inside the steam film, high cladding temperatures are reached, and a cladding water (zirconium water) reaction may take place this chemical reaction results in oxidation of the fuel cladding to a structurally weaker form. This weaker form may lose its integrity, resulting in an uncontrolled release of activity to the reactor coolant.

Operation above the maximum local linear heat generation rate for fuel melting could result in excessive fuel pellet temperature and cause melting of the fuel at its centerline. Fuel centerline melting occurs when the local LHR, or power peaking, in a region of the fuel is high enough to cause the fuel centerline temperature to reach the melting point of the fuel. Expansion of the pellet upon centerline melting may cause the pellet to stress the cladding to the point of failure, allowing an uncontrolled release of activity to the reactor coolant. The melting point of uranium dioxide varies slightly with burnup. As uranium is depleted and fission products produced, the net effect is a decrease in the melting point. Fuel centerline temperature is not a directly measurable parameter during operation. The maximum local fuel pin centerline temperature is maintained by limiting the local linear heat generation rate in the fuel. The local linear heat generation rate in the fuel is limited so that the maximum fuel centerline temperature will not exceed the value defined by the algorithm in SL 2.1.1.2.

SAFETY LIMITS Ifor DNB l-*

emperatuýre j o Det-(AI)

BASES

[or Delta These limiting heat flux conditions are higher than those calculat for the range of all control\\*

rods fully withdrawn to the maximum allowable control rod insertion assu g the axial power imbalanceg is within the limits of the f, (Delta I) function of the Overtemperature Delta trip. When the axial power FimbaIaRR "A not within the tolorGnc, the axial poWor imbhRARancAfoto h

vroprtr et tripe will roduco the Getpoints to proVide protoctionR cone~tonF;t9t w.ith cor)e caf*t limits.

[ertB2.1(a 2.1.2 REACTOR COOLANT SYSTEM PRESSURE The restriction of this Safety Limit protects the integrity of the Reactor Coolant System from overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere.

The reactor pressure vessel and pressurizer are designed to Section III of the ASME Code for Nuclear Power Plant which permits a maximum transient pressure of 110% (2735 psig) of design pressure. The Reactor Coolant System piping, valves and fittings, are designed to ANSI B 31.1 1967 Edition, which permits a maximum transient pressure of 120% (2985 psig) of component design pressure.

The Safety Limit of 2735 psig is therefore consistent with the design criteria and associated code requirements.

The entire Reactor Coolant System is hydro tested at 3107 psig, 125% of design pressure, to demonstrate integrity prior to initial operation.

2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS The Reactor Nominal Trip Setpoint Limits specified in Table 2.2-1 are the values at which the Reactor Trips are set for each functional unit. The Nominal Trip Setpoints have been selected to ensure that the reactor core and reactor coolant system are prevented from exceeding their safety limits during normal operation and design basis anticipated operational occurrences and to assist the Engineered Safety Features Actuation System in mitigating the consequences of accidents. Operation with a trip set less conservative than its Nominal Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Nominal Trip Setpoint and the Allowable Value is equal to or less than the rack allowance assumed for each trip in the safety analyses.

Technical specifications are required by 10 CFR 50.36 to contain Limiting Safety System Settings (LSSS) defined by the regulation as "... settings for automatic protective devices.., so chosen that automatic protective action will correct the abnormal situation before a Safety Limit (SL) is exceeded."

The analytic limit is the limit of the process variable at which a safety action is initiated, as established by the safety analysis, to ensure that a SL is not exceeded. Any automatic protection action that occurs on reaching the analytic limit therefore ensures that the SL is not exceeded. However, in practice, the actual settings for automatic protective devices must be chosen to be more conservative than the analytic limit to account for instrument loop uncertainties related to the setting at which the automatic protective action would actually occur.

September 13, 2006 SEQUOYAH - UNIT 1 B 2-2 Amendment No. 310

Insert B 2.1 (a)

When the axial power imbalance exceeds the tolerance (or deadband) of the f1(Al) trip reset function, the Overtemperature Delta Temperature trip setpoint is reduced by the values in the CORE OPERATING LIMITS REPORT to provide protection required by the core safety limits.

Similarly, the limiting linear heat generation rate conditions for CFM are higher than those calculated for the range of all control rods from fully withdrawn to the maximum allowable control rod insertion assuming the axial power imbalance or Delta-I (AI) is within the limits of the f2(AI) function of the OverPower-Delta Temperature trip. When the axial power imbalance exceeds the tolerance (or deadband) of the f2(AI) trip reset function, the OverPower-Delta Temperature trip setpoint is reduced'by the values specified in the CORE OPERATING LIMITS REPORT to provide protection required by the core safety limits.

POWER DISTRIBUTION LIMITS BASES 3/4.2.4 QUADRANT POWER TILT RATIO The QUADRANT POWER TILT RATIO limit assures that no anomaly exists such that the radial power distribution satisfies the design values used in the power capability analysis. Radial power distribution measurements are made during startup testing and periodically during power operation.

The QUADRANT POWER TILT RATIO limit at which corrective action is required provides DNB and linear heat generation protection with x-y plane power tilts. The QUADRANT POWER TILT RATIO limit is reflected by a corresponding peaking augmentation factor which is included in the generation of the AFD limits.

The 2-hour time allowance for operation with the tilt condition greater than 1.02 but less than 1.09, is provided to allow identification and correction of a dropped or misaligned control rod. In the event such action does not correct the tilt, the margin for uncertainty on FQ(X,Y,Z) is reinstated by reducing the allowable THERMAL POWER by 3 percent for each percent of tilt in excess of 1.02.

3/4.2.5 DNB PARAMETERS The limits on the DNB related parameters assure that each of the parameters are maintained within the normal steady state envelope of operation assumed in the transient and accident analyses.

The limits are consistent with the initial FSAR assumptions and have been analytically demonstrated adequate to maintain a minimum DNBR of greater than or equal to the safety analysis DNBR limit throughout each analyzed transient.

The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> periodic surveillance of these parameters through instrument readout is sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation.

I Th flow pa.amt....

.-rc nicatei in Fi-qurO 2-1 h.a.. begin rou ded down to Was the ahaly.ic in tha t'nmonn

,a ti i rnr.+inn April 21, 1997 Amendment No. 19, 138,155, 223 SEQUOYAH - UNIT I B 3/4 2-4

Proposed TS Bases Changes (Mark-Ups) for SQN, Unit 2

cladding (due to departure from nucleate boiling) and overheating of the fuel pellet (centerline fuel 2.1 SAFETY LIMITS melt), either of which could result in BASES 2.1.1 REACTOR CORE restrictions of this Safety Limit prevent overheating of the fuel a.-.dkte cladding perforation*',

-ýwould result in the release of fission products to the reactor coolant. Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficien is larae and the claddina surface temperature is sliahtlv above the coolant saturation temperature.-

Overheating of the fuel is prevented by maintaining the steady state peak linear heat rate (LHR) below the level at which fuel centerline melting occurs.I Operation Move tne upper uounuary 01 Me nucleate 0o1ng regine cuum resuin ex ie cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficien]K DNB is not a directly measurable parameter during operation and; therefore, THERMAL POWER and actor Coolant Temperature and Pressure have been related to DNB. The DNB correlations have been deeloped to predict the DNB flux and the location of DNB for axially uniform and non-uniform heat flux di tributions. The local DNB heat flux ratio, DNBR, defined as the ratio of the heat flux that would cause D B at a particular core location to the local heat flux, is indicative of the margin to DNB.

t from the outer surface of the cladding to the

/

reactor coolant water I

Th DNbag*onh si. *- that thpre mu m 21.,1.At R Ilz DRIMtMw DTY10111717'7T11,Mv wt-,,

[ 11, a

v.

confidenc To meet the DNB Design Basis, a statistical core design (SCD) process has been

,Z' used to develop an appropriate statistical DNBR design limit. UI

/¢J..

Mir aes'..g-'@.R. 19a.r..

c'.,

unceraiaties in planto operating parameters, nuclear and tnerma-parameters, and fuel fabrication parameters are considered statistically such that there is at least a 95 percent probability at a 95 percent confidence level that the minimum DNBR for the limiting rod is greater than or equal to the DNBR limit. The.uncan iMORnAR tho 9bo04 plant parametersr.

u

d. to d.trmin*

Ithe the plant un*certaity. This DNBR uncertai i¢, combined with the orrelation D4NBR limit, establishes a statistical

-which must be met i pfant safety analysis u g values of input parameters without DNBR noni

'derived from the SCD analysis applicable DNB critical 1

heat flux adjustment design The curves of Figure 2.1-1 show the loci of points of THERMAL

\\.

for limit System pressure and average temperature for which the minimum DNBR is no less than the safety u

-crtant*

analysis DNBR limit, or the average enthalpy at the vessel exit is equal to the enthalpy of saturated liquidu lInsert B (r[T1 curves of Figure 2.1-1 are based on an enthalpy rise hot channel design axial 12.1 0.7%N*

2.

factor FNAH, values have been reduced to include a 4% total rod power uncertainty factor), and a <-

power roforonco c-cino VOWth a peak 9f 1.55 for axial poWor chape. An allowance is included for an increase in distribution, Fz witha FNat reduced pewe* based on the expression:

reference FN =FRj [1+.3 (I-P)]

cosine as THERMAL I

shape at a POWER is where P =

THERMAL POWER peak of 1.55 RATED THERMAL POWER 1.70 - Ma& -

BSA, Fuel FH= Nominal Value&

442-westinig cu8z Fu18l April 21, 1997 SEQUOYAH - UNIT 2 B 2-1 Amendment No. 21,104,130, 146,214

Insert B 2.1 Operation above the boundary of the nucleate boiling regime could result in excessive cladding temperature because of the onset of DNB and the resultant sharp reduction in heat transfer coefficient. Inside the steam film, high cladding temperatures are reached, and a cladding water (zirconium water) reaction may take place this chemical reaction results in oxidation of the fuel cladding to a structurally weaker form. This weaker form may lose its integrity, resulting in an uncontrolled release of activity to the reactor coolant.

Operation above the maximum local linear heat generation rate for fuel melting could result in excessive fuel pellet temperature and cause melting of the fuel at its centerline. Fuel centerline melting occurs when the local LHR, or power peaking, in a region of the fuel is high enough to cause the fuel centerline temperature to reach the melting point of the fuel. Expansion of the pellet upon centerline melting may cause the pellet to stress the cladding to the point of failure, allowing an uncontrolled release of activity to the reactor coolant. The melting point of uranium dioxide varies slightly with burnup. As uranium is depleted and fission products produced, the net effect is a decrease in the melting point. Fuel centerline temperature is not a directly measurable parameter during operation. The maximum local fuel pin centerline temperature is maintained by limiting the local linear heat generation rate in the fuel. The local linear heat generation rate in the fuel is limited so that the maximum fuel centerline temperature will not exceed the value defined by the algorithm in SL 2.1.1.2.

SAFETY LIMITS

[for DNB I

[

BASES Ins

2. 1.1 REACTOR CORE (Continued)

These limiting heat flux conditin are higher than those calu ted for the range of all cont rods fully withdrawn to the maximum allowable control rod insertion assumi the axial power imbalance is within the limits of the fj(delta I) function of the Overtemperature Delta trip. When the axial poWvr imaaneicnt wihnthe t@o9eRae, the axial poWor imbalancc oAot On the vroprtr ot T tpeWill reduce the SetpOintG to provido protection censic;;tonRt w.ith core safety liit.Insert B 2.1 (a) 2.1.2 REACTOR COOLANT SYSTEM PRESSURE The restriction of this Safety Limit protects the integrity of the Reactor Coolant System from overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere.

The reactor pressure vessel and pressurizer are designed to Section III of the ASME Code for Nuclear Power Plant which permits a maximum transient pressure of 110% (2735 psig) of design pressure. The Reactor Coolant System piping, valves and fittings, are designed to ANSI B 31.1 1967 Edition, which permits a maximum transient pressure of 120% (2985 psig) of component design pressure.

The Safety Limit of 2735 psig is therefore consistent with the design criteria and associated code requirements.

The entire Reactor Coolant System is hydrotested at 3107 psig, 125% of design pressure, to demonstrate integrity prior to initial operation.

2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS The Reactor Nominal Trip Setpoint Limits specified in Table 2.2-1 are the values at which the Reactor Trips are set for each functional unit. The Nominal Trip Setpoints have been selected to ensure that the reactor core and reactor coolant system are prevented from exceeding their safety limits during normal operation and design basis anticipated operational occurrences and to assist the Engineered Safety Features Actuation System in mitigating the consequences of accidents. Operation with a trip set less conservative than its Nominal Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Nominal Trip Setpoint and the Allowable Value is equal to or less than the rack allowance assumed for each trip in the safety analyses.

Technical specifications are required by 10 CFR 50.36 to contain Limiting Safety System Settings (LSSS) defined by the regulation as "... settings for automatic protective devices... so chosen that automatic protective action will correct the abnormal situation before a Safety Limit (SL) is exceeded."

The analytic limit is the limit of the process variable at which a safety action is initiated, as established by the safety analysis, to ensure that a SL is not exceeded. Any automatic protection action that occurs on reaching the analytic limit therefore ensures that the SL is not exceeded. However, in practice, the actual settings for automatic protective devices must be chosen to be more conservative than the analytic limit to account for instrument loop uncertainties related to the setting at which the automatic protective action would actually occur.

The Nominal Trip Setpoint is a predetermined setting for a protective device chosen to ensure automatic actuation prior to the process variable reaching the analytic limit and thus ensuring that the SL would not be exceeded. As such, the Nominal Trip Setpoint accounts for uncertainties in setting the device (e.g., calibration), uncertainties in how the device might actually perform (e.g., repeatability),

changes in the point of action of the device over time (e.g., drift during surveillance intervals), and any other factors which may influence its actual performance (e.g., harsh accident environments). In this manner, the Nominal Trip Setpoint plays an important role in ensuring that SLs are not exceeded. As such, the Nominal Trip Setpoint meets the definition of an LSSS in accordance with Regulatory Guide 1.105, Revision 3, "Setpoints for Safety-Related Instrumentation," and could be used to meet the requirements that they be contained in the technical specifications.

September 13, 2006 SEQUOYAH - UNIT 2 B 2-2 Amendment No. 130, 146, 299

Insert B 2.1 (a)

When the axial power imbalance exceeds the tolerance (or deadband) of the f1(AI) trip reset function, the Overtemperature Delta Temperature trip setpoint is reduced by the values in the CORE OPERATING LIMITS REPORT to provide protection required by the core safety limits.

Similarly, the limiting linear heat generation rate conditions for CFM are higher than those calculated for the range of all control rods from fully withdrawn to the maximum allowable control rod insertion assuming the axial power imbalance or Delta-I (AI) is within the limits of the f2(AI) function of the OverPower-Delta Temperature trip. When the axial power imbalance exceeds the tolerance (or deadband) of the f2(AI) trip reset function, the OverPower-Delta Temperature trip setpoint is reduced'by the values specified in the CORE OPERATING LIMITS REPORT to provide protection required by the core safety limits.

POWER DISTRIBUTION LIMITS BASES 3/4.2.4 QUADRANT POWER TILT RATIO The QUADRANT POWER TILT RATIO limit assures that no anomaly exists such that the radial power distribution satisfies the design values used in the power capability analysis. Radial power distribution measurements are made during startup testing and periodically during power operation.

The QUADRANT POWER TILT RATIO limit at which corrective action is required provides DNB and linear heat generation protection with x-y plane power tilts. The QUADRANT POWER TILT RATIO limit is reflected by a corresponding peaking augmentation factor which is included in the generation of the AFD limits.

The 2-hour time allowance for operation with the tilt condition greater than 1.02 but less than 1.09, is provided to allow identification and correction of a dropped or misaligned control rod. In the event such action does not correct the tilt, the margin for uncertainty on FQ(X,Y,Z) is reinstated by reducing the allowable THERMAL POWER by 3 percent for each percent of tilt in excess of 1.02.

3/4.2.5 DNB PARAMETERS The limits on the DNB related parameters assure that each of the parameters are maintained within the normal steady state envelope of operation assumed in the transient and accident analyses.

The limits are consistent with the initial FSAR assumptions and have been analytically demonstrated adequate to maintain a minimum DNBR of greater than or equal to the safety analysis DNBR limit throughout each analyzed transient.

The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> periodic surveillance of these parameters through instrument readout is sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation.

Tho fl

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irn.+0 A

Cm, 1

4hnhn n

4--r rnh

+-. kine +knnf,~

i n

W VW W

M qtgrcp -

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rt

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• n April 21, 1997 Amendment 21, 130, 146, 214 SEQUOYAH - UNIT 2 B 3/4 2-4

ATTACHMENT 3 Proposed TS Changes (Final Typed) for SQN, Units I and 2

Proposed TS Changes (Final Typed) for SQN, Unit I

2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS REACTOR CORE 2.1.1 The combination of THERMAL POWER, pressurizer pressure, and the highest operating loop coolant temperature (Tavg) shall not exceed the limits shown in Figure 2.1-1 and the following SLs shall not be exceeded:

2.1.1.1 The departure from nucleate boiling ratio (DNBR) shall be maintained > 1.132 for the BHTP correlation, > 1.21 for the BWU-N correlation, and > 1.21 for the BWCMV correlation.

2.1.1.2 The maximum local fuel pin centerline temperature shall be maintained < 4901'F, decreasing by 13.7°F per 10,000 MWD/MTU of burnup for COPERNIC applications, and

< 4642°F, decreasing by 58°F per 10,000 MWD/MTU of burnup for TACO3 applications.

APPLICABILITY: MODES 1 and 2.

ACTION:

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

REACTOR COOLANT SYSTEM PRESSURE 2.1.2 The Reactor Coolant System pressure shall not exceed 2735 psig.

APPLICABILITY: MODES 1, 2, 3, 4 and 5.

ACTION:

MODES 1 and 2 Whenever the Reactor Coolant System pressure has exceeded 2735 psig, be in HOT STANDBY with the Reactor Coolant System pressure within its limit within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

MODES 3, 4 and 5 Whenever the Reactor Coolant System pressure has exceeded 2735 psig, reduce the Reactor Coolant System pressure to within its limit within 5 minutes.

SEQUOYAH - UNIT 1 2-1 Amendment No. 41,

Figure 2.1-1 Reactor Core Safety Limit - Four Loops in Operation 080 840 E

i-820 800 580 5401 0.0 02 04 0.6 08 1.0 1.2 FRACTION OF RATED THERMAL POWER H - UNIT 1 2-2 Amendment No. 19, SEQUOYAI

TABLE 2.2-1 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT

1. Manual Reactor Trip

2. Power Range Neutron Flux

3.

Power Range Neutron Flux High Positive Rate

4. Power Range Neutron Flux, High Negative Rate

5.

Intermediate Range, Neutron Flux

6.

Source Range Neutron Flux

7. Overtemperature AT

8. Overpower AT

9. Pressurizer Pressure--Low

10. Pressurizer Pressure--High

11. Pressurizer Water Level-High

12. Loss of Flow NOMINAL TRIP SETPOINT Not Applicable Low Setpoint - 25% of RATED THERMAL POWER High Setpoint - 109% of RATED THERMAL POWER 5% of RATED THERMAL POWER with a time constant > 2 second 5% of RATED THERMAL POWER with a time constant _ 2 second 25% of RATED THERMAL POWER 105 counts per second See Note 1 See Note 2 1970 psig 2385 psig 92% of instrument span 90% of design flow per loop*

ALLOWABLE VALUES Not Applicable Low Setpoint - < 27.4% of RATED THERMAL POWER High Setpoint - _ 111.4% of RATED THERMAL POWER

< 6.3% of RATED THERMAL POWER with a time constant > 2 second

< 6.3% of RATED THERMAL POWER with a time constant > 2 second

< 45.20% of RATED THERMAL POWER

< 1.45 x 105 counts per second See Note 3 See Note 4

> 1964.8 psig

<2390.2 psig

<92.7% of instrument span

> 89.6% of design flow per loop*

• Design flow is 94,600 (91,400 X 1.035) gpm per loop.

Amendment No. 44, 141, 185, 221, 223, 310, SEQUOYAH - UNIT 1 2-5

Figure 3.2-1 Flow vs. Power for 4 Loops in Operation 380000 IL E

E 378000 376000 374000 372000 370000 368000 D

366000 364000D-362000 360000 358000--

A 3.5% measurement uncertainty for ftw ts included in this figure.

Acceptable Operation Region Unacceptable Operation Region 92 90 94 96 98 Thermal Power Fraction (% of RTP) 100 202 Amendment No. 223, 1 SEQUOYAH - UNIT 1 3/4 2-17

ADMINISTRATIVE CONTROLS MONTHLY REACTOR OPERATING REPORT 6.9.1.10 DELETED.

CORE OPERATING LIMITS REPORT 6.9.1.14 Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT before each reload cycle or any remaining part of a reload cycle for the following:

1.

f(Al) limits for Overtemperature Delta T Trip Setpoints and f2(AI) limits for Overpower Delta T Trip Setpoints for Specification 2.2.1.

2.

Moderator Temperature Coefficient BOL and EOL limits and 300 ppm surveillance limit for Specification 3/4.1.1.3,

3.

Shutdown Bank Insertion Limit for Specification 3/4.1.3.5,

4.

Control Bank Insertion Limits for Specification 3/4.1.3.6,

5.

AXIAL FLUX DIFFERENCE Limits for Specification 3/4.2.1,

6.

Heat Flux Hot Channel Factor and K(z) for Specification 3/4.2.2, and

7.

Nuclear Enthalpy Rise Hot Channel Factor for Specification 3/4.2.3.

6.9.1.14.a The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by NRC, specifically those described in the following documents:

The COLR will contain the complete identification for each of the TS referenced topical reports used to prepare the COLR (i.e., report number, title, revision, date, and any supplements).

1.

BAW-10180P-A, Revision 1, "NEMO - Nodal Expansion Method Optimized," March 1993

2.

BAW-10169P-A, Revision 0, "RSG Plant Safety Analysis - B&W Safety Analysis Methodology for Recirculating Steam Generator Plants," October 1989

3.

BAW-10163P-A, Revision 0, "Core Operating Limit Methodology for Westinghouse Designed PWRs," June 1989

4.

EMF-2328(P)(A), "PWR Small Break LOCA Evaluation Model," March 2001 SEQUOYAH - UNIT 1 6-13 Amendment No. 52, 58, 72, 74, 117, 152, 155, 156, 171,216, 223, 281,300, 314, I

ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT (continued)

5.

BAW-1 0227P-A, Revision 1, "Evaluation of Advanced Cladding and Structural Material (M5) in PWR Reactor Fuel," June 2003

6.

BAW-10186-A, Revision 2, "Extended Burnup Evaluation," June 2003

7.

EMF-2103P-A, Revision 0, "Realistic Large Break LOCA Methodology for Pressurized Water Reactors," April 2003

8.

BAW-10241 P-A, Revision 1, "BHTP DNB Correlation Applied with LYNXT," July 2005.

9.

BAW-10199P-A, Revision 0, "The BWU Critical Heat Flux Correlations," August 1996

10.

BAW-10189P-A, "CHF Testing and Analysis of the Mark-BW Fuel Assembly Design,"

January 1996

11.

BAW-10159P-A, "BWCMV Correlation of Critical Heat Flux in Mixing Vane Grid Fuel Assemblies," August 1990

12.

BAW-10231(P)(A), Revision 1, "COPERNIC Fuel Rod Design Computer Code,"

January 2004 6.9.1.14.b The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met.

6.9.1.14.c THE CORE OPERATING LIMITS REPORT shall be provided within 30 days after cycle start-up (Mode 2) for each reload cycle or within 30 days of issuance of any midcycle revision of the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.

REACTOR COOLANT SYSTEM (RCS) PRESSURE AND TEMPERATURE LIMITS (PTLR) REPORT 6.9.1.15 RCS pressure and temperature limits for heatup, cooldown, low temperature operation, criticality, and hydrostatic testing, LTOP arming, and PORV lift settings as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:

Specification 3.4.9.1, "RCS Pressure and Temperature (P/T) Limits" Specification 3.4.12, "Low Temperature Over Pressure Protection (LTOP) System" 6.9.1.15.a The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

1.

Westinghouse Topical Report WCAP-14040-NP-A, "Methodology used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves."

2.

Westinghouse Topical Report WCAP-1 5293, "Sequoyah Unit 1 Heatup and Cooldown Limit Curves for Normal Operation and PTLR Support Documentation."

3.

Westinghouse Topical Report WCAP-15984, "Reactor Vessel Closure HeadNessel Flange Requirements Evaluation for Sequoyah Units 1 and 2."

6.9.1.15.b The PTLR shall be provided to the NRC within 30 days of issuance of any revision or supplement thereto.

SEQUOYAH - UNIT 1 6-13a Amendment No. 52, 58, 72, 74, 117,155,223,241,258,294,297,306,314,320,

ADMINISTRATIVE CONTROLS STEAM GENERATOR TUBE INSPECTION REPORT 6.9.1.16 A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with the Specification 6.8.4.k, Steam Generator (SG) Program. The report shall include:

a. The scope of inspections performed on each SG,

b. Active degradation mechanisms found,

c.

Nondestructive examination techniques utilized for each degradation mechanism,

d.

Location, orientation (if linear), and measured sizes (if available) of service induced indications,

e. Number of tubes plugged during the inspection outage for each active degradation mechanism,

f.

Total number and percentage of tubes plugged to date,

g.

The results of condition monitoring, including the results of tube pulls and in-situ testing, and

h.

The effective plugging percentage for all plugging in each SG.

SPECIAL REPORTS 6.9.2.1 Special reports shall be submitted within the time period specified for each report, in accordance with 10 CFR 50.4.

6.9.2.2 This specification has been deleted.

6.10 RECORD RETENTION (DELETED)

SEQUOYAH - UNIT 1 6-14 Amendment No. 42, 52, 58, 72, 74, 117, 148, 155, 163, 174,178, 223, 233, 241,258, 294, 297, 306,

Proposed TS Changes (Final Typed) for SQN, Unit 2

2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS REACTOR CORE 2.1.1 The combination of THERMAL POWER, pressurizer pressure, and the highest operating loop coolant temperature (Tavg) shall not exceed the limits shown in Figure 2.1-1 and the following SLs shall not be exceeded:

2.1.1.1 The departure from nucleate boiling ratio (DNBR) shall be maintained > 1.132 for the BHTP correlation, > 1.21 for the BWU-N correlation, and > 1.21 for the BWCMV correlation.

2.1.1.2 The maximum local fuel pin centerline temperature shall be maintained < 4901'F, decreasing by 13.7°F per 10,000 MWD/MTU of burnup for COPERNIC applications, and

< 4642°F, decreasing by 58°F per 10,000 MWD/MTU of burnup for TACO3 applications.

APPLICABILITY: MODES 1 and 2.

ACTION:

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

REACTOR COOLANT SYSTEM PRESSURE 2.1.2 The Reactor Coolant System pressure shall not exceed 2735 psig.

APPLICABILITY: MODES 1, 2, 3, 4 and 5.

ACTION:

MODES 1 and 2 Whenever the Reactor Coolant System pressure has exceeded 2735 psig, be in HOT STANDBY with the Reactor Coolant System pressure within its limit within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

MODES 3, 4 and 5 Whenever the Reactor Coolant System pressure has exceeded 2735 psig, reduce the Reactor Coolant System pressure to within its limit within 5 minutes.

SEQUOYAH - UNIT 2 2-1 Amendment No. 33, 1

Figure 2.1-1 Reactor Core Safety Limit-Four Loops in Operation 640 I,-

820 5,80 500 E40 0.0 0-2 0.4 0.6 0.8 FRACTION OF RATED THERMAL POWER 1.0 Amendment No. 21, SEQUOYAH - UNIT 2 2-2

TABLE 2.2-1 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT

1. Manual Reactor Trip

2. Power Range, Neutron Flux

3.

Power Range, Neutron Flux, High Positive Rate

4. Power Range, Neutron Flux, High Negative Rate

5.

Intermediate Range, Neutron Flux

6.

Source Range, Neutron Flux

7.

Overtemperature AT

8.

Overpower AT

9.

Pressurizer Pressure--Low

10. Pressurizer Pressure--High

11. Pressurizer Water Level--

High

12. Loss of Flow NOMINAL TRIP SETPOINT Not Applicable Low Setpoint - 25% of RATED THERMAL POWER High Setpoint - 109% of RATED THERMAL POWER 5% of RATED THERMAL POWER with a time constant

_>2 seconds 5% of RATED THERMAL POWER with a time constant

Ž_2 seconds 25% of RATED THERMAL POWER 105 counts per second See Note 1 See Note 2 1970 psig 2385 psig 92% of instrument span 90% of design flow per loop*

ALLOWABLE VALUES Not Applicable Low Setpoint - __ 27.4% of RATED THERMAL POWER High Setpoint - < 111.4% of RATED THERMAL POWER

• 6.3% of RATED THERMAL POWER with a time constant

Ž2 seconds

• 6.3% of RATED THERMAL POWER with a time constant

Ž_2 seconds

<_ 45.20% of RATED THERMAL POWER

• 1.45 x 105 counts per second See Note 3 See Note 4

> 1964.8 psig

< 2390.2 psig

< 92.7% of instrument span

_ 89.6% of design flow per loop*

• Design flow is 94,600 (91,400 x 1.035) gpm per loop.

SEQUOYAH - UNIT 2 2-5 I

Amendment No. 36,132,177, 203, 212, 214,299,

Figure 3.2-1 Flow vs. Power for 4 Loops in Operation 380000 T.

MO CI E

378000 37600O -

374000 372000 370000 368000 366000 362000 360000 358000 90 A 3.59X measurement uncertain~ty for fkow is included in' Zhis figure.

Acceptable Operation Regkon Unacceptable Operation Region (90,359480) 92 100 102 94 96 98 Thermal Power Fraction (% of RTP)

Amendment No. 214, 1

SEQUOYAH - UNIT 2 3/4 2-15

ADMINISTRATIVE CONTROLS MONTHLY REACTOR OPERATING REPORT 6.9.1.10 DELETED CORE OPERATING LIMITS REPORT 6.9.1.14 Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT before each reload cycle or any remaining part of a reload cycle for the following:

1.

fd(Al) limits for Overtemperature Delta T Trip Setpoints and f2(AI) limits for Overpower Delta T Trip Setpoints for Specification 2.2.1.

2.

Moderator Temperature Coefficient BOL and EOL limits and 300 ppm surveillance limit for Specification 3/4.1.1.3,

3.

Shutdown Bank Insertion Limit for Specification 3/4.1.3.5,

4.

Control Bank Insertion Limits for Specification 3/4.1.3.6,

5.

AXIAL FLUX DIFFERENCE Limits for Specification 3/4.2.1,

6.

Heat Flux Hot Channel Factor and K(z) for Specification 3/4.2.2, and

7.

Nuclear Enthalpy Rise Hot Channel Factor for Specification 3/4.2.3.

6.9.1.14.a The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by NRC, specifically those described in the following documents:

The COLR will contain the complete identification for each of the TS referenced topical reports used to prepare the COLR (i.e., report number, title, revision, date, and any supplements).

1.

BAW-10180P-A, Revision 1, "NEMO - Nodal Expansion Method Optimized," March 1993

2.

BAW-10169P-A, Revision 0, "RSG Plant Safety Analysis - B&W Safety Analysis Methodology for Recirculating Steam Generator Plants," October 1989

3.

BAW-10163P-A, Revision 0, "Core Operating Limit Methodology for Westinghouse Designed PWRs," June 1989

4.

EMF-2328(P)(A), "PWR Small Break LOCA Evaluation Model," March 2001 SEQUOYAH - UNIT 2 6-13 Amendment No. 44, 50, 64, 66, 107, 134,142, 146,161,206, 214, 223, 272, 289, 303,1

ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT (continued)

5.

BAW-10227P-A, Revision 1, "Evaluation of Advanced Cladding and Structural Material (M5) in PWR Reactor Fuel," June 2003

6.

BAW-10186-A, Revision 2, "Extended Burnup Evaluation," June 2003

7.

EMF-2103P-A, Revision 0, "Realistic Large Break LOCA Methodology for Pressurized Water Reactors," April 2003

8.

BAW-10241 P-A, Revision 1, "BHTP DNB Correlation Applied with LYNXT," July 2005.

9.

BAW-10199P-A, Revision 0, "The BWU Critical Heat Flux Correlations," August 1996

10.

BAW-10189P-A, "CHF Testing and Analysis of the Mark-BW Fuel Assembly Design,"

January 1996

11.

BAW-10159P-A, "BWCMV Correlation of Critical Heat Flux in Mixing Vane Grid Fuel Assemblies," August 1990

12.

BAW-10231(P)(A), Revision 1, "COPERNIC Fuel Rod Design Computer Code," January 2004 6.9.1.14.b The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met.

6.9.1.14.c THE CORE OPERATING LIMITS REPORT shall be provided within 30 days after cycle start-up (Mode 2) for each reload cycle or within 30 days of issuance of any midcycle revision of the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.

REACTOR COOLANT SYSTEM (RCS) PRESSURE AND TEMPERATURE LIMITS (PTLR) REPORT 6.9.1.15 RCS pressure and temperature limits for heatup, cooldown, low temperature operation, criticality, and hydrostatic testing, LTOP arming, and PORV lift settings as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:

Specification 3.4.9.1, "RCS Pressure and Temperature (P/T) Limits" Specification 3.4.12, "Low Temperature Over Pressure Protection (LTOP) System" 6.9.1.15.a The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

1.

Westinghouse Topical Report WCAP-14040-NP-A, "Methodology used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves."

2.

Westinghouse Topical Report WCAP-15321, "Sequoyah Unit 2 Heatup and Cooldown Limit Curves for Normal Operation and PTLR Support Documentation."

3.

Westinghouse Topical Report WCAP-1 5984, "Reactor Vessel Closure HeadNessel Flange Requirements Evaluation for Sequoyah Units 1 and 2."

6.9.1.15.b The PTLR shall be provided to the NRC within 30 days of issuance of any revision or supplement thereto.

SEQUOYAH - UNIT 2 6-14 Amendment No. 44, 50, 64, 66, 107, 134,146,206,214,231,249,284,303,305,311,

ADMINISTRATIVE CONTROLS STEAM GENERATOR (SG) TUBE INSPECTION REPORT 6.9.1.16.1 A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with Specification 6.8.4.k, "Steam Generator (SG) Program." The report shall include:

a. The scope of inspections performed on each SG,

b. Active degradation mechanisms found,

c. Nondestructive examination techniques utilized for each degradation mechanism,

d.

Location, orientation (if linear), and measured sizes (if available) of service induced indications,

e. Number of tubes plugged during the inspection outage for each active degradation mechanism,

f.

Total number and percentage of tubes plugged to date,

g.

The results of condition monitoring, including the results of tube pulls and in-situ testing, and

h.

The effective plugging percentage for all plugging in each SG.

6.9.1.16.2 A report shall be submitted within 90 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with the steam generator program (6.8.4.k) when voltage based alternate repair criteria have been applied. The report shall include information described in Section 6.b of Attachment 1 to NRC Generic Letter 95-05, "Voltage-Based Repair Criteria for Westinghouse Steam Generator Tubes Affected by Outside Diameter Stress Corrosion Cracking."

6.9.1.16.3 For implementation of the voltage-based repair criteria for tube support plate (TSP) intersections, notify the staff prior to initial entry into MODE 4 following completion of an inspection performed in accordance with Specification 6.8.4.k, "Steam Generator (SG)

Program," should any of the following conditions arise:

1) If circumferential crack-like indications are detected at the TSP intersections.

2)

If indications are identified that extend beyond the confines of the TSP.

3)

If indications are identified at the TSP elevations that are attributable to primary water stress corrosion cracking.

6.9.1.16.4 For implementation of W*, the calculated steam line break leakage from the application of TSP alternate repair criteria and W* inspection methodology shall be submitted within 90 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with Specification 6.8.4.k, "Steam Generator (SG) Program." The report will include the number of indications within the tubesheet region, the location of the indications (relative to the bottom of the WEXTEX transition [BWT] and TTS), the orientation (axial, circumferential, skewed, volumetric), the severity of each indication (e.g., near through-wall or not through-wall), the side of the tube from which the indication initiated (inside or outside diameter), and an assessment of whether the results were consistent with expectations with respect to the number of flaws and flaw severity (and if not consistent, a description of the proposed corrective action).

SEQUOYAH - UNIT 2 6-14a Amendment No. 305,

ATTACHMENT 4 Proposed TS Bases Changes (Final Typed) for SQN, Units 1 and 2

Proposed TS Bases Changes (Final Typed) for SQN, Unit 1

2.1 SAFETY LIMITS BASES 2.1.1 REACTOR CORE The restrictions of this safety limit prevent overheating of the fuel cladding (due to departure from nucleate boiling) and overheating of the fuel pellet (centerline fuel melt), either of which could result in cladding perforation that would result in the release of fission products to the reactor coolant.

Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature. Overheating of the fuel is prevented by maintaining the steady state peak linear heat rate (LHR) below the level at which fuel centerline melting occurs.

Operation above the upper boundary of the nucleate boiling regime could result in excessive temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient from the outer surface of the cladding to the reactor coolant water.

DNB is not a directly measurable parameter during operation and therefore THERMAL POWER and Reactor Coolant Temperature and Pressure have been related to DNB. The DNB correlations have been developed to predict the DNB flux and the location of DNB for axially uniform and non-uniform heat flux distributions. The local DNB heat flux ratio, DNBR, defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux, is indicative of the margin to DNB.

The DNB design basis is that there must be at least a 95 percent probability with 95 percent confidence that DNB will not occur when the minimum DNBR is at the design DNBR limit.

To meet the DNB Design Basis, a statistical core design (SCD) process has been used to develop an appropriate statistical DNBR design limit. Uncertainties in plant operating parameters, nuclear and thermal parameters, and fuel fabrication parameters are considered statistically such that there is at least a 95 percent probability at a 95 percent confidence level that the minimum DNBR for the limiting rod is greater than or equal to the DNBR limit. This DNBR uncertainty derived from the SCD analysis, combined with the applicable DNB critical heat flux correlation limit, establishes the statistical DNBR design limit which must be met in plant safety analysis using values of input parameters without adjustment for uncertainty.

Operation above the boundary of the nucleate boiling regime could result in excessive cladding temperature because of the onset of DNB and the resultant sharp reduction in heat transfer coefficient.

Inside the steam film, high cladding temperatures are reached, and a cladding water (zirconium water) reaction may take place. This chemical reaction results in oxidation of the fuel cladding to a structurally weaker form. This weaker form may lose its integrity, resulting in an uncontrolled release of activity to the reactor coolant.

Operation above the maximum local linear heat generation rate for fuel melting could result in excessive fuel pellet temperature and cause melting of the fuel at its centerline. Fuel centerline melting occurs when the local LHR, or power peaking, in a region of the fuel is high enough to cause the fuel centerline temperature to reach the melting point of the fuel. Expansion of the pellet upon centerline melting may cause the pellet to stress the cladding to the point of failure, allowing an uncontrolled release of activity to the reactor coolant. The melting point of uranium dioxide varies slightly with burnup. As uranium is depleted and fission products produced, the net effect is a decrease in the melting point. Fuel centerline temperature is not a directly measurable parameter during operation. The maximum location fuel pin centerline temperature is maintained by limiting the local linear heat generation rate in the fuel.

The local linear heat generation rate in the fuel is limited so that the maximum fuel centerline temperature will exceed the value defined by the algorithm in SL 2.1.1.2.

SEQUOYAH - UNIT 1 B 2-1 Amendment No. 19, 114, 138, 155, 223,

SAFETY LIMITS BASES The curves of Figure 2.1-1 show the loci of points of THERMAL POWER, Reactor Coolant System pressure and average temperature for which the minimum DNBR is no less than the safety analysis DNBR limit, or the average enthalpy at the vessel exit is equal to the enthalpy of saturated liquid.

The curves of Figure 2.1-1 are based on an enthalpy rise hot channel factor, FNH, (nominal values have been reduced to include a 4% total rod power uncertainty factor), and a design axial power distribution, Fz with a reference cosine shape at a peak of 1.55. An allowance is included for an increase in FNH as THERMAL POWER is reduced based on the expression:

F M=aFMTP[1+'3 (1-P)]

where

=

THERMAL POWER and RATED THERMAL POWER F P =Nominal Value 1.70 These limiting heat flux conditions for DNB are higher than those calculated for the range of all control rods fully withdrawn to the maximum allowable control rod insertion assuming the axial power imbalance or Delta-I (AI) is within the limits of the f, (Delta I) function of the Overtemperature Delta Temperature trip. When the axial power imbalance exceeds the tolerance (or deadband) of the f1(AI) trip reset function, the Overtemperature Delta Temperature trip setpoint is reduced by the values in the CORE OPERATING LIMITS REPORT to provide protection required by the core safety limits.

Similarly, the limiting linear heat generation rate conditions for CFM are higher than those calculated for the range of all control rods from the fully withdrawn to the maximum allowable control rod insertion assuming the axial power imbalance or Delta-I (AI) is within the limits of the f2(AI) function of the OverPower-Delta Temperature trip. When the axial imbalance exceeds the tolerance (or deadband) of the f2(AI) trip reset function, the OverPower-Delta Temperature trip setpoint is reduced by the values specified in the CORE OPERATING LIMITS REPORT to provide protection required by the core safety limits.

SEQUOYAH - UNIT 1 B 2-1a Amendment No.

SAFETY LIMITS BASES 2.1.2 REACTOR COOLANT SYSTEM PRESSURE The restriction of this Safety Limit protects the integrity of the Reactor Coolant System from overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere.

The reactor pressure vessel and pressurizer are designed to Section III of the ASME Code for Nuclear Power Plant which permits a maximum transient pressure of 110% (2735 psig) of design pressure. The Reactor Coolant System piping, valves and fittings, are designed to ANSI B 31.1 1967 Edition, which permits a maximum transient pressure of 120% (2985 psig) of component design pressure.

The Safety Limit of 2735 psig is therefore consistent with the design criteria and associated code requirements.

The entire Reactor Coolant System is hydro tested at 3107 psig, 125% of design pressure, to demonstrate integrity prior to initial operation.

2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS The Reactor Nominal Trip Setpoint Limits specified in Table 2.2-1 are the values at which the Reactor Trips are set for each functional unit. The Nominal Trip Setpoints have been selected to ensure that the reactor core and reactor coolant system are prevented from exceeding their safety limits during normal operation and design basis anticipated operational occurrences and to assist the Engineered Safety Features Actuation System in mitigating the consequences of accidents. Operation with a trip set less conservative than its Nominal Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Nominal Trip Setpoint and the Allowable Value is equal to or less than the rack allowance assumed for each trip in the safety analyses.

Technical specifications are required by 10 CFR 50.36 to contain Limiting Safety System Settings (LSSS) defined by the regulation as "... settings for automatic protective devices... so chosen that automatic protective action will correct the abnormal situation before a Safety Limit (SL) is exceeded."

The analytic limit is the limit of the process variable at which a safety action is initiated, as established by the safety analysis, to ensure that a SL is not exceeded. Any automatic protection action that occurs on reaching the analytic limit therefore ensures that the SL is not exceeded. However, in practice, the actual settings for automatic protective devices must be chosen to be more conservative than the analytic limit to account for instrument loop uncertainties related to the setting at which the automatic protective action would actually occur.

SEQUOYAH - UNIT 1 B 2-2 Amendment No. 310,

POWER DISTRIBUTION LIMITS BASES 3/4.2.4 QUADRANT POWER TILT RATIO The QUADRANT POWER TILT RATIO limit assures that no anomaly exists such that the radial power distribution satisfies the design values used in the power capability analysis. Radial power distribution measurements are made during startup testing and periodically during power operation.

The QUADRANT POWER TILT RATIO limit at which corrective action is required provides DNB and linear heat generation protection with x-y plane power tilts. The QUADRANT POWER TILT RATIO limit is reflected by a corresponding peaking augmentation factor which is included in the generation of the AFD limits.

The 2-hour time allowance for operation with the tilt condition greater than 1.02 but less than 1.09, is provided to allow identification and correction of a dropped or misaligned control rod. In the event such action does not correct the tilt, the margin for uncertainty on FQ(X,Y,Z) is reinstated by reducing the allowable THERMAL POWER by 3 percent for each percent of tilt in excess of 1.02.

3/4.2.5 DNB PARAMETERS The limits on the DNB related parameters assure that each of the parameters are maintained within the normal steady state envelope of operation assumed in the transient and accident analyses.

The limits are consistent with the initial FSAR assumptions and have been analytically demonstrated adequate to maintain a minimum DNBR of greater than or equal to the safety analysis DNBR limit throughout each analyzed transient.

The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> periodic surveillance of these parameters through instrument readout is sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation.

SEQUOYAH - UNIT 1

223, B 3/4 2-4 Amendment No. 19, 138, 155,

Proposed TS Bases Changes (Final Typed) for SQN, Unit 2

2.1 SAFETY LIMITS BASES 2.1.1 REACTOR CORE The restrictions of this Safety Limit prevent overheating of the fuel cladding (due to departure from nucleate boiling) and overheating of the fuel pellet (centerline fuel melt), either of which could result in cladding perforation that would result in the release of fission products to the reactor coolant.

Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature. Overheating of the fuel is prevented by maintaining the steady state peak linear heat rate (LHR) below the level at which fuel centerline melting occurs.

Operation above the upper boundary of the nucleate boiling regime could result in excessive temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient from the outer surface of the cladding to the reactor coolant water.

DNB is not a directly measurable parameter during operation and; therefore, THERMAL POWER and Reactor Coolant Temperature and Pressure have been related to DNB. The DNB correlations have been developed to predict the DNB flux and the location of DNB for axially uniform and non-uniform heat flux distributions. The local DNB heat flux ratio, DNBR, defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux, is indicative of the margin to DNB.

The DNB design basis is that there must be at least a 95 percent probability with 95 percent confidence that DNB will not occur when the minimum DNBR is at the design DNBR limit.

To meet the DNB Design Basis, a statistical core design (SCD) process has been used to develop an appropriate statistical DNBR design limit. Uncertainties in plant operating parameters, nuclear and thermal parameters, and fuel fabrication parameters are considered statistically such that there is at least a 95 percent probability at a 95 percent confidence level that the minimum DNBR for the limiting rod is greater than or equal to the DNBR limit. This DNBR uncertainty derived from the SCD analysis, combined with the applicable DNB critical heat flux correlation limit, establishes the statistical DNBR design limit which must be met in plant safety analysis using values of input parameters without adjustment for uncertainty.

Operation above the boundary of the nucleate boiling regime could result in excessive cladding temperature because of the onset of DNB and the resultant sharp reduction in heat transfer coefficient.

Inside the steam film, high cladding temperatures are reached, and a cladding water (zirconium water) reaction may take place. This chemical reaction results in oxidation of the fuel cladding to a structurally weaker form. This weaker form may lose its integrity, resulting in an uncontrolled release of activity to the reactor coolant.

Operation above the maximum local linear heat generation rate for fuel melting could result in excessive fuel pellet temperature and cause melting of the fuel at its centerline. Fuel centerline melting occurs when the local LHR, or power peaking, in a region of the fuel is high enough to cause the fuel centerline temperature to reach the melting point of the fuel. Expansion of the pellet upon centerline melting may cause the pellet to stress the cladding to the point of failure, allowing an uncontrolled release of activity to the reactor coolant. The melting point of uranium dioxide varies slightly with burnup. As uranium is depleted and fission products produced, the net effect is a decrease in the melting point. Fuel centerline temperature is not a directly measurable parameter during operation. The maximum location fuel pin centerline temperature is maintained by limiting the local linear heat generation rate in the fuel.

The local linear heat generation rate in the fuel is limited so that the maximum fuel centerline temperature will exceed the value defined by the algorithm in SL 2.1.1.2.

SEQUOYAH - UNIT 2 B 2-1 Amendment No. 21, 104, 130, 146, 214,

SAFETY LIMITS BASES The curves of Figure 2.1-1 show the loci of points of THERMAL POWER, Reactor Coolant System pressure and average temperature for which the minimum DNBR is no less than the safety analysis DNBR limit, or the average enthalpy at the vessel exit is equal to the enthalpy of saturated liquid.

The curves of Figure 2.1-1 are based on an enthalpy rise hot channel N

factor, FAH, (nominal values have been reduced to include a 4% total rod power uncertainty factor), and a design axial power distribution, Fz with a reference cosine shape at a peak of 1.55. An allowance is included for an increase in FrýH as THERMAL POWER is reduced based on the expression:

FNE =FAH[I+.3 (1-P)]

where P THERMAL POWER and RATED THERMAL POWER FH= Nominal Value 1.70 These limiting heat flux conditions for DNB are higher than those calculated for the range of all control rods fully withdrawn to the maximum allowable control rod insertion assuming the axial power imbalance or Delta-I (Al) is within the limits of the fj(delta I) function of the Overtemperature Delta Temperature trip. When the axial power imbalance exceeds the tolerance (or deadband) of the f1(AI) trip reset function, the Overtemperature Delta Temperature trip setpoint is reduced by the values in the CORE OPERATING LIMITS REPORT to provide protection required by the core safety limits.

Similarly, the limiting linear heat generation rate conditions for CFM are higher than those calculated for the range of all control rods from the fully withdrawn to the maximum allowable control rod insertion assuming the axial power imbalance or Delta-I (Al) is within the limits of the f2(AI) function of the OverPower-Delta Temperature trip. When the axial imbalance exceeds the tolerance (or deadband) of the f2(AI) trip reset function, the OverPower-Delta Temperature trip setpoint is reduced by the values specified in the CORE OPERATING LIMITS REPORT to provide protection required by the core safety limits.

SEQUOYAH - UNIT 2 B 2-1 a Amendment No.

SAFETY LIMITS BASES 2.1.2 REACTOR COOLANT SYSTEM PRESSURE The restriction of this Safety Limit protects the integrity of the Reactor Coolant System from overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere.

The reactor pressure vessel and pressurizer are designed to Section III of the ASME Code for Nuclear Power Plant which permits a maximum transient pressure of 110% (2735 psig) of design pressure. The Reactor Coolant System piping, valves and fittings, are designed to ANSI B 31.1 1967 Edition, which permits a maximum transient pressure of 120% (2985 psig) of component design pressure.

The Safety Limit of 2735 psig is therefore consistent with the design criteria and associated code requirements.

The entire Reactor Coolant System is hydrotested at 3107 psig, 125% of design pressure, to demonstrate integrity prior to initial operation.

2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS The Reactor Nominal Trip Setpoint Limits specified in Table 2.2-1 are the values at which the Reactor Trips are set for each functional unit. The Nominal Trip Setpoints have been selected to ensure that the reactor core and reactor coolant system are prevented from exceeding their safety limits during normal operation and design basis anticipated operational occurrences and to assist the Engineered Safety Features Actuation System in mitigating the consequences of accidents. Operation with a trip set less conservative than its Nominal Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Nominal Trip Setpoint and the Allowable Value is equal to or less than the rack allowance assumed for each trip in the safety analyses.

Technical specifications are required by 10 CFR 50.36 to contain Limiting Safety System Settings (LSSS) defined by the regulation as ".... settings for automatic protective devices... so chosen that automatic protective action will correct the abnormal situation before a Safety Limit (SL) is exceeded."

The analytic limit is the limit of the process variable at which a safety action is initiated, as established by the safety analysis, to ensure that a SL is not exceeded. Any automatic protection action that occurs on reaching the analytic limit therefore ensures that the SL is not exceeded. However, in practice, the actual settings for automatic protective devices must be chosen to be more conservative than the analytic limit to account for instrument loop uncertainties related to the setting at which the automatic protective action would actually occur.

The Nominal Trip Setpoint is a predetermined setting for a protective device chosen to ensure automatic actuation prior to the process variable reaching the analytic limit and thus ensuring that the SL would not be exceeded. As such, the Nominal Trip Setpoint accounts for uncertainties in setting the device (e.g., calibration), uncertainties in how the device might actually perform (e.g., repeatability),

changes in the point of action of the device over time (e.g., drift during surveillance intervals), and any other factors which may influence its actual performance (e.g., harsh accident environments). In this manner, the Nominal Trip Setpoint plays an important role in ensuring that SLs are not exceeded. As such, the Nominal Trip Setpoint meets the definition of an LSSS in accordance with Regulatory Guide 1.105, Revision 3, "Setpoints for Safety-Related Instrumentation," and could be used to meet the requirements that they be contained in the technical specifications.

SEQUOYAH - UNIT 2 B 2-2 Amendment No. 130, 146, 299,

POWER DISTRIBUTION LIMITS BASES 3/4.2.4 QUADRANT POWER TILT RATIO The QUADRANT POWER TILT RATIO limit assures that no anomaly exists such that the radial power distribution satisfies the design values used in the power capability analysis. Radial power distribution measurements are made during startup testing and periodically during power operation.

The QUADRANT POWER TILT RATIO limit at which corrective action is required provides DNB and linear heat generation protection with x-y plane power tilts. The QUADRANT POWER TILT RATIO limit is reflected by a corresponding peaking augmentation factor which is included in the generation of the AFD limits.

The 2-hour time allowance for operation with the tilt condition greater than 1.02 but less than 1.09, is provided to allow identification and correction of a dropped or misaligned control rod. In the event such action does not correct the tilt, the margin for uncertainty on FQ(X,Y,Z) is reinstated by reducing the allowable THERMAL POWER by 3 percent for each percent of tilt in excess of 1.02.

3/4.2.5 DNB PARAMETERS The limits on the DNB related parameters assure that each of the parameters are maintained within the normal steady state envelope of operation assumed in the transient and accident analyses.

The limits are consistent with the initial FSAR assumptions and have been analytically demonstrated adequate to maintain a minimum DNBR of greater than or equal to the safety analysis DNBR limit throughout each analyzed transient.

The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> periodic surveillance of these parameters through instrument readout is sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation.

SEQUOYAH - UNIT 2 B 3/4 2-4 Amendment 21, 130, 146, 214,