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License Amendment Request 258, Revise Reactor Core Safety Limit to Reflect WCAP-17642-PA, Revision 1
	ML18127B714
Person / Time
Site:	Turkey Point
Issue date:	05/03/2018
From:	Coffey R; Florida Power & Light Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	L-2018-065
	Download: ML18127B714 (14)
	v • d • e

May 3, 2018 L-2018-065 10 CFR 50.90 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington DC 20555-0001 RE: Turkey Point Nuclear Plant, Unit 3 and 4 Docket Nos. 50-250 and 50-251 Renewed Facility Operating Licenses DPR-31 and DPR-41 License Amendment Request 258, Revise Reactor Core Safety Limit to Reflect WCAP-17642-P-A Revision 1 Pursuant to 10 CFR Part 50.90, Florida Power & Light Company (FPL) hereby requests amendments to Renewed Facility Operating Licenses DPR-31 and DPR-41 for Turkey Point Nuclear Plant Units 3 and 4 (Turkey Point), respectively. The proposed license amendments revise the Turkey Point Technical Specifications (TS) by revising Reactor Core Safety Limit (SL) 2.1.1. b, to reflect the peak fuel centerline melt temperature specified in WCAP-17642-P-A, Revision 1, "Westinghouse Performance Analysis and Design Model (PAD5) ".

The enclosu re to this letter provides FPL's evaluation of the proposed changes. Attachment 1 to the enclosure provides a mark-up of the existing TS page to show the proposed changes. Attachment 2 provides the existing TS Bases page marked up to show the proposed changes. The TS Bases changes are provided for information only and will be incorporated in accordance with the TS Bases Control Program upon implementation of the approved license amendments.

FPL has determined that the proposed changes do not involve a significant hazards consideration pursuant to 10 CFR 50 .92(c), and there are no significant environmental impacts associated with the proposed changes. The Turkey Point Onsite Review Group has reviewed the proposed license amendments. In accordance with 10 CFR 50 .91 (b)(1 ), a copy of the proposed license amendments are being forwarded to the State designee for the State of Florida.

FPL requests that the proposed changes are processed as a normal license amendment request. Once approved, the amendments will be implemented for the Unit 3 Cycle 31 and Unit 4 Cycle 32 reload campaigns, currently scheduled in Spring 2020 and Fall 2020, respectively.

This letter contains no regulatory commitments .

Should you have any questions regarding this submission , please contact Mr. Robert Hess, Turkey Point Licensing Manager, at 305-246-4112.

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

Executed on the _3r_ct__ day of May 2018.

Sincerely,

~

Regional Vice President, Southern Region Turkey Point Nuclear Plant Florida Power & Light Company 9760 SW 344tl, St. , Homestead, FL 33035

Turkey Point Nuclear Plant L-2018-065 Docket Nos. 50-250 and 50-251 Page 2 of 2 Enclosure Attachments cc: USNRC Regional Administrator, Region II USNRC Project Manager, Turkey Point Nuclear Plant USNRC Senior Resident Inspector, Turkey Point Nuclear Plant Ms. Cindy Becker, Florida Department of Health

Turkey Point Nuclear Plant L-2018-065 Docket Nos. 50-250 and 50-251 Enclosure Page 1 of 12 EVALUATION OF THE PROPOSED CHANGES Turkey Point Nuclear Plant Unit 3 and Unit 4 License Amendment Request 258, Revise Reactor Core Safety Limit to Reflect WCAP-17642-P-A, Revision 1, Westinghouse Performance Analysis and Design Model (PAD5) 1.0

SUMMARY

DESCRIPTION............................................................................................................... 2 2.0 DETAILED DESCRIPTION ............................................................................................................... 2 2.1 System Design and Operation ............................................................................................ 2 2.2 Current Requirements ......................................................................................................... 2 2.3 Description of the Proposed Change .................................................................................. 2 2.4 Reason for the Proposed Change ...................................................................................... 2

3.0 TECHNICAL EVALUATION

.............................................................................................................. 3

4.0 REGULATORY EVALUATION

......................................................................................................... 5 4.1 Applicable Regulatory Requirements/Criteria ..................................................................... 5 4.2 No Significant Hazards Consideration ................................................................................ 5 4.3 Conclusion .......................................................................................................................... 7

5.0 ENVIRONMENTAL CONSIDERATION

............................................................................................ 7

6.0 REFERENCES

.................................................................................................................................. 8

.. - Proposed Technical Specification Page (markup) - Proposed Technical Specification Bases Page (markup), Information Only

Turkey Point Nuclear Plant L-2018-065 Docket Nos. 50-250 and 50-251 Enclosure Page 2 of 12 1.0

SUMMARY

DESCRIPTION Florida Power & Light Company (FPL) hereby requests amendments to Renewed Facility Operating Licenses DPR-31 and DPR-41 for Turkey Point Nuclear Plant Units 3 and 4 (Turkey Point), respectively. The proposed license amendments revise the Turkey Point Technical Specifications (TS) Reactor Core Safety Limit (SL) 2.1.1.b, to reflect the peak fuel centerline melt temperature specified in WCAP-17642-P-A, Revision 1, Westinghouse Performance Analysis and Design Model (PAD5).

2.0 DETAILED DESCRIPTION 2.1 System Design and Operation The Turkey Point nuclear units must ensure that acceptable fuel design limits are not exceeded during steady state operation, normal operational transients, and anticipated operational occurrences (AOOs), consistent with the Turkey Point licensing basis as specified in 1967 Proposed General Design Criteria (GDC) 6. To accomplish this, Turkey Point TS 2.1.1, Reactor Core Safety Limits, ensure that departure from nucleate boiling (DNB) does not occur and that the fuel centerline temperature remains below the fuel melting temperature. The proposed amendment revises the fuel centerline melting temperature specified in SL 2.1.1.b, but does not alter the Safety Limit associated with the DNB ratio.

The restrictions of SL 2.1.1.b prevent overheating of the fuel and cladding, as well as possible cladding perforation, which would result in the release of fission products to the reactor coolant. 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.

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 fuel pellet following centerline melting could cause excessive cladding stress leading to failure of the cladding and uncontrolled release of fission products to the reactor coolant.

Proper functioning of the Reactor Protection System (RPS) and the Main Stream Safety Valves prevents violation of the Reactor Core SLs.

2.2 Current Technical Specification Requirements SL 2.1.1.b defines the burnup-dependent temperature below which the fuel centerline temperature must be maintained. SL 2.1.1.b applies during MODES 1 and 2, i.e. when the reactor is critical and requires placing the applicable Unit in MODE 3 (Hot Standby) within one hour in the event the Safety Limit is violated.

2.3 Reason for the Proposed Change Plant-specific safety analyses are performed to ensure that compliance with plant Safety Limits is maintained. Westinghouse Performance Analysis and Design Model (PAD5) methodology (Reference 6.1) defined the fuel pellet melting limit that is included in the PAD5 methodology based on available fuel pellet material properties. The Nuclear Regulatory Commission (NRC) staff reviewed and approved the Westinghouse methodology and concluded that the melting limits defined in Reference 6.1 are acceptable.

Turkey Point Nuclear Plant L-2018-065 Docket Nos. 50-250 and 50-251 Enclosure Page 3 of 12 The proposed change will be implemented to maintain consistency between the value specified in SL 2.1.1.b and the criteria used when performing confirmatory safety analyses that rely on the NRC approved methodology in Reference 6.1.

2.4 Description of the Proposed Change The proposed change revises the peak fuel centerline temperature specified in SL 2.1.1.b, but does not alter the Required Action that must be taken following a violation of the limit..

The following changes are proposed to the Turkey Point TS.

The current version of SL 2.1.1.b reads:

The peak fuel centerline temperature shall be maintained <5080°F, decreasing by 58°F per 10,000 MWD/MTU of burnup.

The revised version of SL 2.1.1.b would read:

The peak fuel centerline temperature shall be maintained <5080°F, decreasing by 9°F per 10,000 MWD/MTU of burnup.

A mark-up of the proposed change to TS Section 2.1.1 is provided in Attachment 1.

3.0 TECHNICAL EVALUATION

The proposed license amendments revise the Turkey Point TS by revising Reactor Core Safety Limit (SL) 2.1.1.b, to reflect the peak fuel centerline melt temperature discussed in WCAP-17642-P/NP, Revision 1, Westinghouse Performance Analysis and Design Model (PAD5).

3.1 Revise Peak Fuel Centerline Temperature in Safety Limit 2.1.1.b The principal design tool used by Westinghouse for evaluating fuel rod performance is the Performance Analysis and Design (PAD) code. This computer program iteratively calculates the interrelated effects of fuel and cladding deformations including fuel densification, fuel swelling, fuel relocation, fuel rod temperatures, fill and fission gas release (FGR), and rod internal pressure (RIP), as a function of time and linear power.

PAD evaluates the power history of a fuel rod as a series of steady-state power levels with instantaneous jumps from one power level to another. The length of the fuel rod is divided into several axial segments and each segment is assumed to operate at a constant set of conditions over its length. Fuel densification and swelling, cladding stresses and strains, temperatures, burnup and fission gas releases are calculated separately for each axial segment and the effects are integrated to obtain the overall fission gas release and resulting internal pressure for each time step. The coolant temperature rise along the fuel rod is calculated based on the flow rate and axial power distribution, and the cladding surface temperature is determined with consideration of corrosion effects and the possibility of local boiling.

Model updates incorporated into the PAD5 code address all of the fuel and cladding performance models required for high burnup fuel design. Key fuel performance updates to the PAD5 models include fuel thermal conductivity degradation (TCD) with burnup, enhanced high burnup athermal fission gas release (pellet rim effects) and enhanced high burnup fission gas bubble swelling. Cladding creep and growth models are also updated to reflect high burnup cladding performance. In addition to high burnup analysis

Turkey Point Nuclear Plant L-2018-065 Docket Nos. 50-250 and 50-251 Enclosure Page 4 of 12 capability, a key driver for the implementation of the PAD5 models in fuel design is to address regulatory concerns associated with fuel thermal conductivity degradation with burnup.

The PAD5 models are the latest evolutions of the Westinghouse PAD code (Reference 6.1). As part of the development, the burnup-dependent term of the fuel melting limits in PAD5 was updated based on journal-published fuel material data. Additional validation performed in Section 2.1 of Appendix A of Reference 6.1 shows that the PAD5 code, in conjunction with the new fuel melt limit, accurately predicts fuel melt based on comparisons to experimental observations. Section 3.7.12 of the NRC Safety Evaluation Report concluded that the fuel melting limits in PAD5 are acceptable.

A comprehensive description of all PAD5 models, NRC Requests for Additional Information (RAI), and the NRCs safety evaluation are also documented in Reference 6.1. The NRC Safety Evaluation Limitations and Conditions are discussed below.

3.2 Limits of Applicability FPL intends that the proposed amendments will be implemented with the approved fuel performance methods in Reference 6.1. As such, the Limitations and Conditions from the NRCs Final Safety Evaluation Report in Reference 6.1 pertinent to this amendment request are detailed below along with details of how each is satisfied.

The NRC staff limits the applicability of the PAD5 code and methodology to the cladding, fuel, and reactor parameters listed in Section 4.1 of Reference 6.1.

Response: FPL will apply PAD5 within the limits specified in Section 4.1 of Reference 6.1 for cladding, fuel and reactor parameters to be used at Turkey Point.

Since these PAD5 inputs depend on the reload design, these parameters are validated on a cycle-specific basis.

The application of PAD5 should at no time exceed the fuel melting temperature as calculated by PAD5 due to the lack of properties for molten fuel in PAD5 and other properties such as thermal conductivity and fission gas release.

Response: FPL will limit the peak fuel centerline temperature per this amendment request.

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria Section 182a of the Atomic Energy Act requires applicants for nuclear power plant operating licenses to include Technical Specifications (TS) as part of the license.

The TS ensure the operational capability of structures, systems, and components that are required to protect the health and safety of the public.

10 CFR 50.90 requires NRC approval for any modification to, addition to, or deletion from the plant TS. Therefore, this activity requires NRC approval prior to making the plant-specific changes in this license amendment request.

10 CFR 50.36 requires that the TS include items in the following specific categories:

(1) safety limits, limiting safety systems settings, and limiting control settings; (2) limiting conditions for operation (LCOs); (3) surveillance requirements per 10 CFR

Turkey Point Nuclear Plant L-2018-065 Docket Nos. 50-250 and 50-251 Enclosure Page 5 of 12 50.36(c)(3); (4) design features; and (5) administrative controls. This amendment application is related to the first category above since a change to the peak fuel centerline melt temperature Safety Limit is proposed.

1967 Proposed General Design Criteria (GDC) 6 states that the reactor core with its related controls and protection systems shall be designed to function throughout its design lifetime without exceeding acceptable fuel damage limits. The core and related auxiliary system designs shall provide this integrity under all expected conditions of normal operation with appropriate margins for uncertainties and for specified transient situations which can be anticipated. The restrictions of SL 2.1.1.b prevent overheating of the fuel and cladding.

Overheating of the fuel is prevented by maintaining the steady state peak temperature below the level at which fuel centerline melting occurs. 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 change to SL 2.1.1.b changes the limit to be consistent with the limit approved in Reference 6.1, thus the requirement of 1967 Proposed General Design Criteria (GDC) 6 continues to be met.

4.2 No Significant Hazards Consideration The proposed license amendments revise the Turkey Point Technical Specifications (TS) by revising Reactor Core Safety Limit (SL) 2.1.1.b, to reflect the peak fuel centerline melt temperature discussed in Final Safety Evaluation for WCAP-17642-P-A, Revision 1, Westinghouse Performance Analysis and Design Model (PAD5) (Reference 6.1).

As required by 10 CFR 50.91(a), FPL has evaluated the proposed change using the criteria in 10 CFR 50.92 and has determined that the proposed change does not involve a significant hazards consideration. An analysis of the issue of no significant hazards consideration is presented below:

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

Response: No There are no design changes associated with the proposed amendments. All design, material, and construction standards that were applicable prior to this amendment request will continue to be applicable. The proposed amendments will not affect accident initiators or precursors or alter the design, conditions, and configuration of the facility, or the manner in which the plant is operated and maintained, with respect to such initiators or precursors. Compliance with Safety Limit 2.1.1.b is required to confirm that fuel cladding failure does not occur as a result of fuel centerline melting. The fuel centerline melt temperature limit is established to preclude centerline melting. The proposed change to the fuel centerline melt temperature limit has been reviewed by the NRC and found to be appropriately conservative with respect to the fuel material properties in the Final Safety Evaluation for WCAP-17642-P-A, Revision 1, (Reference 6.1). Accident analysis acceptance criteria will continue to be met with the proposed amendments. Hence, the proposed amendments will not affect the source term, containment isolation, or radiological release assumptions used in evaluating the radiological consequences of any accident previously evaluated. The proposed amendments will not alter any assumptions or change any mitigation actions in the radiological consequence evaluations in the Turkey Point Updated Final

Turkey Point Nuclear Plant L-2018-065 Docket Nos. 50-250 and 50-251 Enclosure Page 6 of 12 Safety Analysis Report (UFSAR). Consequently, the applicable radiological dose acceptance criteria will continue to be met.

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

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

Response: No There are no proposed design changes nor are there any changes in the method by which any safety-related plant structures, systems, and components perform their specified safety functions. The proposed amendments will not affect the normal method of plant operation or change any operating parameters. No equipment performance requirements will be affected. The proposed amendments will not alter any assumptions made in the safety analyses. The proposed amendments revise Reactor Core Safety Limit 2.1.1.b; however, the change does not involve a physical modification of the plant. No new accident scenarios, transient precursors, failure mechanisms, or limiting single failures will result from this amendment. Hence, there will be no adverse effect or challenges imposed on any safety-related system as a result of these amendments.

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

(3) Do the proposed amendments involve a significant reduction in a margin of safety?

Response: No The revised Safety Limit 2.1.1.b has been calculated based on the NRC-approved methods which ensure that the plant operates in compliance with all regulatory criteria. There will be no effect on those plant systems necessary to effect the accomplishment of protection functions. No instrument setpoints or system response times are affected and none of the acceptance criteria for any accident analysis will be changed. Consequently, the proposed amendments will have no impact on the radiological consequences of a design basis accident.

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

Based upon the above analysis, FPL concludes that the proposed license amendments do not involve a significant hazards consideration, under the standards set forth in 10 CFR 50.92, Issuance of Amendment, and accordingly, a finding of no significant hazards consideration is justified.

4.3 Conclusion In summary, in accordance 10 CFR 50.90, FPL requests NRC review and approval of the change to Turkey Point Technical Specification 2.1.1, Reactor Core Safety Limits.

Based on the above discussions, (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)

Turkey Point Nuclear Plant L-2018-065 Docket Nos. 50-250 and 50-251 Enclosure Page 7 of 12 issuance of the license amendments will not be inimical to the common defense and security or to the health and safety of the public.

5.0 ENVIRONMENTAL CONSIDERATION

The proposed license amendments revise the Turkey Point Technical Specifications (TS) by revising Reactor Core Safety Limit (SL) 2.1.1.b, to reflect the peak fuel centerline melt temperature discussed in the Final Safety Evaluation for WCAP-17642-P-A, Revision 1, Westinghouse Performance Analysis and Design Model (PAD5) (Reference 6.1).

A review of the anticipated construction and operational effects of the proposed changes has determined the requested license amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9), in that:

(i) There is no significant hazards consideration.

As documented in Section 4.2, No Significant Hazards Consideration Determination, of this license amendment request, an evaluation was completed to determine whether or not a significant hazards consideration is involved by focusing on the three standards set forth in 10 CFR 50.92, Issuance of Amendment. The Significant Hazards Consideration determined that (1) the proposed amendments do not involve a significant increase in the probability or consequences of an accident previously evaluated; (2) the proposed amendments do not create the possibility of a new or different kind of accident from any accident previously evaluated; and (3) the proposed amendments do not involve a significant reduction in a margin of safety. Therefore, it is concluded 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.

(ii) There is no significant change in the types or significant increase in the amounts of any effluents that may be released offsite.

The proposed change is unrelated to any aspects of plant construction or operation that would introduce any changes to effluent types (e.g., effluents containing chemicals or biocides, sanitary system effluents, and other effluents) or affect any plant radiological or non-radiological effluent release quantities. The proposed amendments do not adversely impact any functions associated with containing, controlling, channeling, monitoring, or processing radioactive or nonradioactive materials, nor do they diminish the functionality of any design or operational features that are credited with controlling the release of effluents during plant operation. The types and quantities of expected plant effluents are not changed. No effluent release path is associated with these amendments. Neither radioactive nor nonradioactive material effluents are affected by this activity.

Furthermore, the proposed amendments do not diminish the functionality of any design or operational features that are credited with controlling the release of effluents during plant operation. Therefore, it is concluded that the proposed amendments do not involve a significant change in the types or a significant increase in the amounts of any effluents that may be released offsite.

(iii) There is no significant increase in individual or cumulative occupational radiation exposure.

The proposed amendments do not affect plant radiation zones described in Section 11 of the Turkey Point Updated Final Analysis Report (UFSAR), and controls under 10 CFR Part 20 preclude a significant increase in occupational radiation exposure. The proposed

Turkey Point Nuclear Plant L-2018-065 Docket Nos. 50-250 and 50-251 Enclosure Page 8 of 12 amendments do not adversely impact radiologically controlled zones. Plant radiation zones, radiation controls established to satisfy 10 CFR Part 20 requirements, and expected amounts and types of radioactive materials are not affected by the proposed amendments. Therefore, individual and cumulative radiation exposures are not significantly affected by this change. Therefore, the proposed amendments do not involve a significant increase in individual or cumulative occupational radiation exposure.

Based on the above review of the proposed amendments, it has been determined that anticipated construction and operational effects of the proposed amendments do not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in the individual or cumulative occupational radiation exposure. Accordingly, the proposed amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), an environmental impact statement or environmental assessment of the proposed amendments is not required.

6.0 REFERENCES

6.1 WCAP-17642-P-A, Revision 1, Westinghouse Performance Analysis and Design Model (PAD5), November 2017

Turkey Point Nuclear Plant L-2018-065 Docket Nos. 50-250 and 50-251 Enclosure Page 9 of 12 ATTACHMENT 1 PROPOSED TECHNICAL SPECIFICATION PAGE (MARKUP)

(1 page follows)

Turkey Point Nuclear Plant L-2018-065 Docket Nos. 50-250 and 50-251 Enclosure Page 10 of 12 ATTACHMENT 1 PROPOSED TECHNICAL SPECIFICATION PAGE (MARKUP) 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 specified in the COLR, for 3 loop operation; and the following Safety Limits shall not be exceeded:

a. The departure from nucleate boiling ratio (DNBR) shall be maintained > 1.17 for the WRB-1 DNB correlation.

b. The peak fuel centerline temperature shall be maintained < 5080°F, decreasing by 58°F per 10,000 MWD/MTU of burnup.

9oF

APPLICABILITY: MODES 1 and 2.

ACTION:

Whenever the point defined by the combination of the highest operating loop average temperature and THERMAL POWER has exceeded the appropriate pressurizer pressure line, be in HOT STANDBY within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

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 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

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.

TURKEY POINT - UNITS 3 & 4 2-1 AMENDMENT NOS. 247 AND 243

Turkey Point Nuclear Plant L-2018-065 Docket Nos. 50-250 and 50-251 Enclosure Page 11 of 12 ATTACHMENT 2 PROPOSED TECHNICAL SPECIFICATION BASES PAGE (MARKUP)

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Turkey Point Nuclear Plant L-2018-065 Docket Nos. 50-250 and 50-251 Enclosure Page 12 of 12 ATTACHMENT 2 PROPOSED TECHNICAL SPECIFICATION BASES PAGE (MARKUP)

REVISION NO.: PROCEDURE TITLE: PAGE:

27 TECHNICAL SPECIFICATION BASES CONTROL PROGRAM 20 of 211 PROCEDURE NO.:

0-ADM-536 TURKEY POINT PLANT ATTACHMENT 2 Technical Specification Bases (Page 2 of 193) 2.1.1 (Continued)

The DNB design basis is as follows: There must be at least a 95 percent probability with 95 percent confidence that the minimum DNBR of the limiting rod during Condition I and II events is greater than or equal to the DNBR limit of the DNB correlation being used. The correlation DNBR limit is established based on the entire applicable experimental data set such that there is a 95 percent probability with 95 percent confidence that DNB will NOT occur when the minimum DNBR is at the DNBR limit.

The curves (formerly TS Figure 2.1-1) provided in the COLR show the location of points of THERMAL POWER, Reactor Coolant System pressure and average temperature for which the minimum DNBR is NO less than the design DNBR value, or the average enthalpy at the vessel exit is equal to the enthalpy of saturated liquid.

In addition, fuel centerline temperature is required to stay below the melting temperature. Consistent with these design basis requirements, a DNB correlation and peak fuel centerline temperature limits are provided as Safety Limits in this Specification. The DNB correlation and parameter value of WRB 1 and 1.17, respectively, are applicable to the pre-Extended Power Uprate (EPU) and EPU operating cycles which contain residual 15x15 DRFA fuel from previous pre-uprate cycles and which contain 15x15 Upgrade fuel at the EPU conditions. The peak 9ºF centerline temperature limit of less than 5080 ºF, decreasing by 58 ºF per 10,000 MWD/MTU of burnup, is the standard value used for Westinghouse fuel. The automatic enforcement of these Reactor Core Safety Limits is provided by the proper functioning of the reactor protection system and the steam generator safety valves.