L-2016-227, License Amendment Request Change to P /T Limit Curve and LTOP Period of Applicability

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License Amendment Request Change to P /T Limit Curve and LTOP Period of Applicability
ML17006A006
Person / Time
Site: Saint Lucie NextEra Energy icon.png
Issue date: 12/22/2016
From: Costanzo C
Florida Power & Light Co
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
L-2016-227
Download: ML17006A006 (32)
v • d • e


Text

DEC 2 2 2016

, , *** :* ~ ~ I i: '1 I * '. ' ;f '.!I L-2016-227 10 CFR 50.90 .

J * '.; ~, l . ~-,

U.S. Nuclear Regulatory Commission Document Control Desk Washington, DC '20555-0001

Subject:

St. Lucie Unit 2 Docket No. 50-389 License Amendment Request Change to P /T Limit Curve and LTOP Period of Applicability Pursuant to 10 CFR 50.90, Florida Power & Light Company (FPL) is submitting a request for an amendment to the Technical Specifications (TS) for St. Lucie Unit 2. The proposed amendment would revise the period of applicability of the pressure-temperature (P-1) limit curves from 47 EFPY to 31.98 EFPY. The low temperature overpressure protection (LTOP) requirements, which are based on the P-T limit curves, will also be changed to 31.98 EFPY The enclosure provides a description and assessment of the proposed changes, a markup of the TSs and the word processed TS changes, as well as an informational markup of the TS Bases.

This license amendment proposed by FPL has been reviewed by the St. Lucie Plant Onsite Review Group.

In accordance with 10 CFR 50.91(b)(1), a copy of the proposed license amendment is being forwarded to the .

State Designee for the State of Florida.

FPL is requesting that this be processed as a normal amendment request, with approval of the proposed amendment within one year of the submittal date. Once approved, the amendment shall be implemented within 90 days.

If you should have any questions, please contact Mr. Mike Snyder at (772) 467-7036.

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

Executed on DEC 2 2 2016 I*  : '*'.

Sincerely,

. n Christopher R. Costanzo Site Vice President",

  • 1
  • *
  • i St. Lucie Plant , *' . ..
  • Enclosure cc: NRC Region II Administrator St. Lucie Plant NRC Senior Resident Inspector Ms. Cynthia Becker, Florida Department of Health Florida Power & Light Company 6501 S. Ocean Drive, Jensen Beach, FL 34957

L-2016-227 Enclosure Page 1 of 31 License Amendment Request Change to P!T Limit Curve and LTOP Period of Applicability Table of Contents

1. DESCRIPTION OF PROPOSED CHANGE .............................................................. 2
2. BASIS AND JUSTIFICATION OF PROPOSED CHANGE ....................................... 3
3. EVALUATION ........................................................................................................... 4 3.1 CHANGING THE PERIOD OF APPLICABILITY OF THE P-T LIMIT CURVES .. :. 4 3.2 LOW-TEMPERATURE OVERPRESSURE PROTECTION (LTOP) ANALYSIS. 13 3.3 CORRECTION FACTORS ................................................................................. 13
4. NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION ................... 13
5. ENVIRONMENTAL IMPACT CONSIDERATION DETERMINATION ..................... 15
6. REFERENCES ....................................................................................................... 16 Attachments Unit 2 Technical Specification Mark-Ups ............................................................ 17 Unit 2 Technical Specification Bases Mark-Up ..................................................... 23 Unit 2 Word-Processed Technical Specifications .................................................26

L-2016-227 Enclosure Page 2 of 31

1. DESCRIPTION OF PROPOSED CHANGE The current St. Lucie Unit 2 Technical Specification reactor coolant system (RCS)

Pressure Temperature (P-T) limit curves indicate applicability up to 47 effective full power years (EFPY) of operation (Reference 1). The existing low-temperature overpressure protection (LTOP) analysis that is based upon these P-T limit curves is also applicable up to 47 EFPY. The period of applicability is based on the projected Reference Temperature for Nil Ductility Transition (RT Nor) of the limiting reactor vessel materials at the time of the analysis. Analysis of the most recent reactor vessel surveillance capsule indicated the adjusted reference temperature (ART) values that are the basis for the current P-T limit curves and LTOP analysis, will be reached at approximately 31.98 EFPY.

The proposed change of the P-T limit curves from 47 EFPY to 31.98 EFPY is based upon fluence projections, limiting material embrittlement predictions benchmarked by the latest surveillance capsule results, and the Limiting Conditions for Operation (LCOs). These P-T limit curves will ensure that all RCS components will be able to withstand the effects of transient loads due to system temperature and pressure changes without their functions or performance being impaired. These loads are introduced by normal load transients, reactor trips, and startup and shutdown operations.

The LTOP system, provided by the power operated relief valves (PO RVs), and provided by the shutdown cooling (SOC) relief valves when the SOC system is in operation (Reference SL Lucie Unit 2 UFSAR 5.2.6, Amendment No. 19), ensures RCS over pressurization below certain temperatures is prevented, thus maintaining reactor coolant pressure boundary integrity. Since the LTOP analysis is based on the P-T limit curves which are unchanged, the period of applicability of the existing LTOP analysis for 47 EFPY has also been changed to 31.98 EFPY.

The applicable LCOs and the proposed changes are as follows:

a. LCO 3.4.9.1 currently provides the pressure and temperature limits during heatup, cooldown, criticality, and inservice leak and hydrostatic testing in accordance with Figures 3.4 2 and 3.4 3 for the RCS (except the pressurizer) applicable up to 47 EFPY. Tables 3.4-3 and 3.4-4 provide the low temperature RCS overpressure protection range and the minimum cold leg temperature for PORV use for LTOP, respectively, both tables are also applicable up to 47 EFPY.

The footnote applicable to LCO 3.4.1.3 references Table 3.4-3 for the cold leg temperature conditions to start a reactor coolant pump during hot shutdown to limit pump binding during heatup. LCO 3.4.1.3 requires at least two loop(s)/train(s) operable, and it defines four possible configurations of which at least two configurations are required to be operable.

L-2016-227 Enclosure Page 3 of 31 The proposed amendment would use these existing figures, as-is, and only revise the period of applicability from "47 EFPY" to "31.98 EFPY." A note is added to Figures 3.4-3 and 3.4-4 to identify the limiting material ART values used in the analysis of record to be consistent with the Unit 1 equivalent figures. This note has no operational impact. -

b. LCO 3.4.9.3 currently provides LTOP requirements during Modes 4, 5 and 6 and two of the conditions within it references Table 3.4-3 and Table 3.4-4. LCO 3.4.9.3 requires at least one overpressure protection system operable if the RCS is not depressurized and not vented by at least 3.58 square inches, and it defines three possible configurations of which at least one is required to be operable. Two of the possible configurations allow use of the PORVs to satisfy LTOP when the RCS cold leg temperature is greater than the temperature listed in Table 3.4-4 to limit the potential mass transient. When none of the overpressure protection devices are operable within eight hours, the action within LCO 3.4.9.3 requires the RCS to be above the temperatures in Table 3.4-3 or to restore the overpressure protection device .to operable or vent the RCS.

The proposed amendment would use these existing tables, as-is, and only revise the period of applicability from "47 EFPY" to "31.98 EFPY."

The marked up Technical Specification changes are provided in Attachment 1 of this Enclosure.

2. BASIS AND JUSTIFICATION OF PROPOSED CHANGE The current P-T Limit curves and LTOP analysis will expire at 31.98 EFPY based on the updated projection to reach the limiting adjusted reference temperature (ART) material value used in the P-T Limit curve analysis. The period of applicability is being updated due to the test results and evaluation of the most recent surveillance capsule 97° from St. Lucie Unit 2. The surveillance capsule tests resulted in a revised chemistry factor used for future projections of ART. Although the ART is unchanged for the P-T Limit curves and LTOP setpoint analysis of record, the time to reach the ART value is reduced, and therefore the period of applicability is also reduced. This determination was based on the current ASME methods, Regulatory Guides, and Regulatory requirements.

The basis for the new chemistry factor, period of applicability, and proposed change is provided in WCAP-17939-NP (Reference 3), which contains the supporting evaluation.

The fluence projections were prepared using the guidance of Regulatory Guide 1.190, (Reference 11 ). The reactor pressure vessel beltline P-T limits are based upon the irradiation damage prediction methods of Regulatory Guide 1.99 Revision 2 (Reference I

2).

L-2016-227 Enclosure Page 4 of 31

3. EVALUATION The basis and justification to change the period of applicability from 47 EFPY to 31.98 EFPY for the P-T limit curves and LTOP requirements is provided below.

3.1 CHANGING THE PERIOD OF APPLICABILITY OF THE P-T LIMIT CURVES P-T Limit Curve and LTOP Calculation Method:

The current P-T limit curves and LTOP analysis were developed using the requirements of 10 CFR 50, Appendix G. The P-T limit curves are based upon the irradiation damage prediction methods of NRC Regulatory Guide 1.99 Revision 2 (Reference 2).

The plant specific material input to the Appendix G lower bound fracture toughness curve/methodology is performed by determining the adjusted reference temperature (ART) of the limiting reactor vessel

  • beltline material at the end of the period of applicability. ART predictions use: initial material test properties, material chemistry, surveillance capsule results, fluence and margin, and they are the only plant specific material inputs that are considered in the P-T limit curve analysis. The ART values are calculated using the prediction methods of Reference 2.

The method to calculate the new applicability period of the P-T limit curves and by extension the LTOP setpoints is determined by Items a-h shown below:

Item a. Perform credibility evaluation of the surveillance capsule data, includes data from Capsule 97°:

The data from the three capsules, including the recently tested 97° surveillance capsule, was found to meet all five credibility criteria of Reference 2. A credibility analysis was documented in Reference 3 and deemed credible. The results of the credibility analysis are summarized below:

Criterion 1 (Reference 2) is met because the St. Lucie Unit 2 surveillance program was developed to the requirements of ASTM.E185-73. The intermediate shell plate M-605-1 has the highest initial ART or RT NOT value, and it has the second highest Copper weight percentage value. It is defined as the most limiting plate material when compared to all other base and weld materials within the reactor vessel beltline region. The weld metal heat # 83637 in the surveillance capsule is within the reactor vessel beltline but it is not limiting at the end of life (EOL) due to its low chemistry factor (low copper & low nickel content).

L-2016-227 Enclosure Page 5 of 31 Criterion 2 (Reference 2) is met because the scatter in the surveillance capsule data, is small enough to permit determination of the 30 ft-lb temperature and the Upper Shelf Energy (USE).

Criterion 3 (Reference 2) is met by comparing the data from all three surveillance capsules tested using transversely and longitudinally oriented Charpy data to a best fit line using the sum of the squares method. This uses all available data.

Table 1: Chemistry Factor Determination for St. Lucie Unit 2 Weld and Plate Material using Transversely and Longitudinally Oriented Charpy Data Shift in L'.lRTNoT Credible?

Capsule Scatter (Measured IShift Measured Fluence Fluence FFx Predicted Criteria - L'.lRTNoTI Capsule I L'.lRTNoT (x10 19 Factor L'.lRTNoT L'.lRTNoT at 1o Predicted) <17°F Material (oF) n/cm 2 ) (FF) FF 2 (oF) (oF) (oF) (oF) (Plate) 83°/

Intermediate Shell Plate (M-605-1)

Transverse 29.4 0.140 0.488 0.238 14.33 50.01 17 -20.61 No 263°/

Intermediate Shell Plate (M-605-1)

Transverse 102.7 1.00 1.000 1.000 102.70 102.58 17 0.12 Yes 83°/

Intermediate Shell Plate (

(M-605-1)

LonQitudinal 45.11 0.140 0.488 0.238 21.99 50.01 17 -4.90 Yes 97°/

Intermediate -

Shell Plate (M-605-1)

Transverse 127.6 2.25 1.220 1.487 155.62 125.10 17 2.50 Yes 97°/

Intermediate Shell Plate (M-605-1)

LonQitudinal 132.7 2.25 1.220 1.487 161.84 125.10 17 7.60 Yes Sum: 4.450 456.48 Best Fit CFM-6os-1 =1: (FF x ~RTNDT) I 1: (FF 2

) =(456.48) I (4.450) =102.58 °F

L-2016-227 Enclosure Page 6 of 31 Shift in ARTNDT Credible?

Capsule Scatter (Measured !Shift Measured Fluence Fluence FFx Predicted Criteria - ARTNoTI Capsule I ART No~ (x10 19 Factor ARTNDT ARTNDT at 1o Predicted) <28°F Material (oF) n/cm 2 ) (FF) FF2 (oF) (oF) (oF) (oF) (Weld) 83°/

Surveillance Weld (Heat

  1. 83637) 15.8 0.140 0.488 0.238 7.70 11.53 28 4.27 Yes 263°/

Surveillance Weld (Heat

  1. 83637) 26.5 1.00 1.000 1.000 26.50 23.65 28 2.85 Yes 97°/

Surveillance Weld (Heat

  1. 83637) 24.8 2.25 1.220 1.488 30.25 28.84 28 -4.04 Yes Sum: 2.725 64.45 Best Fit CFweld 83637 =~(FF x L\RTNDT) I l:(FF 2

) =(64.45) I (2. 725) =23.65 °F Although the surveillance weld material was also credible, the reduced margin term of 28 °F is not needed, because the St. Lucie Unit 2 reactor vessel is plate-limited as shown in Table 1.

The plate material scatter was determined to be less than 17°F for four of the five surveillance data points shown in Table 1. The one data point that falls outside of the one sigma (1 o) scatter criteria is by 3.61°F (20.61 °F - 17 °F), which is a small amount considering that the band in the scatter criteria is 34 °F (+/- 17°F) and the predicted ARTNDT is conservatively higher than the measured ARTNDT value. Furthermore, the two data points that are at fluence, levels closer to the actual current vessel fluence and projected end of 40-year life fluence have very minimal scatter (2.50 °F and 7.60 °F).

For a one sigma standard deviation, it would encompass 68 percent of the data. The four out of five surveillance data points (or 80 percent of the data) fall inside the one sigma standard deviation 17 °F scatter criteria. Since comparisons of the measured and predicted shifts for the two most recent data points have minimal scatter, it is reasonable to state that the surveillance plate data is deemed credible per the third criterion, (Reference 3).

Criterion 4 (Reference 2) is met because the surveillance capsules are positioned near the reactor vessel inside wall so that the irradiation conditions of the test specimens

L-2016-227 Enclosure Page 7 of 31 resemble, as closely as possible, the irradiation of the reactor vessel. The location of the specimens with respect to the reactor vessel beltline provides assurance that the reactor vessel wall and the specimens experience equivalent operating conditions such that the temperatures will not differ by more than 25°F.

Criterion 5 (Reference 2) is met since the predicted .!lRTNOT for the Standard Reference Mate.rial (SRM) plate is within one standard deviation (17 °F) of the measured ilRTNOT*

Multiple data points are not required to meet this criterion.

Based on the above-mentioned five criteria, the surveillance capsule data from the three capsules 83°, 263°, and 97° are credible, (Reference 3).

Item b. Review current and previous analyses to identify the limiting ART values for 1/4T and 3/4T used in the current P-T limit curve and L TOP analysis:

The current 47 EFPY P-T limit curves and LTOP analysis are based on a limiting material ART of 160°F at 1/4T and 137 °F at 3/4T for the intermediate shell plates M-605-1 & M-605-2, (References 7 and 10).

Item c. Update the accumulated fluence and projections for the reactor vessel beltline materials:

The neutron transport and dosimetry evaluation methodologies follow the guidance of Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," (Reference 11) and are consistent with the NRC-approved methodology described in WCAP-14040-A, Revision 4, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves".

The peak clad/base metal interface vessel fluence or surface fluence at the end of Cycle 20, which ended in March 2014, is 1.73 x 1019 n/cm2 (Reference 3, Table 6-2) with 25.55 EFPY. Vessel fluence projections at the inside diameter of the reactor vessel, clad to base metal interface, are provided for 32, 36, 40, 48, 55, and 60 EFPY at a reactor power level of 3020 MWt. The values of accumulated fluence at the end of Cycle 20 and fluence projections through 60 EFPY are shown in Reference 3 and discussed in more detail within Item g below.

Item d. Evaluate all surveillance results for the St. Lucie Unit 2 beltline materials and consider appropriate chemistry factor and margin term to be applied:

The new CF of 102.58 °F, shown in Table 1 above, is used to determine the new peak neutron fluence or surface fluence.

Item e. Select the most limiting material:

L-2016-227 Enclosure Page 8of31 At 47 EFPY the peak neutron fluence was determined to be 3.67 x 10 19 n/cm 2 at the inside diameter of the reactor vessel, clad to base metal interface in the beltline region, and the intermediate shell plate, M-605-1, was defined as the representative limiting beltline material. The higher of the two ART values, calculated at 1/4T and 3/4T and shown in bold font below, was applied for the limiting ART value for the 47 EFPY period of applicability (EPU conditions), (Reference 7). To ensure the current P-T limit curves are bounding, the limiting ART value is unchanged and used to determine the new surface fluence (defined in Item f).

Initial RTNOT (M-605-1) = +30 °F (Reference 7)

Margin = 17 °F (Reference 7)

ART (M-605-1 1/4T at 47 EFPY) = 160 °F (Reference 7)

ART (M-605-1 3/4T at 47 EFPY) = 137 °F (Reference 7)

ART= Initial RTNOT+ llRTNOT+ Margin (Equation 1 in Reference 2)

Rearranging the above equation and solving for llRTNOT yields the following:

.llRTNOT = 160 °F - 30°F - 17 °F = 113 °F Item f. Determine the "limiting" surface fluence associated with the limiting ART value in the current analysis with updated chemistry factor:

Equation 2 in Reference 2 is used to determine the new surface fluence to reach the limiting ART value at the 1/4T location. The 1/4T ART value is unchanged, 160°F, and the new chemistry factor (CF) is used to solve for surface fluence by iteration:

CF = 102.58 °F llRTNoT =(CF) f ( 0 0*10 logf) (Equation 2 in Reference 2)

Rearranging Equations 1 and 2 in Reference 2 and entering the new CF defined above, yields:

113 oF = (102.58 oF) f(0.28-0.101ogf)

Using Figure 1 from Reference 2 and solving for fluence (f) by iteration, the fluence at the 1/4T location is approximately 1.44 x 1019 n/cm 2. The new surface fluence (fsurr) is calculated using Equation 3 in Reference 2, shown below:

f = fsurf (e -0*24x) (Equation 3 in Reference 2)

Rearranging the equation above and using a nominal reactor vessel thickness of 8.625 inches (Reference 3.) yields:

fsurt = f I ( e -0.24 x 2.15625) = 1.44x1019 I 0.5960

L-2016-227 Enclosure Page 9of31 fsurf = 2.42 x 1019 n/cm 2 This 1/4T neutron fluence value results in a new peak surface fluence of 2.42 x 1019 n/cm 2 ; this yields the limiting ART value in the existing P-T curves analysis.

Item g. Determine the EFPY that corresponds to the limiting surface fluence:

The neutron fluence evaluations are based on fuel-cycle-specific information applied to a three-dimensional geometrical reactor model. The models include the core, the reactor internals, the surveillance capsules, the pressure vessel cladding and vessel wall, the insulation external to the pressure vessel, and the primary biological shield wall. The neutron fluence is determined at the octant symmetric surveillance capsule positions. A representative average fluence exposure is determined and projected info the future for the surveillance capsules removed and still inside the reactor core (referred to as "standby" capsules). Fluence is calculated at the reactor vessel inner radius (clad/base metal interface) at four azimuthal locations.

The fluence forecasts are based on projections of the neutron exposure of the reactor pressure vessel for operating periods extending to 60 effective full-power years (EFPY) at 3020 MWt. This takes into account the extended power uprate (EPU), which resulted in a net thermal power increase from 2700 MWt to 3020 MWt, 11.85% net increase. The fluence rate (and fluence) projections were conservatively increased by a factor of 10%

to allow for variations in future core power distributions, (Reference 3). The exception are the fluence increases projected in Cycle 22 and Cycle 23 based on placing new fuel at the core periphery during both cycles; therefore a penalty factor is applied to the fluence projections at Cycle 22 and Cycle 23, discussed further below.

This bounds the assumption that future core loadings will be low leakage and will not deviate more than a few percentages from the equilibrium cycle. Any larger perturbations that alter the current fluence projections would require prompt report to the NRC as provided in 10 CFR Part 50.61, (Reference 5).

The fluence projections form the basis to determine the new period of applicability for the P-T limit curves. Projections of EFPY and maximum surface fluence through the EOL are shown below in Table 2, (Reference 3, Table 6-2):

Table 2: EFPY and Fluence Projection Data Maximum Maximum Surface Cycle Surface Fluence Fluence I EFPY Identifier (Note 1) EFPY (x10 19 n/cm 2 ) (x 10 17 n/(cm 2 -EFPY))

1 20 25.55 1.73 N/A 2 21 26.93(Note 2) 1.85 8.696 3 Future 32.00 2.33 (Note 3) 9.467

L-2016-227 Enclosure Page 10 of 31 Note 1: Cycles deemed as "Future" are based on fluence rate (and fluence) projections that were increased by a factor of 10% to allow for variations in future core power distributions, (Reference 3).

Note 2: The value of 26.93 EFPY reported in Reference 3 was an estimate prior to the end of Cycle 21. The actual cumulative EFPY is 26.830 EFPY per the operating burnup data database in-use.

19 2 Note 3: The value of 2.33 x 10 n/cm reported in Reference 3 does not include the penalty factor applied to surface fluence in Cycles 22 and 23 in Table 3.

The maximum slope or surface fluence/EFPY of 9.467 x 1017 n/ (cm 2- EFPY) is determined from using the values from Table 2, Identifiers 2 and 3. The value is shown in Table 2, and it is calculated below:

Slope= (2.33 x 1019 n/cm 2 - 1.85 x 1019 n/cm 2 ) I (32.00 EFPY - 26.93 EFPY)

Slope= 9.467 x 1017 n/(cm 2-EFPY)

A penalty factor of 1.3653 (36.53%) is applied to the Cycle 22 and Cycle 23 projected cycle surface fluence in Table 3 below due to placing new fuel assemblies instead of burnt fuel assemblies at the core periphery during both cycles to address industry operating experience with grid-to-rod fretting fuel failure susceptibility. Beyond Cycle 23, St. Lucie Unit 2 will have a new grid design in place with burnt fuel assemblies placed at the core periphery. Therefore, the penalty factor is not applied to fluence projections beyond Cycle 23. Table 3 below shows the EFPY determination at the limiting surface fluence value.

Table 3: Determination of EFPY at the Limiting Surface Fluence of 2.42x10 19 n/cm 2 (Limiting Values in Bold)

Cycle Cycle Length Projected Cumulative Cycle (EFPY) Surface Surface Approximate Length (Note Cumulative Fluence, f surt Fluence, Penalty Cycle End 2

Cycle (EFPH) 2) EFPY (n/cm 2 ) fsurf (n/cm ) Factor Date 20 N/A 1.23 25.55 N/A 1.73E+19 N/A N/A 21 N/A 1.28 26.83 N/A 1.85E+19 N/A N/A 11378 (Note 22 1) 1.30 28.13 1.6742E+18 2.02E+19 1.3653 Spring 2017 23 12400 1.38 29.51 1.7382E+18 2.19E+19 1.3653 Fall 2018 24 12400 1.38 30.89 1.2731 E+18 2.32E+19 1.0 Spring 2020 Mid Cycle 25 9847 1.10 31.98 1.01E+18 2.42E+19 1.0 -May 2021 25 N/A N/A NIA N/A N/A 1.0 Fall 2021 Note 1: EFPH at Cycle 22 uses actual EFPH from cycle start up to 11-2-16 and assumes a 100%

capacity factor for the remainder of the cycle.

Note 2: EFPY at Cycles 23, 24, and 25 assume a 97.5% capacity factor.

L-2016-227 Enclosure Page 11 of 31 The projected surface fluence for Cycles 22, 23, 24, and 25 are calculated as follows:

Slope= 9.467 x 10 17 n/(cm2 -EFPY)

Cycle Projected fsurf = EFPY x Slope x Penalty Factor The limiting cumulative surface fluence (2.42 x 1019 n/cm 2) is reached at 31.98 EFPY in the middle of Cycle 25, approximately during May 2021 as shown in Table 3 above.

l Item h. Project the ART at 1/4T and 3/4T for all beltline materials to show that the P-T limit curves bound all data for the new period of applicability:

The current P-T limit curves are based on a limiting ART values at 1/4T .and 3/4T of 160°F and 137°F (References 7 and 10). In Table 4 below the location, ID#, Copper content (Cu %), Nickel content (Ni %), tabulated chemistry factor (Table CF), Initial RT Nor, and Margin Method regulatory position applied from Reference 2 (Margin Method) data are contained in Reference 10. The calculated chemistry factor (Calculated CF) in Table 4 is calculated in Table 1 above. Using the new chemistry factor of 102.58 °F, a margin term of 17 °F, 2.42 x 1019 n/cm 2 surface fluence value at 31.98 EFPY, and the methodology in Reference 2 to calculate ART (equations also described above), the projected ART values are recalculated in Table 4 below. Table 4 shows the projected limiting ART values do not exceed the limiting ART values used in the current P-T limit curves; therefore, the current P-T limit curves remain bounded.

L-2016-227 Enclosure Page 12 of ;31 Table 4: Projected ART Values at 31.98 EFPY Fluence (Limiting Values in Bold Font) 31.98 EFPY 31.98 31.98 Initial Peak 1/4T 3/4T EFPY EFPY Calculated Table RTNDT Margin Fluence, Fluence Fluence ARTat ARTat 2

Location ID# Cu% Ni% CF CF (oF) Margin Method (n/cm ) (n/cm 2 ) (n/cm 2 ) 1/4T 3/4T Lower shell plate B-8307- Position (M4116-1) 2 0.06 0.57 37.0 20 °F 34 °F 1.1 2.42E+19 1.44E19 5.12E18 95 °F 84 °F Lower shell plate A-3131- Position (M4116-2) 1 0.07 0.60 44.0 20 °F 34 °F 1.1 2.42E+19 1.44E19 5.12E18 102 °F 90 °F Lower shell plate A-3131- Position (M4116-3) 2 0.07 0.60 44.0 20 °F 34 °F 1.1 2.42E+19 1.44E19 5.12E18 102 °F 90 °F Int. shell plate (M- A-8490- Position 605-1) 2 0.11 0.61 74.2 30 °F 34 °F 1.1 2.42E+19 1.44E19 5.12E18 146 °F 124 °F Int. shell plate (M- A-8490- Position 605-1) 2 0.11 0.61 102.58 30 °F 17 °F 1.2 2.42E+19 1.44E19 5.12E18 160 °F 130 °F Int. shell plate (M- B-3416- Position 605-2) 2 0.13 0.62 91.5 10 °F 34 °F 1.1 2.42E+19 1.44E19 5.12E18 145 °F 118 °F Int. shell plate (M- A-8490- Position 605-3) 1 0.11 0.61 74.2 0°F 34 °F 1.1 2.42E+19 1.44E19 5.12E18 116 °F 94 °F Int. shell plate (M- A-8490- Position 605-3) 1 0.11 0.61 102.58 0°F 17 °F 1.2 2.42E+19 1.44E19 5.12E18 130 °F 100 °F Lower shell axial welds (101- 56.0 Position 142A,B,C) 83637 0.05 0.07 34.1 -50 °F OF 1.1 2.42E+19 1.44E19 5.12E18 44 °F 34 °F Int. shell axial welds (101- 65.5 Position 124A,B,C) 83642 0.05 0.09 36.4 -56 °F OF 1.1 2.42E+19 1.44E19 5.12E18 50 °F 39 °F Int. shell axial welds (101-124C 83642/ 56.0 Position Repair) 83637 0.05 0.07 34.1 -50 °F OF 1.1 2.42E+19 1.44E19 5.12E18 44 °F 34 °F Int. to Lower girth 83637 I Position welds (101-171) 3P7317 0.07 0.07 40.1 -50 °F 56 °F 1.1 2.42E+19 1.44E19 5.12E18 50 °F 39 °F L

L-2016-227 Enclosure Page 13 of 31 The fluence is attenuated through the vessel wall using the nominal reactor wall thickness of 8.625". The cladding thickness is neglected in the ART projections in Table 4. Table 4 shows the limiting beltline material is the intermediate shell plate, M-605-1.

3.2 LOW-TEMPERATURE OVERPRESSURE PROTECTION (LTOP) ANALYSIS The low-temperature overpressure protection (LTOP) system is designed to ensure that the 10 CFR Part 50, Appendix G, pressure vessel brittle fracture limits will not be exceeded in Modes 4, 5, or 6 under over-pressurization conditions, (Reference 7). The current LTOP analysis is based on the guidance provided in USN RC Standard Review Plan (SRP) 5.2.2 Revision 3, "Overpressure Protection" and in accordance with the requirements set forth in NRC Branch Technical Position RSB 5-2. The LTOP analysis determines the LTOP setpoints, heatup and cooldown rates, and administrative requirements, (Reference 10). The objective of the current LTOP analysis is to preclude violation of the P-T limit curves during startup and shutdown conditions. The LTOP analysis remains unchanged by the applicability period change from 47 EFPY to 31.98 EFPY because the P-T limit curves are not being changed. Therefore, it is not necessary to re-analyze or modify the LTOP system setpoints.

The fluence calculation results shown in Section 3.1 Item g change the LTOP analysis applicability period from 47 EFPY to 31.98 EFPY.

The P-T limit curve analysis for 47 EFPY also provide the setpoints for the lowest service temperature, minimum boltup temperature and the minimum pressure limits on the P-T limit curves for reference. These limits also remain unchanged and the applicability period is also changed from 47 EFPY to 31.98 EFPY.

3.3 CORRECTION FACTORS Correction factors are not affected by neutron fluence, and therefore, remain unchanged by the period of applicability change from 47 EFPY to 31 :98 EFPY.

4. NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION The standards used to arrive at a determination that a request for amendment involves no significant hazards consideration are included in the Commission's regulation, 10 CFR 50.92, which states that no significant hazards considerations are involved if the operation of the facility in accordance with the proposed amendment would not (1)

' involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety. Each standard is discussed below:

_J

L-2016-227 Enclosure Page 14 of 31 (1) Operation of the facility in accordance with the proposed amendment would not involve a significant increase in the probability or consequences of an accident previously evaluated.

The pressure temperature (P-T) limit curves in the Technical Specifications are conservatively generated in accordance with the fracture toughness requirements of 10 CFR 50 Appendix G as supplemented by the ASME Code Section XI, Appendix G recommendations. The Adjusted Reference Temperature (ART) values are based on the Regulatory Guide 1.99, Revision 2 shift prediction and attenuation formula and have been validated by a credible reactor vessel surveillance program. There are no changes to the P-T limit curves, only a change in the period of applicability based on more recent fluence predictions and new best estimate chemistry information. Based on the current fluence projections, analysis has demonstrated that the current P-T limit curves will remain conservative for up to 31.98 EFPY.

In conjunction with reducing the period of applicability of the existing P-T limit curves, the low temperature overpressure protection (LTOP) analysis for 47 EFPY is also reduced to 31.98 EFPY. The LTOP analysis confirms that the current setpoints for the power operated relief valves (PO RVs) will provide the-,

appropriate overpressure protection at low Reactor Coolant System (RCS) temperatures: Because the P-T limit curves remain unchanged, the existing LTOP values are unchanged, which include the PORV setpoints.

The P-T limit curves and LTOP analysis have not changed; therefore, the proposed amendment does not represent a change in the configuration or operation of the plant. The results of the existing LTOP analysis have not changed, and the limiting pressures for given temperatures will not be exceeded for the postulated transients. Assurance is provided that reactor vessel integrity will be maintained. The proposed amendment does not involve an increase in the probability or consequences of accidents previously evaluated.

(2) Operation of the facility in accordance with the proposed amendment would not create the possibility of a new or different kind of accident from any accident previously evaluated.

The requirements for the P-T limit curves and LTOP limits have been in place since the beginning of plant operation. The only changes to these curves are the reduction of the period of applicability (in EFPY), which is based on a new chemistry factor, new fluence data, and the operating time (in EFPY) required to reach the same limiting adjusted reference temperature (ART) projection used for the current 47 EFPY P-T limit curves. Since there is no change in the configuration or operation of the facility as a result of the proposed amendment, ,

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

L-2016-227 Enclosure Page 15 of 31 (3) Operation of the facility in accordance with the proposed amendment would not involve a significant reduction in a margin of safety.

Analysis has demonstrated that the fracture toughness requirements of 10 CFR 50 Appendix G are satisfied and that conservative operating restrictions are maintained for the purpose of low temperature overpressure protection. The P-T limit curves will provide assurance that the RCS pressure boundary will behave in a ductile manner and that the probability of a rapidly propagating fracture is minimized. Therefore, operation in accordance with the proposed amendment would not involve a significant reduction in a margin of safety.

Summary: Based on the above discussion and the analysis performed, Florida Power

& Light Co. (FPL) has determined that the amendment request does not (1) involve a significant increase in the probability or consequences of an accident previously evaluated, (2) create the probability of a new and different kind of accident from any accident previously evaluated, or (3) involve a significant reduction in a margin of safety; and therefore does not involve a significant hazards consideration.

5. ENVIRONMENTAL IMPACT CONSIDERATION DETERMINATION 10 CFR 51.22(c)(9) provides criteria for identification of licensing and regulatory actions eligible for categorical exclusion from performing an environmental assessment. A proposed amendment to an operating license for a facility requires no environmental assessment if operation of the facility in accordance with the proposed am*endment does not:
i. involve a significant hazards consideration, ii. result in a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, and iii. _ result in a significant increase in individual or cumulative occupational radiation exposure.

The proposed amendment involves no significant increase in the amounts of any effluents and no significant change in the types of any effluents that may be released offsite, and no significant increase in individual or cumulative occupational radiation exposure. FPL has concluded that the proposed amendment involves nci significant hazards consideration and meets the criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9) and that pursuant to 10 CFR 51.22(b) an environmental impact statement or environmental assessment need not be prepared in connection with issuance of the amendments.

L-2016-227 Enclosure Page 16 of 31

6. REFERENCES
1. St. Lucie Unit 2, Technical Specification through Amendment 180, dated March 7, 2016.
2. US NRC Regulatory Guide 1.99, "Radiation Embrittlement of Reactor Vessel Materials," Revision 2, May 1988
3. Westinghouse Report# WCAP-17939, "Analysis of Capsule 97° from the Florida Power & Light Company St. Lucie Unit 2 Reactor Vessel Radiation Surveillance Program," May 2015 [USNRC ADAMS Accession #s ML151548077, ML15154B079, ML15154B080]
4. Deleted
5. US NRC Letter to FPL ADAMS Accession# ML090060049, "St. Lucie Plant Unit No. 2 - Issuance of Amendment Regarding Pressure Vessel Fluence to 55 Effective Full-Power Years of Operations (TAC NO. MD8040)," January 29, 2009
6. Deleted
7. US NRC Letter to FPL ADAMS Accession# ML12235A463, "St. Lucie Plant Unit 2 - Issuance of Amendment Regarding Extended Power Uprate (TAC NO.

ME5843)," September 24, 2012

8. Deleted
9. Deleted
10. FPL Letter to US NRC ADAMS Accession # ML080290135, "Update PT Curve and LTOP for 55 EFPY," January 23, 2008 [FPL Letter# L-2007-198]
11. NRC Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," March 2001

L-2016-227 Enclosure Page 17 of 31 Attachment 1 Unit 2 Technical Specification Mark-Ups

L-2016-227 Enclosure Page 18 of 31 Attachment 1 LIST OF FIGURES FIGURE PAGE 2.1-1 REACTOR CORE THERMAL MARGIN SAFETY LIMIT LINES FOUR REACTOR COOLANT PUMPS OPERATING ........................................................... 2-3 2.2-1 LOCAL POWER DENSITY - HIGH TRIP SETPOINT PART 1 (FRACTION OF RATED THERMAL POWER VERSUS OR2) .................................. 2-7 2.2-2 LOCAL POWER DENSITY - HIGH TRIP SETPOINT PART 2 (QR2 VERSUS Y1) .....................................................................................................2-8 2.2-3 THERMAL MARGIN/LOW PRESSURE TRIP SETPOINT PART 1 (Y 1. VERSUS A 1) ......................................................................................................2-9 2.2-4 THERMAL MARGIN/LOW PRESSURE TRIP SETPOINT PART 2 (FRACTION OF RATED THERMAL POWER VERSUS OR1) ................................ 2-10 3.1-1 MINIMUM BAMTVOLUME VS STORED BORIC ACID CONCENTRATION ... 3/41-15 3.1-1a DELETED .......................................................................................................................

3.1-2 DELETED .......................................................................................................................

3.2-1 DELETED .......................................................................................................................

3.2-2 DELETED .......................................................................................................................

3.2-3 DELETED .......................................................................................................................

4.2-1 DELETED .......................................................................................................................

3.2-4 DELETED .......................................................................................................................

3.4-1 DELETED .......................................................................................................................

3.4-2 ST. LUCIE UNIT 2 REACTOR COOLANT SYSTEM PRESSURE-INSERVICE TEST ................. ,

31 98 r*** . . . . . . . . . . . . . . . . . . . . . . . . . . .

TEMPERATURE LIMITS FOR4TEFPY, HEATUP, CORE CRITICAL, AND

_ 3/4 4-31a ST. LUCIE - UNIT 2 XXI Amendment No. 3, 63, ~. Ba, '142,,

-147, '1-54, 1£3.

L-2016-227 Enclosure Page 19 of 31 Attachment 1 LIST OF FIGURES /continued\

FIGURE PAGE 3.4-3 ST. LUCIE UNIT 2 REACTOR COOLANT SYSTEM PRESSURE-TEMPERATURE LIMITS FOR 4+-EFPY, COOLDOWN AND INSERVICE TEST ......................................1 _ 1....................................................... 3/4 4-31b 31 98 3A-4 DELETED ....................................................................................................... 3/4 4-32 4.7-1 SAMPLING PLAN FOR SNUBBER FUNCTIONAL TEST .............................. 3/4 7-25 5.1-1 SITE AREA MAP .................................................................................................... 5-2 5.6-1a DELETED ....................................................................................................................

5.6-1b DELETED ....................................................................................................................

5.6-1c DELETED ....................................................................................................................

5.6-1d DELETED ....................................................................................................................

5.6-1e DELETED ....................................................................................................................

5.6-1f DELETED ....................................................................................................................

5.6-1 ALLOWABLE REGION 1 STORAGE PATTERNS AND FUEL ARRANGEMENTS ................................................................................................ 5-4h 5.6-2 ALLOWABLE REGION 2 STORAGE PATTERNS AND FUEL ARRANGEMENTS (Sheet 1 of 3) .......................................................................... 5-4i 5.6-2 ALLOWABLE REGION 2 STORAGE PATTERNS AND FUEL ARRANGEMENTS (Sheet 2 of 3) .......................................................................... 5-4j 5.6-2 ALLOWABLE REGION 2 STORAGE PATTERNS AND FUEL ARRANGEMENTS (Sheet 3 of 3) ......................................................................... 5-4k 5.6-3 INTERFACE REQUIREMENTS BETWEEN REGION 1 AND REGION 2 (Sheet 1 of 2) ......................................................................................................... 5-41 5.6-3 INTERFACE REQUIREMENTS BETWEEN REGION 1 AND REGION 2 (Sheet 2 of 2) ....................................................................................................... 5-4m 5.6-4 ALLOWABLE CASK PIT STORAGE RACK PATTERNS ..................................... 5-4n 6.2-1 DELETED ...............................................................................................................6-3 6.2-2 DELETED ...............................................................................................................6-4 ST. LUCIE - UNIT 2 XXll Amendment No.&, 2~. 63, ~. ~.

447, 35, 454, ~

L-2016-227 Enclosure Page 20 of 31 Attachment 1 FIGURE 3.4-2 ST. LUCIE UNIT 2 REACTOR COOLANT SYSTEM PRESSURE-TEMPERATURE LIMITS FOR '4'rEFPY, HEATUP, CORE CRITICAL, AND INSERVICE TEST 31.98 2500

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1/4T, 160 °F 3/4T, 137 °F ST. LUCIE - UNlT 2 314 4*31a Amendment No. 37. 4&, 44.2,

'IM, "ffiS-

L-2016-227 Enclosure Page 21 of 31 Attachment 1 FIGURE 3.4-3 ST. LUCIE UNIT 2 REACTOR COOLANT SYSTEM PRESSURE-TEMPERATURE LIMITS FOR 2"1'-EFPY, COOLDOWN, AND INSERVICE TEST

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Limiting ART Values at 31 .98 EFPY:

1/4T, 160 °F 3/4T, 137 °F ST. LUCIE* UNIT 2 314 4-31b Amendment No. ~. 4&, ~.

-l64, 463-

L-2016-227 EnClosure Page 22 of 31 Attachment 1 TABLE 3.4-3 LOW TEMPERATURE RCS OVERPRESSURE PROTECTION RANGE Operating Cold Leg Temperature, °F Period, ~ri~ ~ri~

EFPY Heatup Coo Id own

~246 ~224 TABLE 3.4-4 MINIMUM COLD LEG TEMPERATURE FOR PORV USE FOR LTOP Cold Leg Temperature, °F Operating During During Period Heatup Cooldown EFPY 80 132 ST. LUCIE - UNIT 2 314 4-37a Amendment No. ~. ~. ~ .

-iM. ~

L-2016-227 Enclosure Page 23 of 31 Attachment 2 Unit 2 Technical Specification Bases Mark-Up

t-2016-227 Enclosure Page 24 of 31 Attachment 2 SECTION NO.;

TITLE. TECHNICAL SPECIFICATIONS 314.4 BASES ATTACHMENT 6 OF ADM-25.04 36 of 40 REVISION NO.; REACTOR COOLANT SYSTEM 16 ST. LUCIE UNIT 2 314.4 REACTOR COOLANT SYSTEM (continued)

BASES (continued) 314.4.9 PRESSUREfTEMPERA TURE LIMITS All components in the Reactor Coolant System are designed to withstand the effects of cyclic loads due to system temperature and pressure changes.

These cyclic loads are introduced by normal load transients, reactor trips, and startup and shutdown operations. The various categories of load cycles used for design purposes are provided in Section 5.2 of the UFSAR. During startup and shutdown , the rates of temperature and pressure changes are limited so that the maximum specified heatup and cooldown rates are consistent with the design assumptions and satisfy the stress limits for cyclic operation.

During heatup, the thermal gradients through the reactor vessel wall produce thermal stresses which are compressive at the reactor vessel inside surface and are tensile at the reactor vessel outside surface. Since reactor vessel internal pressure always produces tensile stresses at both the inside and outside surface locations, the total applied stress is greatest at the outside surface location . However, since neutron irradiation damage is larger at the inside surface location when compared to the outside surface, the inside surface flaw may be more limiti.ng . Consequently, for the heatup analysis both the inside and outside surface flaw locations must be analyzed for the specific pressure and thermal loadings to determine which is more limiting .

During cooldown, the thermal gradients through the reactor vessel wall produce thermal stresses which are tensile at the reactor vessel inside surface and which are compressive at the reactor vessel outside surface.

Since reactor vessel internal pressure always produces tensile stresses at both the inside and outside surface locations, the total applied stress is greatest at the inside surface location . Since the neutron indication damage is also greatest at the inside surface location the inside surface flaw is the limiting location. Consequently, only the inside surface flaw must be evaluated for the cooldown analysis.

The heatup and cooldown limit curves Figures 3.4-2 and 3.4-3 are composite curves which were prepared by determining the most conservative case, with either the inside or outside wall controlling , for any heatup rate of up to 50 degrees F per hour or cooldown rate of up to 100 degrees F per hour. The heatup and cooldown curves were prepared based upon the most limiting value of the predicted adjusted reference temperature at 47 EFPV, and they include adjustments for pressure differences between the reactor vessel beltline and pressurizer instrument taps. the latest approved period of applicability (in EFPY) as defined in Technical Specification Figures 3.4-2

'--- - - - - -----------------land 3.4*3.

L-2D16-227 Enclosure Page 25 of 31 Attachment 2 SECTION NO*. PAGE.

TITLE. TECHNICAL SPECIFICATIONS 3/4.4 BASES ATTACHMENT 6 OF ADM-25.04 37 of 40 REVISION NO.: REACTOR COOLANT SYSTEM 16 ST. LUCIE UNIT 2 .-

3/4.4 REACTOR COOLANT SYSTEM (continued) i BASES (continued) 3/4.4.9 PRESSURE/TEMPERATURE LIMITS (continued)

The reactor vessel materials have been tested to determine their initial RT NoT; the results of these tests are shown in Table B 3/4.4-1 . Reactor operation and resultant fast neutron (E greater than 1 MeV) irradiation will cause an increase in the RT NOT- An adjusted reference temperature can be predicated using a) the initial RT NOT, b) the fluence (E greater than 1 MeV),

including appropriate adjustments for neutron attenuation and neutron energy spectrum variations through the wall thickness, c) the copper and nickel contents of the material, and d) the transition temperature shift as recommended by Regulatory Guide 1.99, Revision 2, "Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials," or other approved method. The heatup and cooldown limit curves Figures 3.4-2 and 3.4-3 include predicted adjustments for this shift in RTNOT at 47 EFPY.

The actual shift in RTNOT of the vessel materials will be b the latest approyed period periodically during operation , by removing and evaluatin ~~~;!~~b~~h~~~rPY} as 10 C~R 50 Appe~d 1x Hand ASTM E185, rea.ctor vessel Specification Figures 3.4- 2 surveillance specimens installed near the ms1de wall oft and 3.4. 3.

the core area. Since the neutron spectra at the irradiatiu--......-~-~---.~

vessel inside radius are essentially identical, the measured transition temperature shift in RT NOT for a set of material samples can be compared to the predictions of RTNOT that were used for preparations of the pressure/temperature limits curves. If the measured delta RT NOT values from the surveillance capsule are not conservatively within the measurement uncertainty of the prediction method , then heat up and cooldown curves must be re-evaluated .

The pressure-temperature limit lines shown on Figures 3.4-2 and 3.4-3 for reactor criticality and for inservice leak and hydrostatic testing have been provided to assure compliance with the minimum temperature requirements for Appendix G to 10 CFR 50.

The maximum RT NOT all Reactor Coolant System pressure-retaining materials, with the exception of the reactor pressure vessel, has been determined to be 60°F. The Lowest Service Temperature limit line shown on Figures 3.4-2 and 3.4-3 is based upon this RT NOT since Article NB-2332 (Summer Addenda of 1972) of Section 111 of the ASME Boiler and Pressure Vessel Code requires the Lowest Service Temperature to be RT NOT + 100°F for piping, pumps, and valves. Below this temperature, the system pressure must be limited to a maximum of 20% of the system 's hydrostatic test pressure.

L-2016-227 Enclosure Page 26 of 31 Attachment 3 Unit 2 Word-Processed Technical Specifications

L-2016-227.

Enclosure Page 27 of 31 Attachment 3 LIST OF FIGURES 2.1-1 REACTOR CORE THERMAL MARGIN SAFETY LIMIT LINES FOUR REACTOR COOLANT PUMPS OPERATING .................... ... ............... . ... .......... 2-3 2.2-1 LOCAL POWER DENSITY - HIGH TRIP SETPOINT PART 1 (FRACTION OF RATED THERMAL POWER VERSUS OR2) .. .. ............ .... ..... ......... 2-7 2.2-2 LOCAL POWER DENSITY - HIGH TRI P SETPOINT PART 2 (QR2 VERSUS Y1) .. ....... .. .... .. ... ..... ....... ... ... ..... ... .......... ...... .... ....... ....... ........ .. ...... .. . 2-8 2.2-3 THERMAL MARGIN/LOW PRESSURE TRIP SETPOINT PART 1 (Y1 , VERSUS A 1) **- -****************** ....................... ....... ............... .. ........ ..... ...... ....... 2-9 2.2-4 TH ERMAL MARGIN/LOW PRESSURE TRIP SETPOINT PART 2 (FRACTION OF RATED TH ERMAL POWER VERSUS OR1) .. .... ....... ...... ..... ........ 2-10 3.1-1 MINIMUM BAMT VOLUME VS STORED BORIC ACID CONCENTRATION ... 3/4 1-1 5 3.1-1a DELETED .... .............. ... ......... .... .. .... ...... .. ....... .... ... ..... ..... ................. ... ... ...... ........... .

3.1-2 DELETED ..... ......... ..... ....... ............. ....... ...... .. ........... ..... ...... .................... ........ .... ....... ... .

3.2-1 DELETED ........ .... ........ ..... .. ...... ... ... ..... ...... ... ...... .... ... .... .... .. .. .... ... ... .... .................. .. ... .. ..

3.2-2 DELETED .... .. .. ... ..... ...... .... ......... .......... ....... ........ .

3.2-3 DELETED .. ... .. .... .............. ...... .... ... ............ ....... ... .

4.2-1 DELETED .................................... ........ .. .... .. ..... .... .................... .... ............... ... ... ....... .

3.2-4 DELETED ............ ..... ...... ...... ..... .... .... ... .. .. ...... ............... ......... ... ..... .. ....... .

3.4-1 DELETED ....... ... ................... .... .. .......... ................... ... .... ............ .... .. ........ ....... .... ........ ...

3.4-2 ST. LUCIE UNIT 2 REACTOR COOLANT SYSTEM PRESSURE-TEMPERATURE LIMITS FOR 31 .98 EFPY, HEATUP, CORE CRITICAL, AND INSERVICE TEST .. ..... .................... .............. ........ ..... .......... ... .. ....... ....... 3/4 4-31a ST. LUCIE - UNIT 2 XXI Amendment No. 8, ea, +a, 92. 442,

~. 4-04.~

L-2016-227 Enclosure Page 28 of 31 Attachment 3 LIST OF FIGURES (continued\

FIGURE 3.4-3 ST. LUCIE UNIT 2 REACTOR COOLANT SYSTEM PRESSURE-TEMPERATURE LIMITS FOR 31.98 EFPY, COOLDOWN AND INSERVICE TEST ............. ......... ........................... ... .. ........... .. .... ........ ... ... 3/4 4-31b 3.4-4 DELETED ...................... ... ............................. ....................................... ...... ... 3/4 4-32 4.7-1 SAMPLING PLAN FOR SNUBBER FUNCTIONAL TEST ...... 314 7-25 5.1-1 SITE AREA MAP ........ ..... ................ ...... ......... ...... . ..... .............. ..... ...... 5-2 5.6-1 a DELETED ... ...... .. .... ................. ....... ..... ....... ... .......... .. ............................ . ...... .......... ..

5.6-1b DELETED .. .................. .. ... .. ... .............. .......................... .. .... ...... .. ... .. ......... ... .... .. .... . ..

5.6-1 c DELETED ....... .. .... .... ..... .. ...... ..... ..... ... ....... .... .. .. .... ... ... ... .. .... .... ....... ...... .. .... ... ..... ....... .

5.6- 1d DELETED ... .... ... . .... ...... .................. ......... ...... ...... ... ......... ... ....... .... ..... .. ..... ........ ....... .

5.6-1e DELETED ........ ...

5.6-1f DELETED .... ....................... ..... ...... .... ......... .............. ......... .. .. .. .... ... ...... ...... ........ ... .... .

5.6-1 ALLOWABLE REGION 1 STORAGE PATTERNS AND FUEL ARRANGEMENTS ..... ....... .. ... ............... .............. ... .. .. .... .. ........... .... ... ................... 5-4h 5.6-2 ALLOWABLE REGION 2 STORAGE PATTERNS AND FUEL ARRANGEMENTS (Sheet 1 of 3) .......... ..... ... ......... ..... ....... ...... ............... .. .......... 5-4i 5.6-2 ALLOWABLE REGION 2 STORAGE PATTERNS AND FUEL ARRANGEMENTS (Sheet 2 of 3) ................. ...... ... .. ... ............... ....... . .... ....... 5-4j 5.6-2 ALLOWABLE REGION 2 STORAGE PATTERNS AND FUEL ARRANGEMENTS (Sheet 3 of 3) ...... ............ . ...... ..... .... ........................... .. ....... 5-4k 5.6-3 INTERFACE REQUIREMENTS BETWEEN REGION 1 AND REGION 2 (Sheet 1 of 2) ..... ......... ..... .. .... .. ..... ....... ... ... .. ...... ......... .... ... ... ... ..... .................. ....... 5-41 5.6-3 INTERFACE REQUIREMENTS BETWEEN REGION 1 AND REGION 2 (Sheet 2 of 2) ............. ....... ... ...... .. . .... .... .... ... ... .. ... .... .... ........ ...... .. . 5-4m 5.6-4 ALLOWABLE CASK PIT STORAGE RACK PATTERNS ................. ........... ... .. .. .. 5-4n 6.2-1 DELETED .. .... .................... ... ... .... ... .... ..... ... ... ...... ..... ... .......... .. ... ..... .. .... .. .... ... ... ... ..6-3 6.2-2 DELETED ....... .... .............. ... ... ... .... ..... ....... .. ............ ... ......................... ........... ........ 6-4 ST. LUCIE - UNIT 2 XXll Amendment No. 8, .29, 83, ~. 442,

~. ~ . 4M. ~

L-2016-227 Enclosure Page 29 of 31 Attachment 3 FIGURE 3.4-2 ST. LUCIE UNIT 2 REACTOR COOLANT SYSTEM PRESSURE-TEMPERATURE LIMITS FOR 31.98 EFPY, HEATUP, CORE CRITICAL, AND INSERVICE TEST 2500 *-,1..J_ .. .J .. _4_: .. 1.. .I- ~ 1..J; .. .J-1..1 J--L .J-L. J. ~ ...

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1/4T, 160°F 3/4T, 137°F ST. LUCIE - UNIT 2 3/4 4-31a Amendment No. :Jrl., 46, ~.

-iM.~

L-2016-227 Enclosure Page 30 of 31 Attachment 3 FIGURE 3.4-3 ST. LUCIE UNIT 2 REACTOR COOLANT SYSTEM PRESSURE-TEMPERATURE LIMITS FOR 31.98 EFPY, COOLDOWN, AND INSERVICE TEST

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0 100 200 300 400 500 Tc - INDIC.ATEO REACTOR COOLANT SYSTEM TEMPERATURE. °F

  • - Includes lnst!'\Jment Uncertalnty Limiting Material: Intermediate Shell Plate (M-605-1 )

Limiting ART Values at 31 .98 EFPY:

1/4T, 160°F 3/4T, 137°F ST. LUCIE - UNIT 2 3/4 4-31b Amendment No. a+, 46, ~ .

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L-201"6-227 Enclosure Page 31 of 31 Attachment 3 TABLE 3.4-3 LOW TEMPERATURE RCS OVERPRESSURE PROTECTION RANGE Operating Cold Leg Temperature, °F Period, During During EFPY Heatup Cooldown

.::: 31 .98 _:::246 _:::224 TABLE 3.4-4 MINIMUM COLD LEG TEMPERATURE FOR PORV USE FOR LTOP Cold Leg Temperature, °F Operating During During Period Heatup Cool down EFPY

.::: 31 .98 80 132 ST. LUCIE - UNIT 2 314 4-37a Amendment No. d-1-, 49, ++2, 4M. ~

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