L-2011-141, Response to NRC Request for Additional Information Regarding Extended Power Uprate License Amendment Request No. 205 and Reactor Systems Issues

Response to NRC Request for Additional Information Regarding Extended Power Uprate License Amendment Request No. 205 and Reactor Systems Issues

ML11137A080
Person / Time
Site:	Turkey Point
Issue date:	05/11/2011
From:	Kiley M Florida Power & Light Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
L-2011-141
Download: ML11137A080 (24)

Text

Florida Power & Light, 9760 S.W. 344 ST. Florida City, FL 33034 MAY 11 2011 L-2011-141 FlPL 10 CFR 50.90 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D. C. 20555-0001 Re:

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 Response to NRC Request for Additional Information Regarding Extended Power Uprate License Amendment Request No. 205 and Reactor Systems Issues

References:

(1) M. Kiley (FPL) to U.S. Nuclear Regulatory Commission (L-2010-113), "License Amendment Request No. 205: Extended Power Uprate (EPU)," (TAC Nos. ME4907 and ME4908), Accession No. ML103560169, October 21, 2010.

(2) Email from J. Paige (NRC) to T. Abbatiello (FPL), "Turkey Point EPU - Reactor Systems (SRXB) Request for Additional Information - Round 2", Accession No. ML111010080, April 11, 2011.

By letter L-2010-113 dated October 21, 2010 [Reference 1], Florida Power and Light Company (FPL) requested to amend Renewed Facility Operating Licenses DPR-31 and DPR-41 and revise the Turkey Point Units 3 and 4 Technical Specifications (TS). The proposed amendment will increase each unit's licensed core power level from 2300 megawatts thermal (MWt) to 2644 MWt and revise the Renewed Facility Operating Licenses and TS to support operation at this increased core thermal power level. This represents an approximate increase of 15% and is therefore considered an extended power uprate (EPU).

On March 31, 2011, a public meeting was held with the U.S. Nuclear Regulatory Commission (NRC) Project Manager (PM), applicable NRC technical reviewers, and FPL representatives to discuss proposed NRC requests for information (RAI) related to the EPU License Amendment Request (LAR). During the meeting, fourteen (14) RAI questions from the Reactor Systems Branch (SRXB) were presented. On April 11, 2011, FPL received an email from the NRC PM containing the final RAI [Reference 2]. The RAI consisted of ten (10) of the questions previously discussed in the above public meeting. The RAI questions and applicable FPL responses are documented in the Attachment to this letter.

In accordance with 10 CFR 50.91(b)(1), a copy of this letter is being forwarded to the State Designee of Florida.

This submittal does not alter the significant hazards consideration or environmental assessment previously submitted by FPL letter L-2010-113 [Reference 1].

This submittal contains no new commitments and no revisions to existing commitments.

Should you have any questions regarding this submittal, please contact Mr. Robert J. Tomonto, Licensing Manager, at (305) 246-7327.

Aoo0 ti OO (

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 L-2011-141 Page 2 of 2 I declare under penalty of perjury that the foregoing is true and correct.

ExecutedonMay I(,2011.

Very truly yours,

/'A -16 ý e ýr Michael Kiley Site Vice President Turkey Point Nuclear Plant Attachment cc:

USNRC Regional Administrator, Region II USNRC Project Manager, Turkey Point Nuclear Plant USNRC Resident Inspector, Turkey Point Nuclear Plant Mr. W. A. Passetti, Florida Department of Health

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 L-2011-141 Attachment Page 1 of 22 Turkey Point Units 3 and 4 RESPONSE TO NRC RAI REGARDING EPU LAR NO. 205 AND SRXB REACTOR SYSTEMS ISSUES ATTACHMENT

Turkey Point Units 3 and 4 L-2011-141 Docket Nos. 50-250 and 50-251 Attachment Page 2 of 22 Response to Request for Additional Information The following information is provided by Florida Power and Light Company (FPL) in response to the U. S. Nuclear Regulatory Commission's (NRC) Request for Additional Information (RAI).

This information was requested to support License Amendment Request (LAR) 205, Extended Power Uprate (EPU), for Turkey Point Nuclear Plant (PTN) Units 3 and 4 that was submitted to the NRC by FPL via letter (L-2010-113) dated October 21, 2010 [Reference 1].

On March 31, 2011, a public meeting was held with the U.S. Nuclear Regulatory Commission (NRC) Project Manager (PM), applicable NRC technical reviewers, and FPL representatives to discuss proposed NRC requests for information (RAI) related to the EPU License Amendment Request (LAR). During the meeting, fourteen (14) RAI questions from the Reactor Systems Branch (SRXB) were presented. On April 11, 2011, FPL received an email from the NRC PM containing the final RAI with ten (10) of the questions previously discussed in the public meeting

[Reference 2]. The ten RAI questions involve specific piping design configurations, reactor fluence calculations, RCS temperature distributions and cooldown times, heat exchanger fouling factors, RHR system capability, natural circulation characteristics, and core outlet temperature monitoring. These ten RAI questions and the applicable FPL responses are documented below.

SRXB-2.1 Provide plant piping configuration drawings that include scale three dimensional (3D) drawings or isometrics with dimensional information and piping and instrumentation diagrams (P&IDs) that describe the feedwater piping associated with the CheckPlus installation. Scaled 3D drawings comparable to Figures 1 and 2 in ER-748 Revision 1 that illustrate the Alden Laboratory test configurations are preferred in place of isometrics that cover the plant piping configuration comparable to the Alden Laboratory figures but the overall response to this RAI is for configuration information that describes piping, valves, flow straighteners (if any), feedwater flow meters, and any other components from the feedwater pumps to at least 10 pipe diameters downstream of each feedwater flow meter.

The following Unit 3 and 4 P&ID and Isometric drawings are provided for the piping associated with the LEFM CheckPlus installations. These drawings are actually portions of engineering modifications which are ongoing, but reflect the expected configuration information requested. This ongoing work, in some cases, results in more than one version of the same drawing due to different modification packages affecting the same piping runs. For Unit 3, dimensions for the LEFM spool pieces do not exactly match the dimensions shown in Cameron's ER-748 Rev. 1 or Alden Laboratory Test Report No. 2009-100-C 1229, Rev. 0. The spool pieces were moved slightly downstream, but within the specified manufacturer's tolerances, due to installation interferences. For Unit 4, the installation matches the dimensions shown in Cameron's ER-752 Rev. 1 and Alden Laboratory Test Report No. 2009-101-C1229, Rev. 1. These vendor diagrams were provided in of the EPU LAR No. 205 submittal package.

Unit 3:

Unit 4:

P&ID 5613-M-3074 Sheets 1, 2, 3 P&ID 5614-M-3074 Sheets 1, 2, 3 Isometric 5613-P-817-S Sheets 1, 2, 3 Isometric 5614-P-770-S Sheets 1, 2, 3, 4 See attached figures for a full set of these Unit 3 and Unit 4 drawings.

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 L-2011-141 Attachment Page 3 of 22 SRXB-2.2 Section 2.8.4.3 discusses reactor vessel neutron fluence calculations performed to support the pressure and temperature limits and cold over-pressure mitigation system setpoint confirmation. Describe how the fluence calculations account for uprated core operation.

SRXB-2.3 In determining the projected fast neutron (E > 1.0 MeV) fluence for the reactor materials, a plant-specific calculation was performed for fuel cycles that have been completed, and fluence projections for future operation were generated based on an assumed mode of operation. Specifically, spatial power distributions and core power level for uprated core designs were used to determine the neutron exposure during future operation. The uprated core designs were based on expected fuel management strategies for the uprate (core power level, core loading, cycle length, core power distributions, loading patterns, etc.).

Assumption 2 in Section 2.8.4.4.2.2 states that the reactor coolant system (RCS) is assumed to be at a uniform temperature during residual heat removal (RHR) system operation and that there are no "hot spots" that could cause an unexpected increase in the bulk RCS temperature. Discuss why the uniform temperature assumption was made. Justify this assumption with respect to the core, upper plenum, upper head, and the pressurizer for both RCPs running and not running. Include a discussion of the implications where RCS temperature is not uniform and equal to average RCS temperature in the hot legs. Include drain down of RCS inventory and emptying the pressurizer in this discussion.

The uniform temperature assumption identified in Licensing Report (LR) Section 2.8.4.4.2.2 is used for the RCS cooldown analyses. The cooldown analyses performed confirm the performance of the RHR system in cooling the plant for various plant shutdown evolutions under EPU conditions. The RCS is modeled as a lumped heat source with a conservative, single-valued heat capacity. This analytical approach has been used and refined over many years and the code results are consistent with actual plant cooldown performance from many reactor-years of PWR plant operation. The consistency of the analyses to actual plant performance has confirmed that individual heat loads associated with specific RCS components (e.g., core, upper plenum, upper head, or pressurizer) do not need to be differentiated in the cooldown model. The lumped RCS heat capacity provides an appropriate representation of the RCS latent heat load on the RHR system during cooldown. Thus, uniform temperature distribution in the RCS is assumed.

The cooldown analysis is intended to accurately characterize the cooldown capabilities of the plant under EPU conditions. It is not intended to address the discrete temperature profiles of the RCS components and piping during RHR system operation. Although the actual fluid temperatures vary throughout the RCS, forced circulation via a Reactor Coolant Pump (RCP) provides for adequate mixing and acts to reduce temperature differentials across the system.

The continued operation of a Reactor Coolant Pump (RCP) during the early stages of the RHR cooldown process improves effective mixing of the reactor coolant and thermal homogeneity within the primary system. With one RCP in operation, one RCS volume is recirculated in less than one minute. Since a typical cooldown rate

Turkey Point Units 3 and 4 L-2011-141 Docket Nos. 50-250 and 50-251 Attachment Page 4 of 22 is approximately 1 0F/minute, a single RCP running assures a relatively constant RCS temperature throughout the system, including the upper plenum, and upper head regions. PTN's procedures require one RCP to remain running down to 180'F to provide added assurance of uniformity of RCS temperature. For cooldown scenarios in which the RCPs are not available (such as Loss of Offsite Power), the operation of the RHR pumps provides for mixing capability. The RHR cooldown analysis commences in Mode 3 (350'F) when the RHR system is aligned to the RCS and continues to Mode 5 (200'F) or Mode 6 (140'F), depending on the case being analyzed.

In these cooldown scenarios, pressurizer level is procedurally maintained and the pressurizer is not drained. This is conservative since it maximizes the water mass and, therefore, the RCS heat capacity. The cooldown analyses also model core decay heat and an appropriate value of RCP heat input, consistent with the RCP normal operating procedures during plant cooldown, in which the last operating RCP is secured below 180'F.

SRXB-2.4 Section 2.8.4.4.2.3 identifies a higher heat load that often extends cooldown time, yet, in the case of normal cooldown, time is 28 hours3.240741e-4 days <br />0.00778 hours <br />4.62963e-5 weeks <br />1.0654e-5 months <br /> versus the lower heat load cooldown time of 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. Explain since the expectation is that an increased heat generation rate would result in a longer cooldown time. The potential concern is that the analysis methodology used for normal cooldown may be applicable to other conditions and an understanding is necessary to assess predicted plant behavior for other conditions.

LR Section 2.8.4.4 states that, in addition to the increased initial core power level, the updated analysis for the EPU also includes other system and component design input changes, all of which affect the cooldown results. These additional changes include seasonally adjusted best-estimate values of the maximum Intake Cooling Water (ICW) system supply temperature and auxiliary plant heat loads that more accurately reflect the actual operation of the plant. The analysis assumes that the maximum ICW temperature (which occurs in the summer) does not occur concurrently with the maximum auxiliary heat load (which is from the spent fuel pool). Note that PTN refueling outages normally occur in the spring or fall. See Table 1 below. The previous cooldown analysis did not consider these refinements in the analysis inputs. Hence, despite the increased core decay heat rate, these additional changes in the cooldown analysis input parameters for the EPU resulted in a net reduction in the reported cooldown time for the worst case normal two-train plant cooldown (EPU Case A in Table 1) when compared to the results of comparable case in previous analysis.

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 L-2011-141 Attachment Page 5 of 22 TABLE 1 Turkey Point EPU RAI SRXB-2.4 Normal Two-Train Plant Cooldown**

Reactor Power Rating Current Extended Power Uprate Cooldown Analysis Power 2644 MWt Input Parameters and Results 2346 MWt Case A Case B Maximum ICW Temp. ('F) 95 92 97 Auxiliary Heat Load 26.0 26.85 21.87 (MBtu/hr)

RHR Cut-in Time (Hrs. after 7.5 7.5 7.0 Shutdown)

Time to Cooldown to 200'F 30.0 28.2 26.8 (Hrs. after Shutdown)

• • 2 RHR pumps, 2 RHR HXs, 2 CCW pumps, 3 CCW HXs, 2 ICW pumps, & 1 RCP Under the most restrictive operating parameters and in the unlikely scenario that a plant cooldown is initiated with less than all of the cooling equipment available, RCS cooldown would continue to be possible, albeit at a much slower rate. The potential for cooldown times in excess of 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> under certain conditions is currently being evaluated under the Turkey Point corrective action program. This performance is consistent with the current licensing basis of the plant.

Sections 2.8.4.4.2.3 and 2.8.4.4.2.5 state that there are no safety-related design criteria for normal plant cooldown times and, therefore, the calculated cooldown times are acceptable. Yet Section 2.8.4.4.2.5 also states a Technical Specification (TS) time limit for achieving cold shutdown of 6 + 30

= 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> after reactor shutdown and states that the 28 hours3.240741e-4 days <br />0.00778 hours <br />4.62963e-5 weeks <br />1.0654e-5 months <br /> applies.

Explain the rationale for concluding that calculated cooldown times are acceptable since there are stated to be no safety-related design criteria in contrast to using these results for meeting TS and other cooldown criteria.

The current licensing basis for Chapter 14 events for Turkey Point defines safe shutdown as hot standby. Thus, cooldown to cold shutdown is not addressed for these events. Cold shutdown is addressed for Appendix R. Accordingly, various cooldown scenarios were addressed in the EPU LAR, two of which are considered quality-related analyses consistent with FPL Quality Assurance Topical Report (QATR) requirements: normal plant TS cooldown to cold shutdown within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> and an Appendix R cooldown to cold shutdown within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

The Appendix R cooldown is discussed in LR Section 2.5.1.4, Fire Protection. As stated in LR Section 2.5.1.4, the cooldown analysis performed for the Appendix R cooldown demonstrates that cooldown to cold shutdown would be accomplished within 63.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />, meeting the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> acceptance criteria. The normal plant TS cooldown, as indicated in FPL responses to SRXB-2.3 and 2.4, assumes availability of two trains, i.e., the normal two-train plant cooldown presented in Table 1 above.

SRXB-2.5

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 L-2011-141 Attachment Page 6 of 22 SRXB-2.6 SRXB-2.7 The designs of the CCW and RHR systems in the Turkey Point plant contain redundant pumps and heat exchangers arranged in parallel cooling trains for system reliability and maintainability. Under the most restrictive operating parameters and in the unlikely scenario that a plant cooldown is initiated with less than all of the cooling equipment available, RCS cooldown would continue to be possible, albeit at a much slower rate. The potential for cooldown times in excess of 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> under certain conditions is currently being evaluated under the Turkey Point corrective action program. This performance is consistent with the current licensing basis of the plant.

There appears to be no information that addresses the effect of the EPU on heat exchanger fouling factors. Address the behavior of heat exchanger fouling factors due to the higher heat load, longer cooldown times, and greater differential temperatures.

Heat exchanger fouling factors typically are a function of water chemistry and cleanliness, which will not change due to the EPU conditions. The fouling factors are not affected by heat load, cooldown times, or differential temperatures, based on the following considerations. The analysis assumes that during normal plant cooldown using the RHR System, the flow rate of reactor coolant through the RHR heat exchanger is throttled as needed to meet two operating limitations: (a) a maximum RCS cooldown rate of 50'F per hour, and (b) a maximum CCW supply temperature of 125'F. This throttling of the RHR flow rate typically is required early in the cooldown transient when favorable heat exchanger conditions exist (i.e., high heat exchanger AT). These same limitations were considered in the previous cooldown analysis for the current power rating of the Turkey Point plant.

As a result of the similar throttling of the RHR flow rate that is assumed to occur early in the cooldown transient for both the current and uprated power ratings, the maximum heat loads and differential temperatures for the RHR and CCW heat exchangers are unaffected by the power uprate and there is no effect on heat exchanger fouling factors.

LR Section 2.8.4.4 states that the CCW and RHR heat exchanger heat transfer rates (U and UA values) assume fouling factors that are consistent with the maintenance program for these heat exchangers. These maintenance programs contain surveillance and cleaning procedures that ensure the continued design basis heat transfer performance for these heat exchangers. In addition, PTN Technical Specification Action 4.7.2 requires that it be verified at least once every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> that two CCW heat exchangers and one CCW pump can remove design basis heat loads. Hence, due to the rigorous heat exchanger maintenance program in effect at PTN, the marginally longer plant cooldown times associated with the EPU conditions are not expected to have a significant effect on the heat exchanger fouling factors.

The end of Section 2.8.4.4.2.3 states "The EPU has no effect on the ability of the RHR system to remove residual heat at reduced reactor coolant system inventory, and therefore the PTN (Turkey Point Nuclear Power Plant) will continue to meet the current licensing basis requirements with respect to NRC Generic Letter 88-17." Justify this conclusion in light of the increased

Turkey Point Units 3 and 4 L-2011-141 Docket Nos. 50-250 and 50-251 Attachment Page 7 of 22 decay heat generation rate that must be removed after shutdown. Include the effect on temperature, RHR flow rate including any limitations on flow rate as a function of RCS water level, and potential hot leg vortexing in your justification.

The core decay heat level associated with the EPU conditions will initially be greater due to the higher core power. The early stages of plant cooldown using the RHR system will be performed with the normal RCS water inventory (i.e.,

main coolant loops filled). However, within a period of less than two days, when the RCS water level could be reduced for plant inspection and maintenance, the decay heat generation rate will be less than one percent of the full power heat load and, therefore, not significantly different than that associated with the current licensed power level of the Turkey Point plant at two days after shutdown. Thus, the RHR flow rate required to remove decay heat at that time will also not differ significantly and will remain within current PTN procedure and Technical Specification limits.

GL 88-17 required utilities operating PWRs to implement improved analyses, procedures and equipment designs that are intended to avoid perturbations on the RCS and supporting systems while in a reduced inventory situation. PTN responded to GL 88-17 with programmatic features that ensure that the RCS will be maintained in a stable and controlled condition during reduced inventory operation. Specifically regarding reduced inventory operation, the design flow rate and operating temperature of the RHR system that will exist during plant cooldown and the level of coolant that is maintained in the main RCS loop piping during reduced inventory conditions in response to GL 88-17 are unchanged under EPU conditions. Other system parameters that are relevant to vortex formation in the RHR suction line (i.e., pipe diameter, water level and fluid flow rate and velocity) also are unchanged as a result of the EPU.

There has been no change to the licensing basis of the plant for reduced inventory operation as a result of the EPU. PTN Technical Specification 3/4.9.8.2 remains applicable to refueling operations with low water level under EPU conditions.

Therefore, it has been concluded that the EPU has no effect on the ability of the RHR system to remove residual heat at reduced reactor coolant system inventory, and PTN will continue to meet the current licensing basis requirements with respect to NRC Generic Letter 88-17.

SRXB-2.8 Section 2.8.7.2 states that a minimum subcooling margin of 50'F is maintained during natural circulation cooldown until RCS temperature is below 350'F. What temperature sensors are used to determine RCS temperature and where are they located within the RCS? If hot leg temperatures are used, address the distribution of temperature that is expected in the hot legs. What is used to determine maximum upper head temperature during natural circulation cooldown? Compare the upper head temperatures predicted to exist during natural circulation cooldown for the existing power level and the proposed power level. Include saturation temperature at the uppermost upper head elevation in the comparison.

Turkey Point Units 3 and 4 L-2011-141 Docket Nos. 50-250 and 50-251 Attachment Page 8 of 22 As directed by Turkey Point's Emergency Operations Procedures (EOPs),

subcooled margin during a natural circulation cooldown is monitored using the Core Exit Thermocouples (CETs). The CETs are terminated in the upper support column where they provide the most direct indication of upper head fluid temperature.

For purposes of a natural circulation cooldown the primary concern with respect to upper head temperature is to ensure voiding does not occur in the upper head region due to temperature differences that could potentially develop if the upper head is cooled at a significantly slower rate than the RCS. After the St. Lucie cooldown event that resulted in the formation of a steam bubble in the upper head region and in support of responses to NRC Generic Letter (GL) 81-21, the Westinghouse Owners Group (WOG) performed analyses of the natural circulation cooldown capability for Westinghouse 2, 3, and 4 loop plants to prevent void formation in the upper head. The results of these analyses classified plants into two categories with cooldown rate limits to prevent head voiding based on classification type. These cooldown rates were incorporated into the WOG Emergency Response Guidelines (ERGs) which provides the guidance for creating plant specific EOPs.

To ensure the temperature of the upper head does not significantly deviate from the RCS, the RCS cooldown rate is limited to 25°F/hr by the EOPs for natural circulation cooldown. This 25°F/hr cooldown rate is the rate recommended for a Thot plant as classified by the ERGs to prevent head voiding caused by significant deviations between the cooldown rates of the upper head and RCS. For both the current power and proposed power levels, the RCS cooldown rate remains limited to 25°F/hr.

Additionally, during a natural circulation cooldown one of the primary means of heat removal from the upper head and associated fluid is through the use of Control Rod Drive Mechanism (CRDM) cooling fans. For the proposed increased power level, the heat removal ability of the CRDM fans is unaffected and can adequately cool the upper head as shown in LR Section 2.8.4.1.2. Therefore, the upper head cooldown rate at the current power level and the proposed power level is the same. Therefore, both head cooldown rate and RCS cooldown rate remains unchanged due to the proposed increase in power.

SRXB-2.9 Provide representative RCS natural circulation temperatures that have been observed during operation of the Turkey Point plant and compare these to selected natural circulation temperatures provided in Section 2.8.7.2.2.3.

Since the PTN stretch power uprate in 1996, there have been multiple unit trips including a dual unit trip on 2/26/2008 but no cases of a natural recirculation cool down from full power to RHR system entry conditions (350TF). There was only one case of a partial natural circulation cooldown which occurred on 10/31/2005 while Unit 4 was in Mode 3 at 400TF to the RHR system entry conditions. This cooldown event took place over six hours or more and was punctuated by a variable cooldown rates such that actual plant performance data is would not be representative of data provided in LR Section 2.8.7.2.2.3.

Turkey Point Units 3 and 4 L-2011-141 Docket Nos. 50-250 and 50-251 Attachment Page 9 of 22 SRXB-2.10 Is the Table 2.8.7.2-1 620.8°F core outlet temperature an average value or the peak value located immediately above the hottest region of the core?

The core outlet temperature of 620.8°F is the average temperature of the water directly exiting the core. After this water gets mixed with the flow which bypasses the core, the temperature of the water which enters the hot leg is reduced. The maximum average hot leg temperature established for EPU conditions is 616.8°F. This is provided in LR Table 1.1-2.

References

1. M. Kiley (FPL) to U.S. Nuclear Regulatory Commission (L-2010-113), "License Amendment Request No. 205: Extended Power Uprate (EPU)," (TAC Nos. ME4907 and ME4908), Accession No. ML103560169, October 21, 2010.

2. Email from J. Paige (NRC) to T. Abbatiello (FPL), "Turkey Point EPU - Reactor Systems (SRXB) Request for Additional Information - Round 2," Accession No. ML111010080, April 11, 2011.

Drawings Unit 3 P&IDs (PC/M #s) 5613-M-3074 SH 1 5613-M-3074 SH 2 (PC/M 09-015) 5613-M-3074 SH 3 Unit 3 Isometrics (PC/M #s) 5613-P-817-S SH 1 (PC/M 09-015) 5613-P-817-S SH 2 (PC/M 09-015) 5613-P-817-S SH 3 (PC/M 08-078)

Unit 4 P&IDs (PC/M #s) 5614-M-3074 SH 1 5614-M-3074 SH 2 (PC/M 09-019) 5614-M-3074 SH 3 (PC/M 08-093)

Unit 4 Isometrics (PC/M #s) 5614-P-770-S SH 1 (PC/M 08-093) 5614-P-770-S SH 2 (PC/M 08-093) 5614-P-770-S SH 3 (PC/M 08-093) 5614-P-770-S SH 4 (PC/M 09-019)

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'J5 RErERENCE OPERATING DIAGRAM NO.5613-N-3074 SH 1, 2 L 3 23PIPING FROM FEED'ATER PUMPS 3PIA AND 3PIB TO HIGH SA POINT HEATERS 3E6A AND 3E6B AND FROM ABOVE HEATERS A1 TO CONTAINMENT PENETRATIONS P-27A, P-27B, AND P-27C

'Cl-*

AND TO STRESS PROBLEMS AFW-DO3, F.R-2, C-2 AND DCI I+(

1. • I THIS DRAWING MADE FROM 79-14 WALKDOWN INDEX*

09

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PCM 08-078 /

CONSTRUCTION NOTES <CONT'D)

DES PIE 11, INSTALLATION OF PIECES 1111-BE 11 FOLLOW&

IGN T HP.

Aso FE 47 (S/N 29019)

,,:3 PER DESIGN

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ELI..LET V."E ATIC S

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1. ALL ISI INFORPATIO, WELD NUHJERS T LEC"TIGN$,

ZONES.

LINE NHONERS

& P11P RESTRAINT LOCATIONS ARE F*I SI REFrERENCE ONLY.

ZONE NBPER.

LIRE R,*EN.R 3-10 14'-FWA-2302 S-IllO 1R-TVE,-S04

• -lll 14 -FC-2 306 a THIS SHEET SHALL RE CODRIDI NTE1O AND WORKED WITH ALL OTTER GREETS IN DRAWING 9613-P-8I7-S.

Pcm 00-078 /

CONSTRUCTION NOTES A

WR ON THIG PTM PP-OTB SNALL BE COE]ORNATED WISH PEN a BEE Fp&L HIPE WELD STANDARD SDV-0W 7 O

IDR WELDINGA ENE PREPARATIO

]S AND INSTALLATIION INSTRUCTIONS.

IL INSTALL SEW FE-S-49SA FE-3S4-IR AND FE-3-478 ULTRASONIC FLOW H'IJElRS AT LI]CALIONtS SREW',N.

4. TPSTHLL NTW CENTRAL

TUWING, SWAGELDK VALVES AND FITTINGS AT EACI NEV LEFM ULTRASONIC FLOW METER, SEE DETAIL N

5. VERIFY VALVE AND EOUIPMENT TAGS/NAPTPLAPHS FOR VALVES AND FLOV PETERS LISTED AIOVE ARE PRESENT AND INFORMATION COPTAINED WITHIN IL UP TO DATE.

6 AFTER COMPLETION OF ALL LErH METERS, REMOVE TEMPORARY PIPING SUPPIOTS.

I PLACES' TO FAR REMOVAL OF PIPING FOR THE INSTALLATION OF THE LEFM SEE DECHTEL

)EPREITIDN DVG. PTN-DB-078-L-L A THE PIPING INSULATIDN SRALL BE REI'OVED AND RESTORED AS PER INHULATION SPEC N-P68.

9. TUHING SNALL BE SUPPORTED AS PER SPEC C-R1V.

I1. FIELD TO LEAVE CAP IN PLACE UNTIL THE IPPLNEAIArI IF THE TRAsMITrER INSOALLATIAN TN A LATER REVIGIIMI.

1I. INSTALLATION WILL BE PERFOARNEP IN ACCORDANCE VITH VENDOR DOCTUPET V0009A4 "CAMERON FLOW MEASUREMENT OVSTEK" CPMR

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56G0-P-217 REV. 0 5610-P-21T REV. I SEISMICALLY ANALYZED PIPING HYDRO DESIGN I

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