L-2011-028, Response to NRC Request for Additional Information Regarding Extended Power Uprate License Amendment Request No. 205 and Safety Analyses Issues - Round 1

Response to NRC Request for Additional Information Regarding Extended Power Uprate License Amendment Request No. 205 and Safety Analyses Issues - Round 1

ML110770019
Person / Time
Site:	Turkey Point
Issue date:	03/16/2011
From:	Kiley M Florida Power & Light Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
L-2011-028
Download: ML110770019 (26)

Text

0 MAR 16 2011 FPL.

POWERING TODAY. L-2011-028 EMPOWERING TOMORROW.0 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 Safety Analyses Issues - Round 1

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) Requests for Additional Information - Round 1," Accession No. ML110460085, February 15,2011 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).

By email from the U.S. Nuclear Regulatory Commission (NRC) Project Manager (PM) dated February 15, 2011 [Reference 2], additional information regarding the Steam Generator Tube Rupture (SGTR) Margin-To-Overfill (MTO) analysis, Best Estimate Large Break Loss of Coolant Accident (LBLOCA) analysis, and Turkey Point's (PTN)

General Design Criteria (GDC) 30 on Reactor Holddown Capability was requested by the NRC staff in the Reactor Systems Branch (SRXB) to support their review of the EPU License Amendment Request (LAR). The Request for Additional Information (RAI) consisted of seven (7) questions: five (5) questions regarding the SGTR MTO analysis, one (1) question regarding the Best Estimate LBLOCA analysis, and one (1) question regarding PTN GDC 30 requirements. These seven RAI questions and the 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-2009-133 [Reference 1].

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

400(

an FPL Group company

Turkey Point Units 3 and 4 L-2011-028 Docket Nos. 50-250 and 50-251 Page 2 of 2 Should you have any questions regarding this submittal, please contact Mr. Robert J.

Tomonto, Licensing Manager, at (305) 246-7327.

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

Executed on March /6 ,2011.

Very truly yours, Michael Kiley Site Vice President Turkey Point Nuclear Plant Attachments 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 L-2011-028 Docket Nos. 50-250 and 50-251 Attachment 1 Page 1 of 24 Turkey Point Units 3 and 4 RESPONSE TO NRC RAI REGARDING EPU LAR NO. 205 AND SRXB SAFETY ANALYSES ISSUES - ROUND 1 ATTACHMENT 1

Turkey Point Units 3 and 4 L-2011-028 Docket Nos. 50-250 and 50-251 Attachment I Page 2 of 24 Response to Request for Additional Information The following information is provided by Florida Power & Light (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].

In an email dated February 15, 2011 [Reference 2], the NRC staff requested additional information regarding FPL's request to implement the Extended Power Uprate. The RAI consisted of seven (7) questions from the NRC's Reactor Systems Branch (SRXB): five (5) questions regarding the Steam Generator Tube Rupture (SGTR) Margin to Overfill (MTO) analysis, one (1) question regarding the Best Estimate Large Break Loss-of-Coolant Accident (LBLOCA) analysis, and one (1) question regarding PTN GDC-30 requirements on Reactor Holddown Capability. These seven RAI questions are documented below with the applicable FPL responses.

Steam Generator Tube Rupture SRXB-1.1: Provide a thermal hydraulic analysis for Turkey Point at the proposed, uprated conditions, for a limiting margin-to-overfill/overfill scenario. One acceptable methodology would be for the analysis to align as closely as possible to what is approved in WCAP-10698-P-A; however, since the licensee has asserted that a limiting single failure is not in the Turkey Point licensing basis, this exception to the WCAP-10698-P-A methodology would be acceptable.

Consider limiting single failures and discuss what they could be.

FPL has performed analyses of the limiting margin-to-overfill scenario for operation at the proposed Extended Power Uprate (EPU) core power level of 2644 MWt. The analysis aligned closely to WCAP-10698-P-A. Exceptions are discussed in response SRXB 1.2 below. However it is recognized that a single failure assumption is not in the Turkey Point licensing basis, therefore it is an exception from the consideration of limiting single failures discussed in WCAP-10698-P-A methodology.

The analyses were performed using the LOFTTR2 thermal hydraulic model consistent with the methodology in WCAP- 10698-P-A.

In addition to the changes made to incorporate the modeling presented in WCAP-10698-P-A, updated operator action times to remove excess conservatism in the MTO analysis have also been implemented. These operator response times during recovery from a SGTR event were recorded using the plant training simulator with various operating crews. The times Were tabulated and a bounding set of response times were selected for use in the margin to overfill analysis. Table 2 shows the comparison between the operator action times used on Reference 1 and the times used in the revised analysis. These simulator-based action times have been modeled in the LOFTTR2 analysis to predict the dynamic system response to the Turkey Point specific recovery actions.

FPL has a plant simulator and training programs which provide the required assurance that the necessary actions and times can be taken consistent with those assumed for the WCAP- 10698-P-A design basis analysis.

Turkey Point Units 3 and 4 L-2011-028 Docket Nos. 50-250 and 50-251 Attachment 1 Page 3 of 24 The results indicate a margin to overfill greater than 300 ft3 in the ruptured steam generator (SG) for the EPU case. No water is transferred into the steam lines. The sequence of events for the revised analysis is provided in Table 1. Figures 1, 2, and 3 provide the time-dependent primary and secondary pressures, primary-to-secondary break flow, and ruptured steam generator water volume, respectively, for the limiting EPU scenario.

Table 1: Limiting MTO Scenario Sequence of Events Event EPU Time (sec)

Tube Rupture 0 Reactor Trip and LOOP 102 AFW Initiation 103 SI Actuation 113 Ruptured SG AFW Isolation 403 Reduce SI Pumps Running 704 isolate Ruptured SG MSIV 1304 Initiate Cooldown with Intact SG 1784 Establish Charging Flow 1788 Terminate Cooldown 2060 Initiate Depressurization 2420 Terminate Depressurization 2476 Terminate SI Flow .2656 Balance Charging and Letdown Flows 2780 Break Flow Termination 3132

Turkey Point Units 3 and 4 L-2011-028 Docket Nos. 50-250 and 50-251 Attachment 1 Page 4 of 24 Table 2: SGTR Operator Action Times Action EPU Time (as provided Revised EPU Time in Reference 1)

Operator action time to isolate auxiliary 5 minutes 5 minutes feedwater flow to the ruptured steam generator following reactor trip Operator action time to isolate safety 18 minutes 10 minutes injection flow from two of the four safety injection pumps following reactor trip Operator action time to close main steam 27 minutes 20 minutes*

isolation valve to isolate steam flow from the ruptured steam generator following reactor trip Operator action time to initiate cooldown 10 minutes (following 28 minutes (after isolation of the ruptured reactor trip) steam generator)

Operator action time to establish maximum Start of cooldown Start of cooldown charging flow OR OR 37 minutes from reactor 28 minutes from trip** reactor trip**

Plant response to complete cooldown LOFTTR2-calculated LOFTTR2-calculated Operator action time to initiate 5 minutes 6 minutes depressurization following completion of cooldown Plant response to complete depressurization LOFTTR2-calculated LOFTTR2-calculated Operator action time to terminate ECCS flow 3 minutes 3 minutes following completion of depressurization Operator action time to balance letdown and 2 minutes 2 minutes charging flow following safety injection termination Plant response until break flow termination LOFTTR2-calculated LOFTTR2-calculated resulting from primary and secondary pressure equalization

• Required to be closed prior to initiation of cooldown. Not an explicit operator response time.

• • The assumption of a minimum time to perform this step decreases the margin to overfill the SG and results in a conservative assumption in the analysis.

Turkey Point Units 3 and 4 L-2011-028 Docket Nos. 50-250 and 50-251 Attachment 1 Page 5 of 24 Figure 1: RCS and Secondary Pressures (EPU)

. . . .- RCS RInto re SO Ruptured SC 0ýS Mt 0 500 1000 1500 2000 2500 3000 3500 Time (s)

Turkey Point Units 3 and 4 L-2011-028 Docket Nos. 50-250 and 50-251 Attachment 1 Page 6 of 24 Figure 2: Ruptured Steam Generator Break Flow (EPU)

U)

E C3, 0D

_z:

3500

Turkey Point Units 3 and 4 L-2011-028 Docket Nos. 50-250 and 50-251 Attachment 1 Page 7 of 24 Figure 3: Ruptured Steam Generator Water Volume (EPU)

Ava iIob e Ruptured SG JU1/4JU 4000-3000-t

/

E 0

2000" -I 1000" "

I I I II liii II I*I I I I I I I V I1 1 I I snO 10,0 1500 2000 2500 3000 .3500 0

Time (s)

Turkey Point Units 3 and 4 L-2011-028 Docket Nos. 50-250 and 50-251 Attachment 1 Page 8 of 24 SRXB-1.2: For the revised margin to overfill/overfill analysis, provide a table comparing analytic assumptions used in WCAP-10698 to those used in the Turkey Point analyses, and justify any differences.

Table 3 provides a comparison of the analytical assumptions used in WCAP-10698-P-A to those used in the Turkey Point analyses.

Table 3 Comparison of WCAP-10698-P-A Modeling to the Revised Analysis Assumptions Parameter WCAP-10698 Model PTN Revised SGTR MTO Analysis Direction of Conservatism EPU Initial Conditions Power (1) Full Power (Nominal + Full Power (Nominal +

Uncertainty) Uncertainty)

RCS Pressure Minimum Minimum Pressurizer Water Level Maximum Maximum SG Secondary Mass Maximum Maximum Break Location Cold-leg Cold-leg Offsite Power Availability Offsite Power Loss of Offsite Power Loss of Offsite Power (LOOP) (LOOP)

Protection Setpoints and Errors Reactor Trip Delay Minimum Minimum Turbine Trip Delay Minimum Minimum SG Relief or Safety Valve Minimum (PORV)' Minimum (PORV)

Pressure Setpoint Pressurizer Pressure Trip Maximum Maximum Setpoint Pressurizer Pressure SI Maximum Maximum Setpoint Safeguards Capacity SI Flow Rate Maximum Maximum AFW Flow Rate Maximum Maximum (isolation on operator action time)

AFW System Delay Minimum Minimum AFW Temperature Maximum Minimum(21 Control. Systems CVS Operation, PZR Not Operating Not Operating Heater Control Turbine Runback Mass Included Not Included(3)

Penalty RCP Running Not Operating Not Operating Decay Heat Decay Heat Maximum ANS 1979-2a(2)

Single Failure Single Failure Included Not Included, consistent with current licensing basis (CLB)

Turkey Point Units 3 and 4 L-2011-028 Docket Nos. 50-250 and 50-251 Attachment 1 Page 9 of 24 (1) Consistent with the discussion of power in WCAP-10698-P-A, the initial steam generator mass is more conservatively calculated without inclusion of the initial power uncertainty since it results in a higher mass.

(2) For this revised analysis, the 1979 American Nuclear Society (ANS) decay heat model minus 2u uncertainty is used. Plant specific sensitivities performed to address the NSAL-07-1 1 issue regarding the use of a higher decay heat uncertainty confirmed that the use of the 1979-2u decay heat is conservative compared to the 1971+20% ANS decay heat model specified by the methodology of WCAP- 10698-P-A. Additionally plant-specific sensitivities for Turkey Point-concluded that it is more conservative (i.e., less margin to overfill) to model AFW temperature differently than prescribed by WCAP-10698-P-A.

(3) There is no automatic OTAT turbine runback system at Turkey Point, and thus no penalty is included. This is an acceptable deviation from the WCAP-10698-P-A methodology since it incorporates plant-specific configuration.

SRXB-1.3: For the SGTR analyses, provide a list of systems, components, and instruments that are credited for accident mitigation in the plant-specific EOPs. Specify whether each component is safety grade, consistent with Requirement (4) of the NRC staff SER approving WCAP-10698.

Information pertaining to credited systems, components, and instruments is presented in the following table. Equipment specific to Unit 3 is shown, but identical equipment is available for Unit 4. Any single component is shown in the table only once, even though some components are relied upon several times throughout the EOPs. The list presents equipment which is specifically utilized in the EOP for mitigating a SGTR event, to include terminating the release from the ruptured steam generator, stopping primary-to-secondary leakage, and restoring RCS pressure, temperature, and inventory control.

Safety Related Equipment/Tag Description or Quality Related 3P215A/B SI Pumps SR MOV-3-843A/B SI Cold Leg Injection Iso Valves SR FT-3-943 SI Cold Leg Injection Flow Indication SR PT-3-455/456/457 Pressurizer Pressure Indication SR EDG 3K4A / B Emergency Diesel Generators SR P2A / P2B I P2C AFW Pumps A, B, and C SR MOV-3-1403/1404/1405 MS Isol to AFW Pumps SR CV-3-2816/2817/2818 AFW Flow Control Valves. SR(1)

CV-3-2831/2832/2833 FT-3-1401A/B;1457A/B; AFW Flow Indication SR 1458A/B LT-3-474/475/476 LT-3-484/485/486 S/G Narrow Range Level Indication SR LT-3-494/495/496 TR-3-410 RCS Cold Leg Indicator/Recorder SR PCV-3-455C/456 Pressurizer PORVs SR (2

Turkey Point Units 3 and 4 L-2011-028 Docket Nos. 50-250 and 50-251 Attachment 1 Page 10 of 24 PT-3-474/475/476 PT-3-484/485/486 S/G Pressure Indication SR PT-3-494/495/496 RD-3-15 SJAE Radiation Monitor QR(3)

RD-3-19 SGBD Radiation Monitor QR_(3)

RAD-3-6417 SJAE SPING Radiation Monitor QR(3)

FT-3-474/475 FT-3-484/485 S/G Steam Flow Indication SR FT-3-494/495 CV-3-1606/1607/1608 MSL Steam Dumps to Atmosphere SR(4) 3CM / 3CD Instrument Air Compressors NNS CV-3-6275A/B/C SG Blowdown Isolation Valves SR POV-3-2604/2605/2606 MSIVs SR MOV-3-1400/1401/1402 MS bypass valves SR MOV-3-1425/1426/1427 S/G C Sample Line Isolation Valves SR TE-3-#E (Various) Core exit thermocouples SR CV-3-2827/2828 Steam Dump to Condenser Valves QR(5)

CV-3-2829/2930 MOV-3-535/536 Pressurizer PORV block valve SR PT-3-403 RCS Wide Range Pressure Indication SR FT-3-605 Flow Indicator for RHR SR 3P201A/B/C Charging Pumps SR PCV-3-455A/B Pressurizer Spray Valves SR(6)

CV-3-31 1 Auxiliary Spray Valves SR(6)

LT-3-460/461/462 PRZ Level Indicator SR MOV-3-865A/B/C SI Accumulator Isolation MOVs SR Table Notes:

(1) Equipment is safety-related with a dedicated, safety-related nitrogen backup supply.

(2) Equipment is safety-related with a dedicated, quality-related nitrogen backup supply.

(3) Operators are directed to perform steam line surveys and monitor steam generator level indications to identify the affected steam generator, in addition to steam generator sampling. Delays associated with sampling will not delay the performance of mitigating actions, since steam and feedwater flow mismatch, level indications, and radiation surveys will provide clear indication of the affected steam generator.

(4) Equipment is safety related with nitrogen backup. Nitrogen for the control signal is from a dedicated, quality-related bottled source. Nitrogen for motive force on the valve operator is from the plant nitrogen system via quality-related supply piping.

(5) Equipment is quality-related with non-nuclear safety (NNS) instrument air supplied to the operators.

(6) Equipment is safety-related as an RCS pressure boundary, but the operator is supplied by NNS instrument air.

Turkey Point Units 3 and 4 L-2011-028 Docket Nos. 50-250 and 50-251 Attachment 1 Page 11 of 24 SRXB-1.4: Under assumed loss of offsite power (LOOP) conditions, address the functionality of each atmospheric dump valve (ADV). Discuss what, if any, mitigating function the ADV provides and its capability to perform that function under the assumed LOOP.conditions.

An SGTR event is mitigated by isolating the affected steam generator, cooling down the RCS to maintain adequate subcooling, and depressurizing the RCS to eliminate reactor coolant leakage through the tube rupture and maintain RCS inventory. When offsite power is available, the steam dump to condenser valves are used to dump steam from the intact steam generators to the condenser to cooldown the RCS. However, during a LOOP, main feedwater and condensate systems are unavailable; instead, the ADVs on the intact steam generators are used for cooldown, in conjunction with the turbine-driven auxiliary feedwater pumps or the diesel-driven standby steam generator feedwater pump. -

The Turkey Point ADVs are air-operated angle globe valves, configured as air-to-.

open / spring-to-close. One ADV is provided on each steam header upstream of the main steam isolation valves, totaling three ADVs per Unit.

Air for the ADV pneumatic operators is normally supplied by the instrument air system. Each Unit is equipped with one electric motor-driven instrument air compressor, and one diesel-driven air compressor, for a total of four compressors.

The two Units' instrument air systems are normally cross connected, and any one of the four compressors alone can supply the combined instrument air load for both Units operating simultaneously.

A LOOP to either Unit will de-energize that Unit's motor-driven instrument air compressor. However, the associated diesel-driven compressor will automatically start on a loss of power to maintain continuity of instrument air service. In addition, the affected Unit's instrument air dryer is automatically sequenced onto the emergency diesel generators during LOOP conditions. With this arrangement, instrument air is automatically and immediately restored during a LOOP without operator action.

The ADVs are controlled from panel-mounted hand/auto digital controllers in the main control room. A separate controller is provided for each ADV. In the automatic mode, the controller issues a pneumatic valve position signal based on a comparison between the operator-selected setpoint and a digital non-safety related main steam pressure signal. In the manual mode, the operator adjusts the digital controller to directly manipulate the pneumatic valve position signal. To generate the pneumatic valve position signal, the controller receives an air supply that is auctioneered between instrument air (normal) and a dedicated bottled nitrogen source (backup).

Each controller receives electrical power from vital inverter-backed 120 VAC panels. Use of diverse vital AC power panels assures that for any failure event, at least two steam dump controllers will always be available. The use of vital AC power ensures controller availability during LOOP conditions. A closed-position limit switch is installed on each ADV to provide Control Room operators with closed/not-closed position indication via the plant's Digital Control System.

Turkey Point Units 3 and 4 L-2011-028 Docket Nos. 50-250 and 50-251 Attachment 1 Page 12 of 24 With a LOOP, a total loss of instrument air would require the failure of both diesel-driven compressors. In that event, motive force for the valves' operators is backed up through reducers from the plant's nitrogen system (about 80 psig). Separately, the 3 to 15 psig pneumatic control signal to the ADV positioners would be supplied from a dedicated nitrogen bottle station, where one bottle has sufficient capacity to allow continuous operation of one Unit's ADVs for 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />, and subsequently to maintain their position for 8.5 hrs. One bottle is normally valved in, and one additional cylinder is added to act as a common source for both units should extended steady state operation be required. This arrangement of redundant air supplies, along with the inverter-backed controller power supplies, ensures the reliability of each ADV.

SRXB-1.5: Identify any new operator actions credited in the revised margin to overfill/overfill analysis.

There are no new operator actions credited in the above analysis to that provided in Reference 1.

SRXB-1.6: Section 2.8.5.6.3 describes a more refined downcomer model. Provide the following specific information concerning the downcomer model:

a. Provide a detailed description and diagram of the downcomer nodalization, including both fluid and heat structures.

b. Identify the sources of heat modeled in the downcomer.

Parts a and b of this RAI are addressed together since they are related questions.

The noding diagram for Turkey Point Units 3 and 4 with nine downcomer channel stacks is presented as Figure 4. The numbers enclosed in squares represent channel numbers. Channels are used to make vertical connections in the vessel model. The numbers enclosed in circles represent gap numbers.

Gaps are used to make lateral connections in the vessel model. The gap numbers which have a horizontal arrow through them connect the channels shown at the start and end of the horizontal arrow. The gaps which have a diagonal arrow through them have a corresponding numbered gap shown elsewhere on the noding diagram. These gaps connect the channels with the matching gap numbers shown.

The downcomer channels are modeled with the long channel stacks shown on the outer portion of the noding diagram (Figure 4).

Cross-sections of the vessel noding at each section elevation are presented as Figures 5, 6, and 7. The cold legs are connected to channels 30, 31, and 32 and the hot legs are connected to channels 37, 38, and 39 in Section 6 as shown in Figure 6.

The metal structures connected to the downcomer which serve as a heat source during a LBLOCA are shown in Figure 8. Only the downcomer channels are shown in this figure, and the gaps are omitted for clarity. The numbers in squares are again the channel numbers, and the unheated conductors are designated with diamonds. The structure to channel connections and a description of the structures are contained in Table 4.

Turkey Point Units 3 and 4 L-2011-028 Docket Nos. 50-250 and 50-251 Attachment 1 Page 13 of 24 473,53 , m

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Turkey Point Units 3 and 4 L-2011-028 Docket Nos. 50-250 and 50-251 Attachment 1 Page 14 of 24 D Channel SECTION 1: LOWER HEAD 0: Gap SECTION 2: LOWER PLENUM SECTION 3: CORE Figure 5: Turkey Point Units 3 and 4 Cross-Section Diagram for Vessel Sections 1, 2, and 3

Turkey Point Units 3 and 4 L-2011-028 Docket Nos. 50-250 and 50-251 Attachment 1 Page 15 of 24 SECTION 4: CCFL REGION. Channel Q Gap SECTION 5: UPPER PLENUM SECTION 6: NOZZLE REGION BELOW NOZZLES Figure 6: Turkey Point Units 3 and 4 Cross-Section Diagram for Vessel Sections 4, 5, and 6

Turkey Point Units 3 and 4 L-2011-028 Docket Nos. 50-250 and 50-251 Attachment 1 Page 16 of 24 SECTION 7: UPPER PLENUM

[c] Channel ABOVE NOZZLES Q Gap SECTION 8: UPPER HEAD UP TO SECTION 9: -UPPER HEAD ABOVE TOP OFGUIDE TUBES GUIDE TUBES.

Figure 7: Turkey Point Units 3 and 4 Cross-Section Diagram for Vessel Sections 7, 8, and 9

Turkey Point Units 3 and 4 L-2011-028 Docket Nos. 50-250 and 50-251 Attachment 1 Page 17 of 24 171 166 167 1 68 176] 1 I672 1 5I I iii 15 2

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Turkey Point Units 3 and 4 L-2011-028 Docket Nos. 50-250 and 50-251 Attachment 1 Page 18of 24 Table 4 (Page 1 of 4): Metal Structures connected to the Downcomer in the Turkey Point Units 3 and 4 Vessel Model Unheated Downcomer Conductor Channel Description Number Connected 3 2 4 3 One-ninth of vessel wall in Vessel Section 2 5 4 6 2 7 332 One-ninth of thermal shield and radial keys in 8 4_ 4 Vessel

_ Section

_ _2 _ _ _ _ __

11 6 12 7 One-ninth of vessel wall in Vessel Section 3 13 8 14 6 15 7 One-ninth of thermal shield in Vessel Section 3 16 8 17 17 66 One-ninth of outer half of core barrel in Vessel 19 7 Section 3 19 *8 _____________________

22 14 23 15 One-ninth of vessel wall in Vessel Section 4 24 16 25 14 26 15 One-ninth of thermal shield in Vessel Section 4 27 16 36 22 37 23 One-ninth of vessel wall in Vessel Section 5 38 24 39 40 22 23 One-ninth of outer half of core barrel in Vessel 40 23 Section 5 41 24 ____________________

42 22 43 23 One-ninth of thermal shield in Vessel Section 5 44 24 51 2 One-ninth of outer half of core barrel in Vessel 52 3 Section 2 53 30 54 31 One-ninth of vessel wall in Vessel Section 6 55 32 ,

56 56 30 30

• One-ninth of outer half of core barrel in. Vessel 57 31 58 32 Section 6 66 40 67 41 One-ninth of vessel wall in Vessel Section 7 68 42

Turkey Point Units 3 and 4 *L-2011-028 Docket Nos. 50-250 and 50-251 Attachment 1 Page 19 of 24 Table 4 (Page 2 of 4): Metal Structures connected to the Downcomer in the Turkey Point Units 3 and 4 Vessel Model 69 40 One-ninth of outer half of core barrel in Vessel 70 71 41 42 Section 7 71 42 85 4 One-ninth of outer half of core barrel in Vessel Section 2 87 14 One-ninth of outer half of core barrel in Vessel 88 89 15 16 Section 4 89 16 90 52 91 53 92 54 _ One-ninth of vessel wall in Vessel Section 2 93 55 94 56 95 57 96 52 97 53 98 54 One-ninth of outer half of core barrel in Vessel 99 55 Section 2 100 56 101 57 103 52 104 53 105 54 One-ninth of thermal shield and radial keys in Vessel 106 55 Section 2 107 56 108 57 109 58 110 59 111 60 One-ninth of vessel wall in Vessel Section 3 112 61 113 62 114 63 115 58 116 59 117 60 One-ninth of outer half of core barrel in Vessel 118 61 Section 3 119 62 120 63 122 58 123 59 124 60 One-ninth of thermal shield in Vessel Section'3 125 61 126 62 127 63

Turkey Point Units 3 and 4 L-2011-028 Docket Nos. 50-250 and 50-251 Attachment 1 Page 20 of 24 Table 4 (Page 3 of 4): Metal Structures connected to the Downcomer in the Turkey Point Units 3 and 4 Vessel Model 128 64 129 65 130 66 One-ninth of vessel wall in Vessel Section 4 131 67 132 68 133 69 134 64 135 65 136 66 One-ninth of outer half of core barrel in Vessel 137 67 Section 4 138 68 139 69 141 64 142 65 143 66 One-ninth of thermal shield in Vessel Section 4 144 67 145 68 146 69 147 70 148 71 149 72 150 73 One-ninth of vessel wall in Vessel Section 5 151 74 152 75 153 70 154 71 155 72 One-ninth of outer half of core barrel in Vessel 156 73 Section 5 157 74 158 75 159 30 One-ninth of the core barrel ring in Vessel Section 6 160 70.

161 71 162 72 One-ninth of thermal shield in Vessel Section 5 163 73 164 74 165 75 166 76 167 77 168 78 One-ninth of vessel wall in Vessel Section 6 169 79 170 80 171 81

Turkey Point Units 3 and 4 L-2011-028 Docket Nos. 50-250 and 50-251 Attachment 1 Page 21 of 24 Table 4 (Page 4 of 4): Metal Structures connected to the Downcomer in the Turkey Point Units 3 and 4 Vessel Model 172 76 173 77 174 78 One-ninth of outer half of core barrel in Vessel 175 79 Section 6 176 80 177 81 178 31 179 32 180 76 181 77 One-ninth of the core barrel ring in Vessel Section 6 182 78 183 79 184 80 185 81 186 82 187 83 188 84 One-ninth of vessel wall in Vessel Section 7 189 85 190 86 191 87 192 82 193 83 194 84 One-ninth of outer half of core barrel in Vessel 195 85 Section 7 196 86 197 87

c. Discuss how subcooled boiling in the downcomer is modeled.

The treatment of subcooled boiling in the downcomer for the nine downcomer channel stack model is the same as for the three downcomer channel stack model.

Subcooled boiling in the downcomer is calculated using the Chen (Reference 4) correlation. While the Chen correlation was developed for saturated boiling, it can be extended into the subcooled region. The Chen correlation superimposes a forced convective and a nucleate boiling component. Moles and Shaw (Reference

6) compared the Chen correlation to boiling data for several fluids and reported satisfactory agreement for low to moderate subcooling.

During subcooled boiling vapor generation occurs and a significant void fraction may exist despite the presence of subcooled water. In this regime, three processes are of interest relative to the downcomer region:

1. forced convection to the liquid,

2. vapor generation at the wall, and

3. condensation near the wall.

Forced convection to the liquid is treated by the forced convective component of the Chen correlation to determine the heat input into the liquid. The nucleate

Turkey Point Units 3 and 4 L-2011-028 Docket Nos. 50-250 and 50-251 Attachment 1 Page 22 of 24 boiling component of the Chen correlation defines the amount of heat available to cause vapor generation at the wall. The near-wall condensation is estimated using the Hancox-Nicoll (Reference 5) correlation for heat flux at the point where all the bubbles generated collapse in the near-wall region.

Please refer to Section 6-2-3 of the CQD (Bajorek et al., Reference 3) for additional information regarding the treatment of subcooled boiling.

SRXB-1.7: By letter dated October 21, 2010, the license amendment request (LAR) states, "As noted in PTN Updated Final Safety Analysis Report (UFSAR), Section 1.3, the General Design Criteria (GDC) used during licensing of the Turkey Point Nuclear Plant predate those provided today in 10 CFR 50, Appendix A. The PTN GDCs were developed based on the 1967 Atomic Energy Commission Proposed General Design Criteria and are addressed in various sections of the UFSAR."

The LAR also identifies, as one of the GDCs in the Turkey Point licensing basis, PTN GDC-30,Reactivity Holddown Capability: "The reactivity control systems providedshall be capable of making the core subcriticalunder credible accidentconditions with appropriatemarginsfor contingencies and limiting any subsequent return to power such that there will be no undue risk to the health andsafety of the public." The LAR states that PTN GDC-30 is comparable to the current GDC-27.

However, the 1967 proposed GDC (32 FR 10213) that corresponds to PTN GDC-30 is "Criterion30--Reactivity Holddown Capability(Category B). At least one of the reactivity control systems provided shall be capable of making and holding the core subcritical under any conditions with appropriate margins for contingencies."

Apparently, the 1967 proposed GDC-30 is more restrictive than PTN GDC-30.

Explain and justify the difference between PTN GDC-30 and its basis, and the 1967 proposed GDC-30.

Explain how PTN GDC-30 is considered to be equivalent to the current GDC-27, not the current GDC-26.

PTN's licensing basis was and is based on the proposed AEC GDCs as amended by the Atomic Industrial Forum (AIF).

1967 AEC GDC 30, Reactivity Holddown Capability, states: "At least one of the reactivity control systems provided shall be capable of making and holding the core subcritical under any conditions with appropriate margins for contingencies."

1967 AIF Reworded AEC GDC 30, Reactivity Holddown Capability,states: "The reactivity control systems provided shall be capable of making the core subcritical under credible accident conditions with appropriate margins for contingencies and limiting any subsequent return to power such that there will be no undue risk to the health and safety of the public."

10 CFR 50, Appendix A, Criterion 27, Combined reactivity control systems capability,states: "The reactivity control systems shall be designed to have a combined capability, in conjunction with poison addition by the emergency core

Turkey Point Units 3 and 4 L-2011-028 Docket Nos. 50-250 and 50-251 Attachment 1 Page 23 of 24 cooling system, of reliably controlling reactivity changes to assure that under postulated accident conditions and with appropriate margin for stuck rods the capability to cool the core is maintained."

The original proposed AEC GDC 30 requirements are more restrictive than those in either the version adopted by Turkey Point or the current 10 CFR 50 Appendix A.

The first would require a single reactivity system alone be capable of holding the core subcritical prohibiting return to power under all conditions. The PTN version, described in Section 3.1.2 of the UFSAR, requires that the reactivity systems together be capable of initially making the core subcritical for all credible accident conditions and limit any subsequent return to power.

10 CFR 50, Appendix A, Criterion 26, Reactivity control system redundancy and capability,states: "Two independent reactivity control systems of different design principles shall be provided. One of the systems shall use control rods, preferably including a positive means for inserting the rods, and shall be capable of reliably controlling reactivity changes to assure that under conditions of normal operation, including anticipated operational occurrences, and with appropriate margin for malfunctions such as stuck rods, specified acceptable fuel design limits are not exceeded. The second reactivity control system shall be capable of reliably controlling the rate of reactivity changes resulting from planned, normal power changes (including xenon burnout) to assure acceptable fuel design limits are not exceeded. One of the systems shall be capable of holding the reactor core subcritical under cold conditions."

Both of the above GDCs address the capability of the reactivity control systems.

GDC 26 addresses the requirements under normal operation and anticipated operational occurrences. It also addresses reactor holddown capability under cold conditions. GDC 27 addresses the requirements under postulated accident conditions, consistent with PTN-GDC 30.

GDC-27 and PTN-GDC-30 provide for appropriate margins in reactivity capability

("with appropriate margins for contingencies" vs. "with appropriate margin for

• stuck rods", respectively). Both provide for multiple reactivity control systems to satisfy the requirements of the GDC ("The reactivity control systems" for both GDCs). Both have similar success criteria ("limiting any subsequent return to power such that there will be no undue risk to the health and safety of the public" vs. "to assure... the capability to cool the core is maintained", respectively).

Turkey Point Units 3 and 4 L-2011-028 Docket Nos. 50-250 and 50-251 Attachment 1 Page 24 of 24 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) Requests for Additional Information - Round 1", Accession No. ML110460085, February 15, 2011.

3. Bajorek, S. M., et al., March 1998, "Code QualificationDocument for Best Estimate LOCA Analysis," Volume 1 Revision 2, and Volumes 2 through 5, Revision 1., WCAP-12945-P-A (Proprietary).

4. Chen, J. C., 1963, "A Correlation for Boiling Heat Transfer to Saturated Fluids in Convective Flow," ASME 63-HT-34.

5. Hancox, W. T. and Nicoll, W. B., 1971, "A General Technique for the Prediction of Void Distributions in Nonsteady Two-Phase Forced Convection," Int. J. Heat and Mass Transfer, Vol. 14.

6. Moles, F. D., and Shaw, J. F. G., 1972, "Boiling Heat Transfer to Subcooled Liquids Under Conditions of Forced Convection," Trans. Inst. Chem. Eng., Vol. 50.