Crystal River, Unit 3 - Attachment E to 3F0712-03 ... · location is a 1.418 ft2 break in one of the side branches of steam generator B at the junction connecting the main feedwater

FLORIDA POWER CORPORATION

CRYSTAL RIVER UNIT 3

DOCKET NUMBER 50-302 / LICENSE NUMBER DPR-72

ATTACHMENT E

TECHNICAL REPORT ANP-3052:CR-3 EPU FEEDWATER LINE BREAK ANALYSIS WITH

FAILURE OF FIRST SAFETY GRADE TRIP,REVISION 2

J! 11 1

ANP-3052Revision 2

CR-3 EPU Feedwater Line Break Analysiswith Failure of First Safety Grade Trip

June 2012

AAREVAAREVA NP Inc.

Controlled Document

ANP-3052Revision 2

June 2012

CR-3 EPU Feedwater Line Break Analysis with Failure of First Safety Grade Trip

Copyright © 2012

AREVA Inc.All Rights Reserved

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Record of Revision

Revision PageslSections/ParagraphsNo. Changed Brief Description I Change Authorization

000 All Initial Release

001 All The FWLB analysis with the first safety grade trip failedwas re-evaluated using the model and limiting conditionsfrom Reference [3]. Revision 001 is a complete revision.

002 Page 10 Added discussion on the rod worth and break location

Page 11 Added Figure 3-1

Table 3-1 Added critical flow models used

i +

4 +

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CR-3 EPU Feedwater Line Break Analysis Revision 2with Failure of First Safety Grade Trip Page 4

Table of Contents

1.0 Introduction ....................................................................................................................................... 82.0 Analytical M ethodology ............................................................................................................ 93.0 Analysis Inputs ................................................................................................................................ 104.0 Results / Conclusions ..................................................................................................................... 145.0 References ..................................................................................................................................... 31

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ANP-3052CR-3 EPU Feedwater Line Break Analysis Revision 2with Failure of First Safety Grade Trip Page 5

List of Tables

Table 3-1: Input to Feedwater Line Break Analysis with First Trip Failed ........................................... 12Table 4-1: Sequence of Events for FWLB with First Trip Failed ......................................................... 15Table 4-2: Results for FW LB with First Trip Failed ............................................................................... 15

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List of Figures

Figure 3-1: Feedwater Line Break Locations ........................................................................................ 11Figure 4-1: FWLB with Fail First Trip - RCS Pressure ........................................................................ 16Figure 4-2: FWLB with Fail First Trip - Reactor Power ........................................................................ 17Figure 4-3: FWLB with Fail First Trip - Reactivity ................................................................................. 18Figure 4-4: FWLB with Fail First Trip - Primary System Temperatures .............................................. 19Figure 4-5: FWLB with Fail First Trip - Indicated Pressurizer Level ..................................................... 20Figure 4-6: FWLB with Fail First Trip - Pressurizer Collapsed Liquid Level ......................................... 21Figure 4-7: FWLB with Fail First Trip - Pressurizer Surge Line Flow ................................................... 22Figure 4-8: FWLB with Fail First Trip - RCS Volumetric Flow Rate ...................................................... 23Figure 4-9: FWLB with Fail First Trip - Pressurizer Safety Valve Flow ................................................ 24Figure 4-10: FWLB with Fail First Trip - SG Secondary Side Liquid Level .......................................... 25Figure 4-11: FWLB with Fail First Trip - SG Secondary Side Inventory .............................................. 26Figure 4-12: FWLB with Fail First Trip - SG % Operating Range ........................................................ 27Figure 4-13: FWLB with Fail First Trip - SG Pressure ........................................................................ 28Figure 4-14: FWL1B with Fail First Trip - EFW Flow ............................................................................. 29Figure 4-15: FWLB with Fail First Trip - Integrated MSSV Flow ......................................................... 30

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Nomenclature

Acronym Definition

BOC Beginning of Cycle

CR-3 Crystal River Unit 3

DSS Diverse Scram System

EFIC Emergency Feedwater Initiation and Control

EFW Emergency Feedwater

EPU Extended Power Uprate

FWLB Feedwater Line Break

LAR Licensing Amendment Request

LOOP Loss of Offsite Power

MFW Main Feedwater

MSSV Main Steam Safety Valves

NRC Nuclear Regulatory Commission

PORV Pilot Operated Relief Valve

PSV Pressurizer Safety Valve

RCP Reactor Coolant Pump

RCPB Reactor Coolant Pressure Boundary

RCS Reactor Coolant System

RPS Reactor Protection System

SG Steam Generator

Tave Average RCS Temperature

TSV Turbine Stop Valves

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1.0 INTRODUCTION

The Nuclear Regulatory Commission (NRC) Reactor Systems Branch staff requested an additional analysis beperformed to support the review of the Crystal River Unit 3 (CR-3) extended power uprate (EPU) licensingamendment request (LAR). In particular, the staff requested an analysis of the feedwater line break (FWLB)transient assuming that the first safety grade reactor protection system (RPS) trip function fails to trip the reactor.This report documents the results of the requested analysis.

The analysis documented in this report assumes that the RPS high reactor coolant system (RCS) pressure tripfunction fails to trip the reactor. This analysis models the non-safety grade diverse scram system (DSS) trip,which inserts the regulating control rod banks upon reaching the DSS high RCS pressure trip setpoint. The peakRCS pressure during the FWLB transient is reported and compared to an acceptance criterion of 120% of thereactor coolant pressure boundary (RCPB) design pressure (1.20 * 2500 = 3000 psig).

The evaluations of the FWLB transient with the first safety grade trip failed are based on the model andconclusions in Reference (3]. The model refinements described in Section 4.1 of Reference [3] and the limitinginitial conditions described in Section 4.12 of Reference [3] are included in the model for evaluating the first safetygrade trip being failed.

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2.0 ANALYTICAL METHODOLOGY

The thermal-hydraulic analysis of the FWLB transient at the CR-3 EPU power level with the first safety grade tripfailed is performed using the RELAP5/MOD2-B&W computer program (Reference [1]). The code simulates RCSand secondary system operation. The reactor core model is based on a point kinetics solution with reactivityfeedback for control rod assembly insertion, fuel temperature changes, moderator temperature changes, andchanges in boron concentration. The RCS model provides for heat transfer from the core, transport of the coolantto the steam generators (SG), and heat transfer to the steam generators. The secondary model includes adetailed depiction of the main steam system, including steam relief to the atmosphere through the main steamsafety valves (MSSVs) and simulation of the turbine stop valves (TSVs). The secondary model also includes thedelivery of feedwater, both main and emergency, to the steam generators.

The RELAP5/MOD2-B&W code has been approved by the NRC for use in non-LOCA safety analyses (Reference

[2]). The analysis documented in this report is consistent with Reference [2] with two exceptions:

1) The first safety grade trip is not credited. Instead, this analysis credits the next available trip, which is thenon-safety grade DSS trip on high RCS pressure.

2) Reference [2] was not used to define the initial conditions for the transient. The initial conditions arebased on the FWLB sensitivity studies documented in Reference (3]. The conditions in Reference [3] thatresult in the highest peak RCS pressure during a FWLB transient are:

a. Nominal RCS pressure (2170 psia hot leg pressure)b. Maximum RCS flow (398,850 gpm)c. 100.4 %FPd. Beginning of Cycle (BOC)e. 585 °F Average RCS Temperature (Tave)f. 940 psia turbine header pressureg. 80 %OR SG levelh. 290 inch indicated pressurizer level

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3.0 ANALYSIS INPUTS

The FWLB analysis with the first safety grade trip failed is based the FWLB sensitivity analyses in Reference [3].Section 3.0 of Reference [3] describes how the sensitivity analyses compare to the FWLB analyses withoutpressurizer spray performed to support Section 2.8.5.2.4 of the CR-3 EPU LAR.

The key input changed to model the first safety grade trip being failed includes:

1. The RPS high RCS pressure trip function is disabled.

2. The DSS high RCS pressure trip function is modeled. The DSS trip setpoint is modeled as 2465 psia.The DSS high RCS pressure trip setpoint includes margin to bound possible changes in the containmentpressure during a FWLB. The DSS trip delay time is modeled as 1.23 seconds. Finally, the rod worthavailable to the DSS system is modeled as only 2.0 %AkIk, since the DSS system only inserts theregulating control rod groups. The rod worth is based on the regulating control rod groups having a worthof 2.2 %Ak/k at the rod insertion limits. The worth is then reduced by 10% (0.2 %Ak/k) for uncertainty. Anexplicit allowance for the worth of a stuck regulating control rod is not considered.

3. Section 4.10 of Reference [3] contains a sensitivity study on the break location. The failed first tripanalysis was performed at the three limiting break locations from that study, to ensure that the limitingbreak location is captured. The break locations considered are shown in Figure 3-1. The peak RCSpressures for the three break locations ranged from 2911.61 psia to 2915.39 psia. The limiting breaklocation is a 1.418 ft2 break in one of the side branches of steam generator B at the junction connectingthe main feedwater line to the side branches.

Table 3-1 summarizes all of the key input to the FWLB analysis with the first safety grade trip failed, including thechanges highlighted above.

The peak RCS pressure is reached shortly after reactor trip before any active safety systems such as emergencyfeedwater (EFW) are credited to mitigate the results. Therefore, there are no single failure assumptions thatwould result in a more limiting peak RCS pressure. However, the FWLB overpressure protection analysis with thefirst safety grade trip failed assumes the same single failure assumption as the FWLB analysis performed tosupport Section 2.8.5.2.4 of the CR-3 EPU LAR. The single failure assumption modeled is the failure of one trainof Emergency Feedwater Initiation and Control (EFIC) such that EFW flow is not initiated automatically in onetrain. Consequently, only one of the two EFW pumps is assumed available to provide flow to the SGs. Thissingle failure assumption produces a conservative long-term transient response.

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Figure 3-1: Feedwater Line Break Locations

Isolation Check Valve.

18"

14" 14"

SG

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Table 3-1: Input to Feedwater Line Break Analysis with First Trip Failed

Parameter Value

RCS Conditions

Core Power, MWt 3014 * 1.004 = 3026.1

Decay Heat 1.0*ANS71 plus B&W Actinides

Total Net Reactor Coolant Pump (RCP) Heat, MWt 16.4

Average RCS Temperature, OF 585

Initial Hot Leg Pressure, psia 2170

Total RCS Flow Rate, gpm 398,850

Pressurizer

Initial Indicated Pressurizer Level, in 290

Pressurizer Spray Not Modeled

Pressurizer Heaters Not Modeled

Pilot Operated Relief Valve (PORV) Not Modeled

Pressurizer Safety Valve (PSV) Setpoints, psig 2500 * (1 + 0.03) (open)2500 * (1 - 0.04) (close)

Total PSV Rated Capacity, Ibm/hr 2 * 317,973 @ 2750 psig

Secondary Side

Initial Main Feedwater (MFW) Temperature, OF 460 OF

Tube Plugging, % 5

Initial SG Level, %OR 80

EFW Temperature, OF 120

EFW Minimum Required Flow, gpm 660

EFW Delay Time, sec 40

Turbine Trip Delay Time, s 0.0

TSV Stroke Time, s 0.2

Number of Main Steam Safety Valves (MSSVs) per SG 8

MSSV Capacity per SG 7 @ 845,759 Ibm/hr1 @ 583,574 Ibm/hr

MSSV Nominal Setpoints 2 @ 1050 psig2 @ 1070 psig2 @ 1090 psig

2 @ 1100 psig including small MSSV

MSSV Setpoint Tolerance +3%

MSSV Accumulation +3%

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Parameter Value

MSSV Blowdown -5%

Core Kinetics Parameters

Doppler Temperature Coefficient (Ak/°kIF) -1.30 E-5

Moderator Temperature Coefficient (Ak/k/IF) 0.0 E-4

Prompt Neutron Generation Time, (ps) 24.8

Effective Delayed Neutron Fraction 0.0070

DSS Insertable Rod Worth, %Ak/k 2.000

RPS High RCS Pressure Trip Assumed Failed

DSS High RCS Pressure Trip Setpoint, psia 2465

DSS Trip Delay Time, s 1.23

Critical Flow Models

Subcooled Liquid Homogeneous Equilibrium Model

Two-Phase and Superheated Fluid Moody

Miscellaneous

Offsite Power Available

Single Failures One Train of EFIC

Operator Actions None

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4.0 RESULTS I CONCLUSIONS

The sequence of events for the FWLB accident with the first safety grade trip failed is listed in Table 4-1 and thecalculated results are tabulated in Table 4-2. Plots that demonstrate the transient response following a FWLB areprovided in Figures 4-1 through 4-15.

Following initiation of the FWLB, the blowdown of the affected SG results in a reduction in the secondary heatremoval. The mismatch between energy addition to the reactor coolant and the secondary heat removal causesthe reactor coolant to heat up and pressurize. The pressure increases to the RPS high RCS pressure tripsetpoint, but the trip is assumed to fail. The pressure continues to increase to the DSS high RCS pressure tripsetpoint. After the appropriate delay time, the DSS trip inserts the regulating control rod banks.

After reactor trip, the RCS pressure continues to increase until the PSVs lift. Shortly after the PSVs lift, thepressure begins to decrease. The peak RCS pressure occurs in the bottom of the reactor vessel and does notexceed 120% of the design pressure of 2500 psig (3000 psig).

EFIC actuates on low SG level in the affected SG. After considering the possible 40 second delay, EFW isprovided to the unaffected SG at the flow rate of one EFW pump (660 gpm). During this time, the PSVs maintainthe RCS pressure based on the PSV open and close setpoints. The FWLB event is sufficiently severe that thepressurizer fills. As a result, the PSVs begin to pass single-phase liquid. The PSVs of the type installed at CR-3achieve satisfactory performance for fluid temperatures greater than -550 0F. An additional check is performed toshow that the PSV fluid inlet temperature remains greater than 600 OF to ensure that the PSVs operate asintended. Figure 4-4 demonstrates that at all times throughout the FWLB transient, the liquid temperature at thetop of the pressurizer remains above 600 OF. The PSVs close for the final time at -170 seconds. At -8 minutes,the secondary heat removal from EFW causes the pressurizer level to drop.

A sensitivity study was performed modeling the FWLB transient with the first safety grade trip failed and a loss ofoffsite power (LOOP). The LOOP is conservatively considered to occur coincident with the turbine trip thatfollows reactor trip. The case with a LOOP included the insertion of the safety control rod banks on LOOP. Thesensitivity study determined that the peak RCS pressure from a LOOP case is less limiting than the peak RCSpressure without a LOOP. Therefore, the FWLB evaluation using the DSS trip function and no LOOP is thebounding case.

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Table 4-1: Sequence of Events for FWLB with First Trip Failed

Parameter Time, sec

Transient Initiated 0.0MFW to unaffected SG Interrupted 0.01Peak Thermal Power Occurs 7.59

DSS High RCS Pressure Trip Setpoint Reached 10.26EFIC Actuated on Low SG-B Level 10.42Regulating Control Rod Groups Begin to Insert 11.49Turbine Trip, TSVs Begin to Close 11.49Initial PSV Lift -12.5

Peak RCS Pressure occurs 15.02Pressurizer becomes liquid solid -50

Affected SG depressurization complete -50EFW to Unaffected SG Begins 50.43Final PSV closure -170Peak Tave occurs -240Transient Analysis Ends 600

Table 4-2: Results for FWLB with First Trip Failed

Parameter ValuePeak RCS pressure (psia) 2915.39Peak thermal power (%RTP) 100.51Peak Tave ('F) 622.22

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Figure 4-1: FWLB with Fail First Trip - RCS Pressure3000

n Hot Leg I (CV 110-4)- Hot Leg 2 (CV 210-4)A Core Exit (CV 352-01)

2900 ...........- Top of PZR (CV 405-01)-- Lower RV.Downcomer (CV 324-04)

2800- -- ---- -- ---

2700O-- - .... ....2600 .............................. ..... .................. ............... .. -....... ...... ......

2600

2 2 0 .. .. ...... ... ... ....... .....

2500

200630 60 120 180 240 300 360 420 480 540 600

Time (s)

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Figure 4-2: FWLB with Fail First Trip - Reactor Power3200

-- a Total Reactor Power (CVAR 370)-- Thermal Power (CVAR 379)

2800

2400

2000 . ... .. .. ..

o 1600

1200

800

400

60 120 180 240 300 360 420 480 540 600Time (s)

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Figure 4-3: FWLB with Fail First Trip - Reactivity

-• CVAR 1970.010

0.000 @W•

-0.010

$$0

a- -~---45A------{A - -a-- ,J -Fl fl;

-0.020

-0.030

-0.040

-0.0500 60 120 180 240 300

Time (s)360 420 480 540 600

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Figure 4-4: FWLB with Fail First Trip - Primary System Temperatures720-

RCS Ave Temp (CVAR 900)-- 0 Loop I Hot Leg (CV 110-4)

A Loop 2 Hot Leg (CV 210-4)

700 .- < Loop 1A Cold Leg (CV 160-4)--V Loop 1 B Cold Leg (CV 180-4)

E> Loop 2A Cold Leg (CV 280-4)Loop 213 Cold Leg (CV 280-4)

680 -- Top of Pressurizer Liquid Temperature (CV 405o --44-

660. ...........

640 . . ...

(-

CI)620

S 600 : ::

540 ............. i ............................. - ... ...... ..................................... ........ .......

5200 60 120 180 240 300 360 420 480 540 600

Time (s)

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4)

4)

._N

400

380

360

340

320

300

280

260

240

220

Figure 4-5: FWLB with Fail First Trip - Indicated Pressurizer Level

-4 CVAR 409

... .. .... ... .. ... .. ---- ---- --- -- ----

.. . .. . ... . ... .. . . .. . ... .. . . ---------. .

iAnn60 120 180 240 300

Time (s)360 420 480 540 600

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Figure 4-6: FWLB with Fail First Trip - Pressurizer Collapsed Liquid Level44

-- CVAR 325

42 --R: - - -•-- ,

40

38

0 36

S34

S32

30

28

26

240 60 120 180 240 300 360 420 480 540 600

Time (s)

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Figure 4-7: FWLB with Fail First Trip - Pressurizer Surge Line Flow500

-- mflow4-41 001 0000

250

0 El E 9R-----2 5 0- ----- ----. . ... . .. ... . .. --- -- --- -- -- --- -- -- --- -- --.-- -- -- --- -- -- --- -- -- --- -- --

r-250

-500

• ) -750... . ... ... ..

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Figure 4-8: FWLB with Fail First Trip - RCS Volumetric Flow Rate

Ei

U

500000

450000

400000

350000

300000

250000

200000

150000

100000

50000

-• CVAR 332

.- . . . . ... . U -.. ...... - .... . . ... . . g .. [.. . . . ... ....

060 120 180 240 300

Time (s)360 420 480 540 600

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Figure 4-9: FWLB with Fail First Trip - Pressurizer Safety Valve Flow500

--- mflowj-492000000

450

4003 5 0 . . . . . . . . . . . . .7 . . . . . . . . . . . . . .. ... . . . . . . . . . . .. .... . . . . . . . . . . . .. ..... - - -- . . . ..

~n 300

o 250 --- - ...

S 200 -- -- - --- -- -- -

4 0 0 . . . . . . . . . . . . . . . ..- .- . . . . . ... . . . . ....-- --

1 0 0 -.-.-.-. .-.- --.-. .-.-.- -.... ....... . . . .. ... . ........... ... . i.. .... .. ... ........... .. ....... ... ... . .. ...

100

50

0 60 120 180 240 300 360 420 480 540 600

Time (s)

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Figure 4-10: FWLB with Fail First Trip - SG Secondary Side Liquid Level12

SG-A (CVAR 814)~SG-B (CVAR 714)

10

-~6

0

4C# 4

60 120 180 240 300Time (s)

360 420 480 540 600

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24

0

80000

70000

60000

50000

40000

30000

20000

10000

Figure 4-11: FWLB with Fail First Trip - SG Secondary Side Inventory

-• SG-A (CVAR 607)SG-B (CVAR 707)

E3 R R U

U -0 60 120 180 240 300

Time (s)360 420 480 540 600

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100

90

80o

0

0

70

60

50

40

30

20

10

0

Figure 4-12: FWLB with Fail First Trip - SG % Operating Range

-u SG-A (CVAR 612)-1SG-B V . AR72)

.. .... .. ... .. ...... ....

.. ....... ... .. ....... ...... ......... ..........

T60 120 180 240 300

Time (s)360 420 480 540 600

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Figure 4-13: FWLB with Fail First Trip - SG Pressure1200U

Unaffected (CV 648-1) 1L Affected (CV 748-1)1 10 • 'A .( • _ _8 ! . ............. -- ---J

801000 ..... ....................................................................................................-----

1900 ----

1 000 -- -- -- .. . . . ... .. . . .. ..

700 ...... .... ....... ........... ........... ......... ....... ... .. ... ........ .. ........

W 300 -- ---............. ......... ....... ................. . ..... . ... . ........................ .... ...

S500

400

300 .... ..

200 - -- ... .....

100 - --.. .

60 120 180 240 300 360 420 480 540 600Time (s)

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Figure 4-14: FWLB with Fail First Trip - EFW FlowIOU

To Unaffected SG (JUN 626)--o To Affected SG (JUN 726)

.2

90

80

70

60

50

40

30

20

10

.......... - ...

a- - - ; ul- e

o 60 120 180 240 300

Time (s)360 420 480 540 600

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Figure 4-15: FWLB with Fail First Trip - Integrated MSSV Flow100000

o Unaffected SG (CVAR 669)-- Affected SG (CVAR 660)

90000

80000

70000

E!

60000

CIO 50000IT

40000

30000

20000

10000

0-0 60 120 180 240 300 360 420 480 540 600

Time (s)

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5.0 REFERENCES

1. BAW-1 01 64PA-06, "RELAP5/MOD2-B&W - An Advanced Computer Program for Light Water ReactorLOCA and Non-LOCA Transient Analysis."

2. BAW-1 01 93PA-00, "RELAP5/MOD2-B&W for Safety Analysis of B&W-Designed Pressurizer WaterReactors."

3. ANP-3114NP-000, "CR-3 EPU - Feedwater Line Break Analysis Sensitivity Studies."

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Documents

Crystal River, Unit 3 - Attachment E to 3F0712-03 ... · location is a 1.418 ft2 break in one of the side branches of steam generator B at the junction connecting the main feedwater