3F0712-03, Attachment E to 3F0712-03, Technical Report, ANP-3052, Rev. 2, CR-3 EPU Feedwater Line Break Analysis with Failure of First Safety Grade Trip.

Attachment E to 3F0712-03, Technical Report, ANP-3052, Rev. 2, CR-3 EPU Feedwater Line Break Analysis with Failure of First Safety Grade Trip.

ML12205A359
Person / Time
Site:	Crystal River
Issue date:	06/30/2012
From:	AREVA NP, Progress Energy Florida
To:	Office of Nuclear Reactor Regulation
References
TAC ME6527, 3F0712-03 ANP-3052, Rev. 2
Download: ML12205A359 (32)

Text

{{#Wiki_filter: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! 111 ANP-3052 Revision 2 CR-3 EPU Feedwater Line Break Analysis with Failure of First Safety Grade Trip June 2012 AREVA NP Inc. A AREVA

Controlled Document ANP-3052 Revision 2 June 2012 CR-3 EPU Feedwater Line Break Analysis with Failure of First Safety Grade Trip Copyright © 2012 AREVA Inc. All Rights Reserved Page 2

A AREVA ANP-3052 CR-3 EPU Feedwater Line Break Analysis Revision 2 with Failure of First Safety Grade Trip Page 3 Record of Revision Revision PageslSections/Paragraphs No. Changed Brief Description I Change Authorization 000 All Initial Release 001 All The FWLB analysis with the first safety grade trip failed was re-evaluated using the model and limiting conditions from 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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A AREVA ANP-3052 CR-3 EPU Feedwater Line Break Analysis Revision 2 with Failure of First Safety Grade Trip Page 4 Table of Contents 1.0 Introduction ....................................................................................................................................... 8 2.0 Analytical Methodology ............................................................................................................ 9 3.0 Analysis Inputs ................................................................................................................................ 10 4.0 Results / Conclusions ..................................................................................................................... 14 5.0 References ..................................................................................................................................... 31 Page 4

A AREVA ANP-3052 CR-3 EPU Feedwater Line Break Analysis Revision 2 with Failure of First Safety Grade Trip Page 5 List of Tables Table 3-1: Input to Feedwater Line Break Analysis with First Trip Failed ........................................... 12 Table 4-1: Sequence of Events for FWLB with First Trip Failed ......................................................... 15 Table 4-2: Results for FW LB with First Trip Failed ............................................................................... 15 Page 5

A AR EVA ANP-3052 CR-3 EPU Feedwater Line Break Analysis Revision 2 with Failure of First Safetv Grade Trio Page 6 List of Figures Figure 3-1: Feedwater Line Break Locations ........................................................................................ 11 Figure 4-1: FWLB with Fail First Trip - RCS Pressure ........................................................................ 16 Figure 4-2: FWLB with Fail First Trip - Reactor Power ........................................................................ 17 Figure 4-3: FWLB with Fail First Trip - Reactivity ................................................................................. 18 Figure 4-4: FWLB with Fail First Trip - Primary System Temperatures .............................................. 19 Figure 4-5: FWLB with Fail First Trip - Indicated Pressurizer Level ..................................................... 20 Figure 4-6: FWLB with Fail First Trip - Pressurizer Collapsed Liquid Level ......................................... 21 Figure 4-7: FWLB with Fail First Trip - Pressurizer Surge Line Flow ................................................... 22 Figure 4-8: FWLB with Fail First Trip - RCS Volumetric Flow Rate ...................................................... 23 Figure 4-9: FWLB with Fail First Trip - Pressurizer Safety Valve Flow ................................................ 24 Figure 4-10: FWLB with Fail First Trip - SG Secondary Side Liquid Level .......................................... 25 Figure 4-11: FWLB with Fail First Trip - SG Secondary Side Inventory .............................................. 26 Figure 4-12: FWLB with Fail First Trip - SG % Operating Range ........................................................ 27 Figure 4-13: FWLB with Fail First Trip - SG Pressure ........................................................................ 28 Figure 4-14: FWL1B with Fail First Trip - EFW Flow ............................................................................. 29 Figure 4-15: FWLB with Fail First Trip - Integrated MSSV Flow ......................................................... 30 Page 6

A AR EVA ANP-3052 CR-3 EPU Feedwater Line Break Analysis Revision 2 with Failure of First Safety Grade Trip Page 7 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 Page 7 M

H A AR EVA ANP-3052 CR-3 EPU Feedwater Line Break Analysis Revision 2 with Failure of First Safety Grade Trip Page 8

1.0 INTRODUCTION

The Nuclear Regulatory Commission (NRC) Reactor Systems Branch staff requested an additional analysis be performed to support the review of the Crystal River Unit 3 (CR-3) extended power uprate (EPU) licensing amendment 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 trip function 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 peak RCS pressure during the FWLB transient is reported and compared to an acceptance criterion of 120% of the reactor 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 and conclusions in Reference (3]. The model refinements described in Section 4.1 of Reference [3] and the limiting initial conditions described in Section 4.12 of Reference [3] are included in the model for evaluating the first safety grade trip being failed. Page 8

A AREVA ANP-3052 CR-3 EPU Feedwater Line Break Analysis Revision 2 with Failure of First Safety Grade Trip Page 9 2.0 ANALYTICAL METHODOLOGY The thermal-hydraulic analysis of the FWLB transient at the CR-3 EPU power level with the first safety grade trip failed is performed using the RELAP5/MOD2-B&W computer program (Reference [1]). The code simulates RCS and secondary system operation. The reactor core model is based on a point kinetics solution with reactivity feedback for control rod assembly insertion, fuel temperature changes, moderator temperature changes, and changes in boron concentration. The RCS model provides for heat transfer from the core, transport of the coolant to the steam generators (SG), and heat transfer to the steam generators. The secondary model includes a detailed depiction of the main steam system, including steam relief to the atmosphere through the main steam safety valves (MSSVs) and simulation of the turbine stop valves (TSVs). The secondary model also includes the delivery 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 the non-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 are based on the FWLB sensitivity studies documented in Reference (3]. The conditions in Reference [3] that result 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 %FP

d. Beginning of Cycle (BOC)

e. 585 °F Average RCS Temperature (Tave)

f. 940 psia turbine header pressure

g. 80 %OR SG level

h. 290 inch indicated pressurizer level Page 9

A AR EVA ANP-3052 CR-3 EPU Feedwater Line Break Analysis Revision 2 with Failure of First Safety Grade Trip Page 10 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 without pressurizer 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 containment pressure during a FWLB. The DSS trip delay time is modeled as 1.23 seconds. Finally, the rod worth available to the DSS system is modeled as only 2.0 %AkIk, since the DSS system only inserts the regulating control rod groups. The rod worth is based on the regulating control rod groups having a worth of 2.2 %Ak/k at the rod insertion limits. The worth is then reduced by 10% (0.2 %Ak/k) for uncertainty. An explicit 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 trip analysis was performed at the three limiting break locations from that study, to ensure that the limiting break location is captured. The break locations considered are shown in Figure 3-1. The peak RCS pressures for the three break locations ranged from 2911.61 psia to 2915.39 psia. The limiting break location is a 1.418 ft 2 break in one of the side branches of steam generator B at the junction connecting the 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 the changes highlighted above. The peak RCS pressure is reached shortly after reactor trip before any active safety systems such as emergency feedwater (EFW) are credited to mitigate the results. Therefore, there are no single failure assumptions that would result in a more limiting peak RCS pressure. However, the FWLB overpressure protection analysis with the first safety grade trip failed assumes the same single failure assumption as the FWLB analysis performed to support Section 2.8.5.2.4 of the CR-3 EPU LAR. The single failure assumption modeled is the failure of one train of Emergency Feedwater Initiation and Control (EFIC) such that EFW flow is not initiated automatically in one train. Consequently, only one of the two EFW pumps is assumed available to provide flow to the SGs. This single failure assumption produces a conservative long-term transient response. Page 10

A AR EVA ANP-3052 CR-3 EPU Feedwater Line Break Analysis Revision 2 with Failure of First Safety Grade Trio Page 11 Figure 3-1: Feedwater Line Break Locations Isolation Check Valve. 18" 14" 14" SG Page 11

A AREVA ANP-3052 CR-3 EPU Feedwater Line Break Analysis Revision 2 with Failure of First Safety Grade Trip Page 12 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/hr 1 @ 583,574 Ibm/hr MSSV Nominal Setpoints 2 @ 1050 psig 2 @ 1070 psig 2 @ 1090 psig 2 @ 1100 psig including small MSSV MSSV Setpoint Tolerance +3%

MSSV Accumulation +3% Page 12

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A AREVA ANP-3052 CR-3 EPU Feedwater Line Break Analysis Revision 2 with Failure of First Safety Grade Trip Page 13 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 Page 13

A AREVA ANP-3052 CR-3 EPU Feedwater Line Break Analysis Revision 2 with Failure of First Safety Grade Trip Page 14 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 the calculated results are tabulated in Table 4-2. Plots that demonstrate the transient response following a FWLB are provided 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 heat removal. The mismatch between energy addition to the reactor coolant and the secondary heat removal causes the reactor coolant to heat up and pressurize. The pressure increases to the RPS high RCS pressure trip setpoint, but the trip is assumed to fail. The pressure continues to increase to the DSS high RCS pressure trip setpoint. 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, the pressure begins to decrease. The peak RCS pressure occurs in the bottom of the reactor vessel and does not exceed 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 is provided to the unaffected SG at the flow rate of one EFW pump (660 gpm). During this time, the PSVs maintain the RCS pressure based on the PSV open and close setpoints. The FWLB event is sufficiently severe that the pressurizer fills. As a result, the PSVs begin to pass single-phase liquid. The PSVs of the type installed at CR-3 achieve satisfactory performance for fluid temperatures greater than -550 0 F. An additional check is performed to show that the PSV fluid inlet temperature remains greater than 600 OF to ensure that the PSVs operate as intended. Figure 4-4 demonstrates that at all times throughout the FWLB transient, the liquid temperature at the top 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 of offsite power (LOOP). The LOOP is conservatively considered to occur coincident with the turbine trip that follows reactor trip. The case with a LOOP included the insertion of the safety control rod banks on LOOP. The sensitivity study determined that the peak RCS pressure from a LOOP case is less limiting than the peak RCS pressure without a LOOP. Therefore, the FWLB evaluation using the DSS trip function and no LOOP is the bounding case. Page 14

A AREVA ANP-3052 CR-3 EPU Feedwater Line Break Analysis Revision 2 with Failure of First Safetv Grade Trio Page 15 Table 4-1: Sequence of Events for FWLB with First Trip Failed Parameter Time, sec Transient Initiated 0.0 MFW to unaffected SG Interrupted 0.01 Peak Thermal Power Occurs 7.59 DSS High RCS Pressure Trip Setpoint Reached 10.26 EFIC Actuated on Low SG-B Level 10.42 Regulating Control Rod Groups Begin to Insert 11.49 Turbine Trip, TSVs Begin to Close 11.49 Initial PSV Lift -12.5 Peak RCS Pressure occurs 15.02 Pressurizer becomes liquid solid -50 Affected SG depressurization complete -50 EFW to Unaffected SG Begins 50.43 Final PSV closure -170 Peak Tave occurs -240 Transient Analysis Ends 600 Table 4-2: Results for FWLB with First Trip Failed Parameter Value Peak RCS pressure (psia) 2915.39 Peak thermal power (%RTP) 100.51 Peak Tave ('F) 622.22 Page 15

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A AR EVA ANP-3052 CR-3 EPU Feedwater Line Break Analysis Revision 2 with Failure of First Safety Grade Trip Page 16 Figure 4-1: FWLB with Fail First Trip - RCS Pressure 3000 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-                                   --                                                        -- --- -                                                              - --

2600 .............................. ..... -....... 2700O-- - 2600 2500 22 0 .. ... ... ........ .. ... .. .. ... 200630 60 120 180 240 300 360 420 480 540 600 Time (s) Page 16

(-It "',(JIý A AREVA ANP-3052 CR-3 EPU Feedwater Line Break Analysis Revision 2 with Failure of First Safety Grade Trip Page 17 Figure 4-2: FWLB with Fail First Trip - Reactor Power 3200

                                                                       -- a Total Reactor Power (CVAR 370)
                                                                       -- Thermal Power (CVAR 379) 2800 2400 2000                                                                                  . ... .. .. ..

o 1600 1200 800 400 60 120 180 240 300 360 420 480 540 600 Time (s) Page 17

A AREVA ANP-3052 CR-3 EPU Feedwater Line Break Analysis Revision 2 with Failure of First Safety Grade Trio Page 18 Figure 4-3: FWLB with Fail First Trip - Reactivity 0.010

                                                                                    -*  CVAR 197 0.000 @W*

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

      -0.010
  $   -0.020 0
      -0.030
      -0.040
      -0.050 0        60       120       180              240      300    360    420            480  540       600 Time (s)

Page 18

A AREVA ANP-3052 CR-3 EPU Feedwater Line Break Analysis Revision 2 with Failure of First Safety Grade Trip Page 19 Figure 4-4: FWLB with Fail First Trip - Primary System Temperatures 720-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 1B Cold Leg (CV 180-4)

E>Loop 2A Cold Leg (CV 280-4) Loop 213Cold Leg (CV 280-4) 680 -- Top of Pressurizer Liquid Temperature 44- (CV 405o -- 660. ........... 640 .. ... (- 620 CI) S 600  :  :: 540.............

                  - i.............................

520 0 60 120 180 240 300 360 420 480 540 600 Time (s) Page 19

A AREVA ANP-3052 CR-3 EPU Feedwater Line Break Analysis Revision 2 with Failure of First Safety Grade TriD Page 20 Figure 4-5: FWLB with Fail First Trip - Indicated Pressurizer Level 400

                                                                       -4 CVAR 409 380                      ...                                          .. .... ... .. ... ..            ---- ---- ---------

360 340 4) 320 4) 300

 ._N 280 260 240 220 iAnn 60        120      180    240   300    360     420                   480                     540                 600 Time (s)

Page 20 I

A AREVA ANP-3052 CR-3 EPU Feedwater Line Break Analysis Revision 2 with Failure of First Safety Grade Trip Page 21 Figure 4-6: FWLB with Fail First Trip - Pressurizer Collapsed Liquid Level 44

                                                                  -- CVAR 325 42               --R:- -                                             -*--    ,

40 38 0 36 S34 S32 30 28 26 24 0 60 120 180 240 300 360 420 480 540 600 Time (s) Page 21

A AREVA ANP-3052 CR-3 EPU Feedwater Line Break Analysis Revision 2 with Failure of First Safety Grade Trip Page 22 Figure 4-7: FWLB with Fail First Trip - Pressurizer Surge Line Flow 500

                                                                                 --   mflow4-41 001 0000 250
           -250-                              . . . ..... ...

0 El E 9R---- r-250

          -500
    * ) -750...                                  .   ...                ...                        ..

Page 22

Ai,k nIr A AREVA ANP-3052 CR-3 EPU Feedwater Line Break Analysis Revision 2 with Failure of First Safety Grade TriOD Page 23 Figure 4-8: FWLB with Fail First Trip - RCS Volumetric Flow Rate 500000

                                                                                                              -* CVAR 332 450000
                      .-                    . . . . ... . U -..  - .......... . . ... . . g..      . ..           ... ....           [..

400000 350000 300000 Ei 250000 U 200000 150000 100000 50000 0 60 120 180 240 300 360 420 480 540 600 Time (s) Page 23

A AREVA ANP-3052 CR-3 EPU Feedwater Line Break Analysis Revision 2 with Failure of First Safety Grade Trip Page 24 Figure 4-9: FWLB with Fail First Trip - Pressurizer Safety Valve Flow 500

                                                                                                                                                  --- mflowj-492000000 450 3 50   . . . . .. . ... .. .7 . . .. . ... . . .. . .. ...

400

   ~n 300 40 0 . . . .. . .      .     ..       ..    . . . ..- .-. .  . ..      ...                                         --                                 ..     .        . ....--

o 250 --- - ... 100 .-.-.-.-.---.-. ....--.......

                             -.-.-.-.                                        -. - ---.--. ..--...- . ........... ... . .... ..i..... ........... .. .......

S 200 100 50 0 60 120 180 240 300 360 420 480 540 600 Time (s) Page 24

A AREVA ANP-3052 CR-3 EPU Feedwater Line Break Analysis Revision 2 with Failure of First Safety Grade Trip Page 25 Figure 4-10: FWLB with Fail First Trip - SG Secondary Side Liquid Level 12 SG-A (CVAR 814)

                                                                 ~SG-B (CVAR 714) 10
   -~6 0

C# 44 60 120 180 240 300 360 420 480 540 600 Time (s) Page 25

A AREVA ANP-3052 CR-3 EPU Feedwater Line Break Analysis Revision 2 with Failure of First Safety Grade Trip Page 26 Figure 4-11: FWLB with Fail First Trip - SG Secondary Side Inventory 80000

                                                                     -* SG-A (CVAR 607)

SG-B (CVAR 707) 70000 60000 24 50000 0 40000 30000 20000 10000 E3 R R U U - 0 60 120 180 240 300 360 420 480 540 600 Time (s) Page 26

A AREVA ANP-3052 CR-3 EPU Feedwater Line Break Analysis Revision 2 with Failure of First Safety Grade Trip Page 27 Figure 4-12: FWLB with Fail First Trip - SG % Operating Range 100

                                                                  -u SG-A (CVAR 612)
                                                                  -1SG-B
                                                                       . VAR72) 90 80o 70 0    60 50 0

40 30 20 10 0 T 60 120 180 240 300 360 420 480 540 600 Time (s) Page 27

A AREVA ANP-3052 CR-3 EPU Feedwater Line Break Analysis Revision 2 with Failure of First Safety Grade Trip Page 28 Figure 4-13: FWLB with Fail First Trip - SG Pressure 1200U 1 10 L Unaffected (CV 648-1) Affected .((CV

                                                                                                                                                                               * 'A

• 748-1) 1

_ _8 ! . ............. J -- --- 1900 ---- 000 ------ ..... 1801000 ....................................................................................................----- 700 ...... .... ....... ........... ........... ......... ....... ... .. ... ........ .. ........ S500 W 300 -- ---............. . ..... ......... . .......... ................. 400 300 .... .. 200 - -- ... .. .. . 100 - --.. . 60 120 180 240 300 360 420 480 540 600 Time (s) Page 28

ri t

                                                                     ,ýk A

AREVA ANP-3052 CR-3 EPU Feedwater Line Break Analysis Revision 2 with Failure of First Safety Grade TriD Page 29 Figure 4-14: FWLB with Fail First Trip - EFW Flow IOU To Unaffected SG (JUN 626)

                                                                              --o To Affected SG (JUN 726) a- - - ; ul-     e 90 80                     - ...

70 60 50

.2 40 30 20 10 o         60              120       180    240      300   360      420            480         540        600 Time (s)

Page 29

A AR EVA ANP-3052 CR-3 EPU Feedwater Line Break Analysis Revision 2 with Failure of First Safety Grade Trip Page 30 Figure 4-15: FWLB with Fail First Trip - Integrated MSSV Flow 100000 o Unaffected SG (CVAR 669)

                                                                      --   Affected SG (CVAR 660) 90000 80000 70000 E!

60000 CIO 50000 IT 40000 30000 20000 10000 0-0 60 120 180 240 300 360 420 480 540 600 Time (s) Page 30

A AREVA ANP-3052 CR-3 EPU Feedwater Line Break Analysis Revision 2 with Failure of First Safety Grade Trip Page 31

5.0 REFERENCES

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

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

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

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