m-relap5 code supplementary manual volume ii: code …6as-1 e-uap-1 00001 (ro) m-relap5...

ENCLOSURE4

UAP-HF-10004

6AS-1 E-UAP- 100001 (RO)

M-RELAP5 Code Supplementary ManualVolume II: Code Assessment

January 2010(Non-Proprietary)

6AS-1 E-UAP-1 00001 (RO)Non-Proprietary VersionM-RELAP5 Supplementary Manual Vol. III

M-RELAP5 Code Supplementary ManualVolume II: Code Assessment

Non-Proprietary Version

January 2010

©2010 Mitsubishi Heavy Industries, Ltd.All Rights Reserved

Mitsubishi Heavy Industries, LTD.

6AS-1 E-UAP-1 00001 (RO)M-RELAP5 Supplementary Manual Vol. III Non-Proprietary Version

© 2010MITSUBISHI HEAVY INDUSTRIES, LTD.

All Rights Reserved

This document has been prepared by Mitsubishi Heavy Industries, Ltd. (CMHI") inconnection with its request to the U.S. Nuclear Regulatory Commission ("NRC") for adesign certification review of the US-APWR nuclear power plant design. No right todisclose, use or copy any of the information in this document, other that by the NRC andits contractors in support of MHI's design certification of the US-APWR, is authorizedwithout the express written permission of MHI.

This document contains technology information and intellectual property relating to theUS-APWR and it is delivered to the NRC on the express condition that it not be disclosed,copied or reproduced in whole or in part, or used for the benefit of anyone other than MHIwithout the express written permission of MHI, except as set forth in the previousparagraph.

This document is protected by the laws of Japan, U.S. copyright law, international treatiesand conventions, and the applicable laws of any country where it is being used.

Mitsubishi Heavy Industries, Ltd.16-5, Konan 2-chome, Minato-ku

Tokyo 108-8215 Japan



ABSTRACT

The M-RELAP5 code has been developed by the Mitsubishi Heavy Industries, Ltd. (MHI)and currently used for the safety analyses of US-APWR plant in responding to a set ofpostulated small break LOCA transients. The code is based on the Best-EstimateRELAP5-3D code originally developed at the Idaho National Laboratory. The MHI modifiedthe code to incorporate the requirements set forth in the Appendix K to 10 CFR 50 forconservative safety analyses approach adopted for the SBLOCA analyses of theUS-APWR. The base code of RELAP5 has a long history and has been widely assessedusing experimental data obtained from various Separate Effects Test (SET) and IntegralEffects Test (lET) facilities for the application to PWR SBLOCAs as a best-estimate code.Taking into account the major modifications from the original code, the M-RELAP5 isapplicable as a conservative code for PWR's SBLOCA analyses.

The M-RELAP5 has been developed and assessed in conformance to the EvaluationModel Development and Assessment (EMDAP) by the U.S. Nuclear RegulatoryCommission. Important phenomena and processes occurring under the specified accidentand transient are identified first, and then code models, correlations, and capabilitiesrelated to the important phenomena and processes are verified and validated by usingappropriate experimental data obtained from SET and lET facilities scalable to thespecified plant. In addition to the developmental assessment for the RELAP5-3D reportedin the original manual, this supplementary manual describes the developmentalassessment strategy adopted by MHI for the M-RELAP5 code, and presents the results ofthe code assessment using various SET and lET data.


6AS-1 E-IJAP-1 00001 (RO)M-RELAP5 Supplementary Manual Vol. III Non-Proprietary Version

TABLE OF CONTENTS

A B S T R A C T .......................................................................................................................... iTABLE O F CONTENTS ....................................................................................................... iiLIST O F TABLES ...................................................................... b ......................................... ivLIST OF FIGURES ............................................................................................................... vLIST OF ACRONYM S ....................................................................................................... xiv1. INTRODUCTION .......................................................................................................... 1-12. CO DE QUALIFICATION STRATEGY .......................................................................... 2-1

2.1 Phenom ena Identification and Ranking Table ........................................................ 2-12.2 Code Assessment M atrix ......................................................................................... 2-32.3 References ............................................................................................................. 2-6

3. SEPARATE EFFECTS TESTS ..................................................................................... 3-13.1 ROSA-IV/LSTF Void Profile Test ............................................................................ 3-1

3.1.1 Test Description ................................................................................................ 3-13.1.2 Code Validation ................................................................................................ 3-23.1.3 Sum m ary .......................................................................................................... 3-3

3.2 ORNUTHTF Void Profile and Uncovered-Bundle Heat Transfer Tests ................. 3-103.2.1 Test Description .............................................................................................. 3-103.2.2 Code Validation .............................................................................................. 3-113.2.3 Sum m ary ........................................................................................................ 3-12

3.3 ORNLITHTF High-Pressure Reflood Test ............................................................. 3-453.3.1 Test Description .............................................................................................. 3-453.3.2 Code Validation .............................................................................................. 3-453.3.3 Sum m ary ........................................................................................................ 3-47

3.4 FLECHT-SEASET Forced-Reflood Test ............................................................... 3-603.4.1 Test Description .............................................................................................. 3-603.4.2 Code Validation .............................................................................................. 3-623.4.3 Sum m ary ........................................................................................................ 3-63

3.5 UPTF CCFIL Test ................................................................................................... 3-703.5.1 Test Description .............................................................................................. 3-703.5..2 Code Validation ............................................................................................... 3-723.5.3 Sum m ary ........................................................................................................ 3-73

3.6 DuklerAir-W ater Flooding Test ............................................................................. 3-813.6.1 Test Description .............................................................................................. 3-813.6.2 Code Validation .............................................................................................. 3-833.6.3 Sum m ary ........................................................................................................ 3-83

3.7 UPTF Test 5 .......................................................................................................... 3-913.7.1 Test Description .............................................................................................. 3-913.7.2 Code Validation .............................................................................................. 3-923.7.3 Sum m ary ........................................................................................................ 3-93

3.8 Advanced Accum ulator Test .................................................................................. 3-983.8.1 Test Description .............................................................................................. 3-983.8.2 Code Validation .............................................................................................. 3-983.8.3 Sum m ary ........................................................................................................ 3-99

3.9 References ......................................................................................................... 3-1074. INTEGRAL EFFECTS TESTS ...................................................................................... 4-1

4.1 ROSA-IV/LSTF SB-CL-1 8 Test ............................................................................... 4-14.1.1 Test Description ................................................................................................ 4-14.1.2 Code Validation ................................................................................................ 4-3

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4.1.3 Steam Generator Heat Transfer Validation ...................................................... 4-84.1.4 Sum m ary .......................................................................................................... 4-9

4.2 ROSA-IV/LSTF SB-CL-09 Test ............................................................................. 4-454.2.1 Test Description .............................................................................................. 4-454.2.2 Code Validation .............................................................................................. 4-464.2.3 Sum m ary ....................................................................................................... 4-46

4.3 ROSA-IV/LSTF SB-CL-12 Test ............................................................................. 4-554.3.1 Test Description .............................................................................................. 4-554.3.2 Code Validation .............................................................................................. 4-564.3.3 Sum m ary ........................................................................................................ 4-56

4.4 LOFT L3-1 Test ..................................................................................................... 4-654.4.1 Test Description .............................................................................................. 4-654.4.2 Code Validation .............................................................................................. 4-674.4.3 Sum m ary ........................................................................................................ 4-69

4.5 SEM ISCALE S-LH-1 Test ...................................................................................... 4-794.5.1 Test Description .............................................................................................. 4-794.5.2 Code Validation ............................................................ 4-804.5.3 Sum m ary ........................................................................................................ 4-82

4.6 References ........................................................................................................... 4-925. REVIEW FOR CO DE ASSESSM ENT RESULTS ........................................................ 5-16. CONCLUSIONS ........................................................................................................... 6-1

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LIST OF TABLES

Table 2.1-1 PIRT for US-APWR SBLOCA (High Rank) .................................................... 2-2

Table 2.2-1 M-RELAP5 Code Assessment Matrix for US-APWR SBLOCAs ................... 2-5

Table 3.1-1 Primary Core Section Specifications for ROSA-IV ........................................ 3-4

Table 3.1-2 ROSA-IV/LSTF Void Profile Test Conditions ................................................. 3-4

Table 3.2-1 Primary Test Section Specifications for ORNL/THTF .................................. 3-13

Table 3.2-2 ORNL/THTF Void Profile and Uncovered-Bundle Heat Transfer Test

C o n d itio n s ............................................................................................... 3 -14

Table 3.3-1 ORNL/THTF High-Pressure Reflood Test Conditions ................................. 3-47

Table 3.4-1 Primary Test Section Specifications for FLECHT-SEASET ......................... 3-63

Table 3.4-2 FLECHT-SEASET Forced-Reflood Test Conditions .................................... 3-63

Table 3.5-1 Comparison of UPTF Hot Leg Configuration with Typical Westinghouse and

Combustion Engineering (CE) PWRs ..................................................... 3-74

Table 3.5-2 Flowrate Conditions for UPTF CCFL Test Analysis ..................................... 3-74

Table 3.5-3 Results of UPTF CCFL Test Analysis (1 5bar) .............................................. 3-74

Table 3.5-4 Results of UPTF CCFL Test Analysis (3bar)..: ............................................. 3-74

Table 3.6-1 Four Different Input Liquid Flowrates for Dukler Air-Water Flooding Test .... 3-84

Table 3.6-2 Experimental Results for Dukler Air-Water Flooding Test ............................ 3-84

Table 3.7-1 Boundary and Initial Conditions for UPTF Test 5 ......................................... 3-93

Table 3.8-1 Full-Height 1/2-Scaled Advanced Accumulator Test Conditions ....... 3-100

Table 4.1-1 Primary Test Facility Specification for ROSA-IV/LSTF ................................ 4-10

Table 4.1-2 Steady-State Parameters for ROSA-IV/LSTF SB-CL-18 ............................. 4-11

Table 4.1-3 Operational Setpoints for ROSA-IV/LSTF SB-CL-18 ................................... 4-11

Table 4.1-4 ECCS Conditions for ROSA-IV/LSTF SB-CL-18 ......................................... 4-12

Table 4.1-5 Core Power Decay Curve for ROSA-IV/LSTF ............................................. 4-13

Table 4.1-6 Primary Test Chronology for ROSA-IV/LSTF SB-CL-18 .............................. 4-14



Table 4.4-1 Steady-State Parameters for LOFT/L3-1 ..................................................... 4-70

Table 4.4-2 Primary Test Chronology for LOFT/L3-1 ...................................................... 4-70

Table 4.5-1 Steady-State Parameters for Semiscale/S-LH-1 ......................................... 4-83

Table 4.5-2 Primary Test Chronology for Semiscale/S-LH-1 .......................................... 4-83

Table 4.5-3 Summary of PCTs during Loop Seal for Semiscale/S-LH-1 ........................ 4-83

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LIST OF FIGURES

Figure 3.1-1 R O SA-IV/LSTF Test Facility ......................................................................... 3-5

Figure 3.1-2 Axial Power Profile and Location of Pressure Instrumentations .................. 3-5

Figure 3.1-3 M-RELAP5 Noding Scheme for ROSA-IV/LSTF Void Profile Analysis ........ 3-6

Figure 3.1-4 Simulated Void Fraction Profile for Test ST-NC-06E .................................... 3-7

Figure 3.1-5 Simulated Averaged Void Fraction Profile for 7.3 MPa Tests ....................... 3-7

Figure 3.1-6 Simulated Void Fraction Profile for Test ST-VF-01D .................................... 3-8

Figure 3.1-7 Simulated Averaged Void Fraction Profile for 1.0 MPa Tests ....................... 3-8

Figure 3.1-8 Comparison of Predicted and Measured Void Fractions for ROSA/LSTF and

O R N L/THTF Void Profile Tests ................................................................. 3-9

Figure 3.2-1 THTF in Small-Break Test Configuration .................................................... 3-15

Figure 3.2-2 Cross Section of THTF Test Section .......................................................... 3-16

Figure 3.2-3 Cross Section of Typical Fuel Rod Simulator ............................................. 3-17

Figure 3.2-4 Axial Location of Spacer Grids and FRS Thermocouples .......................... 3-18

Figure 3.2-5 THTF In-Bundle Pressure Instrumentation ................................................ 3-19

Figure 3.2-6 M-RELAP5 Noding Scheme for ORNL/THTF Test Analysis ...................... 3-20

Figure 3.2-7 Comparison of Predicted and Measured Void Fraction Profiles for

O R N L/T H T F Test 3.09.1O J ...................................................................... 3-21

Figure 3.2-8 Comparison of Predicted and Measured FRS Surface Temperature Profiles

for O R N L/THTF Test 3.09.1OJ ................................................................ 3-22

Figure 3.2-9 Comparison of Predicted and Measured Vapor Temperature Profiles for

O R N L ITH TF Test 3.09.1OJ ...................................................................... 3-23

Figure 3.2-10 Comparison of Predicted and Measured Heat Transfer Coefficient Profiles

for O RN L/THTF Test 3.09.10J ................................................................ 3-24


O R N LIT HT F Test 3.09.10K ..................................................................... 3-25


for O R N L/THTF Test 3.09.1 OK ............................................................... 3-26


O R N L/T H TF Test 3.09.10 K ..................................................................... 3-27


for ORNL/THTF Test 3.09.10K ................................ 3-28


O R N L/THTF Test 3.09.1O M .................................................................... 3-29

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for O RNLITHTF Test 3.09.1O M ............................................................... 3-30


O R NL/T H TF Test 3.09.10M .................................................................... 3-31


for ORNL/THTF Test 3.09.1OM ........................... 3-32


O R N L/THTF Test 3.09.1O N ..................................................................... 3-33


for O RNLITHTF Test 3.09.1 ON ................................................................ 3-34


O R N L/T HTF Test 3.09.1O N ..................................................................... 3-35


for O RNLITHTF Test 3.09.1O N ................................................................ 3-36


O R N L/THTF Test 3.09.1O AA ................................................................... 3-37


O R N L/THTF Test 3.09.1O BB ................................................................... 3-38


O R N L/THTF Test 3.09.10C C .................................................................. 3-39


O R N LFTHTF Test 3.09.1O DD .................................................................. 3-40


O R N L/THTF Test 3.09.1O EE ................................................................... 3-41


O R N L/THTF Test 3.09.10FF ................................................................... 3-42

Figure 3.2-29 Comparison of Predicted and Measure Collapsed Liquid Levels for

O R N LIT H T F Tests ................................................................................... 3-4 3

Figure 3.2-30 Comparison of Predicted and Measure Mixture Levels for ORNL/THTF

T e s ts ......................................................................................................... 3 -4 4

Figure 3.3-1 M-RELAP5 Noding Scheme for ORNL/THTF Reflood Test Analysis ......... 3-48

Figure 3.3-2 Test Section Inlet Flows for ORNL/THTF Test 3.09.1OP Test .................... 3-49

Figure 3.3-3 Test Section Inlet Temperatures for ORNLFTHTF Test 3.09.1OP Test ........ 3-49

Figure 3.3-4 Test Section Pressures for ORNL/THTF Test 3.09.1OP Test ..................... 3-50

Figure 3.3-5 Test Section Inlet Flows for ORNL/THTF Test 3.09.10Q Test .................... 3-50

Figure 3.3-6 Test Section Inlet Temperatures for ORNLITHTF Test 3.09.10Q Test ....... 3-51

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Figure 3.3-7 Test Section Pressures for ORNLITHTF Test 3.09.10Q Test ..................... 3-51

Figure 3.3-8 FRS and Fluid Temperatures at Level F for ORNL/THTF Test 3.09.10P Test

....................................................................... ........................................ 3 -5 2

Figure 3.3-9 FRS Temperatures at Level G for ORNLITHTF Test 3.09.1OP Test ........... 3-53

Figure 3.3-10 Collapsed Liquid Level for ORNL/THTF Test 3.09.1OP Test .................... 3-54

Figure 3.3-11 Quench Level for ORNL/THTF Test 3.09.10P Test .................................. 3-55

Figure 3.3-12 FRS and Fluid Temperatures at Level F for ORNLITHTF Test 3.09.10Q Test

................................................................................................................ 3 -5 6

Figure 3.3-13 FRS Temperatures at Level G for ORNL/THTF Test 3.09.10Q Test ........ 3-57

Figure 3.3-14 Collapsed Liquid Level for ORNLFTHTF Test 3.09.1OQ Test .................... 3-58

Figure 3.3-15 Quench Level for ORNLITHTF Test 3.09.10Q Test .................................. 3-59Figure 3.4-1 FLECHT-SEASET Flow Diagram for Forced Reflood Configuration ......... 3-64

Figure 3.4-2 FLECHT-SEASET Bundle Cross Section .................................................. 3-65

Figure 3.4-3 M-RELAP5 Noding Diagram for FLECHT-SEASET Forced-Reflood Test. 3-66

Figure 3.4-4 Rod Surface Temperature at 72-in Elevation (Run 31504) ........................ 3-67






Figure 3.5-1 Overall View of UPTF Test Facility ............................................................. 3-75

Figure 3.5-2 UPTF Hot Leg Configuration ...................................................................... 3-75

Figure 3.5-3 Configuration of International ECC Injection Pipe (Hutze) ......................... 3-76

Figure 3.5-4 UPTF Hot Leg Separate Effect Test Overall Flow Conditions .................... 3-76

Figure 3.5-5 UPTF Hot Leg Separate Effect Test Comparison of UPTF Hot Leg Void

Fractions to W allis C orrelation ................................................................ 3-77

Figure 3.5-6 Regression Analysis Results of UPTF CCFL Parameters13, c and m for

M -R E L A P 5 .............................................................................................. 3 -7 8

Figure 3.5-7 Nodalization Scheme for UPTF CCFL Test Analysis ................................. 3-79

Figure 3.5-8 UPTF CCFL Test Analysis, Comparison of Flooding Curves of Analysis

R esults and Test R esults ........................................................................ 3-80

Figure 3.6-1 Schematic of Dukler Air-Water Test Facility ............................................... 3-85

Figure 3.6-2 Air Inlet Section of Dukler Air-Water Test Facility ....................................... 3-86

Figure 3.6-3 Liquid Entrance Device of Dukler Air-Water Test Facility ........................... 3-86

Figure 3.6-4 Exit Section of Dukler Air-Water Test Facility ............................................. 3-87

Figure 3.6-5 Flooding Velocities for Air and Water in Vertical Tubes Designed to Minimize

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End Effects (Atm ospheric Pressure) ....................................................... 3-88

Figure 3.6-6 Noding Scheme for Dukler Air-Water Flooding Test Analysis .................... 3-89

Figure 3.6-7 Comparison of Calculated and Measured Results using Wallis Flooding

Correlation Constants (C=0.88 and m=1.0) ........................................... 3-90

Figure 3.6-8 Comparison of Calculated and Measured Results using Wallis Flooding

Correlation Constants (C=0.88 and m=1.0) ............................................ 3-90

Figure 3.7-1 Overall View of UPTF Test Facility ............................................................. 3-94

Figure 3.7-2 Loop Seal Portion and Instrumentation Configuration of UPTF ................. 3-94

Figure 3.7-3 Comparison of Residual Loop Seal Water Levels from UPTF Test 5 and

E xperim ental R esults .............................................................................. 3-95

Figure 3.7-4 Noding Scheme and Injection Procedure for Vapor and Water against UPTF

T e s t 5 ...................................................................................................... 3 -9 6

Figure 3.7-5 Assessment Results for Residual Water Amount in UPTF Test 5 .............. 3-97

Figure 3.8-1 Noding Scheme for Advanced Accumulator Test ..................................... 3-100

Figure 3.8-2 Injection Volumetric Flowrate for Advanced Accumulator Test (Case 1).. 3-101

Figure 3.8-3 Tank Pressure for Advanced Accumulator Test (Case 1) ......................... 3-101

Figure 3.8-4 Tank Water Level for Advanced Accumulator Test (Case 1) .................... 3-102



Figure 3.8-7 Tank Water Level for Advanced Accumulator Test (Case 2) .................... 3-103



Figure 3.8-10 Tank Water Level for Advanced Accumulator Test (Case 3) .................. 3-105

Figure 3.8-11 Injection Volumetric Flowrate for Advanced Accumulator Test (Case 4) 3-105



Figure 4.1-1 General Structure of ROSA-IV/LSTF Facility ............................................. 4-15

Figure 4.1-2 Pressure Vessel Assembly of ROSA-IV/LSTF Facility ............................... 4-16

Figure 4.1-3 Break Assembly of ROSA-IV/LSTF Facility ............................................... 4-17

Figure 4.1-4 Break Orifice of ROSA-IV/LSTF Facility .................................................... 4-17Figure 4.1-5 M-RELAP5 Noding Scheme for ROSA-IV/LSTF SBLOCA Test Analysis .. 4-18

Figure 4.1-6 M-RELAP5 Vessel Noding for ROSA-IV/LSTF SBLOCA Test Analysis ..... 4-19

Figure 4.1-7 M-RELAP5 Hot Leg Noding for ROSA-IV/LSTF SBLOCA Test Analysis ... 4-20

Figure 4.1-8 M-RELAP5 Steam Generator Noding for ROSA-IV/LSTF SBLOCA Test

A n a ly s is ................................................................................................... 4 -2 1

Figure 4.1-9 M-RELAP5 Crossover Leg Noding for ROSA-IV/LSTF SBLOCA Test Analysis

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................................................................................................................ 4 -2 2

Figure 4.1-10 Total Core Power for ROSA-IV/LSTF SB-CL-18 ...................................... 4-23

Figure 4.1-11 RCP Rotation Speed in Primary Loop-A for ROSA-IV/LSTF SB-CL-18 ... 4-23

Figure 4.1-12 RCP Rotation Speed in Primary Loop-B for ROSA-IV/LSTF SB-CL-1 8... 4-24

Figure 4.1-13 SG-A Steam Dome Pressure for ROSA-IV/LSTF SB-CL-18 .................... 4-24

Figure 4.1-14 SG-B Steam Dome Pressure for ROSA-IV/LSTF SB-CL-18 ................... 4-25

Figure 4.1-15 Break Flowrate for ROSA-IV/LSTF SB-CL-18 ......................................... 4-25

Figure 4.1-16 Pressurizer Pressure for ROSA-IV/LSTF SB-CL-18 ................................ 4-26

Figure 4.1-17 Core Differential Pressure for ROSA-IV/LSTF SB-CL-18 ........................ 4-26

Figure 4.1-18 Loop-A Crossover Leg Flowrate for ROSA-IV/LSTF SB-CL-18 ............... 4-27

Figure 4.1-19 Loop-B Crossover Leg Flowrate for ROSA-IV/LSTF SB-CL-18 ............... 4-27

Figure 4.1-20 Loop-A Hot Leg to U-Tube Top Differential Pressure for ROSA-IV/LSTF

S B -C L -1 8 ................................................................................................ 4 -2 8

Figure 4.1-21 Loop-B Hot Leg to U-Tube Top Differential Pressure for ROSA-IV/LSTF

S B -C L -1 8 ................................................................................................ 4 -2 8

Figure 4.1-22 Loop-A Hot Leg to SG Inlet Plenum Bottom Differential Pressure for

R O SA -IV /LST F S B-C L-18 ....................................................................... 4-29

Figure 4.1-23 Loop-B Hot Leg to SG Inlet Plenum Bottom Differential Pressure for

R O SA -IV/LSTF S B-C L-18 ....................................................................... 4-29

Figure 4.1-24 Loop-A SG Inlet Plenum Differential Pressure for ROSA-IV/LSTF SB-CL-18

................................................................................................................ 4 -3 0

Figure 4.1-25 Loop-B SG Inlet Plenum Differential Pressure for ROSA-IV/LSTF SB-CL-18

................................................................................................................ 4 -3 0

Figure 4.1-26 Loop-A SG U-Tube Uphill Side Differential Pressure for ROSA-IV/LSTF

S B -C L -1 8 ................................................................................................ 4 -3 1

Figure 4.1-27 Loop-B SG U-Tube Uphill Side Differential Pressure for ROSA-IV/LSTF

S B -C L -1 8 ................................................................................................ 4 -3 1

Figure 4.1-28 Loop-A SG U-Tube Downhill Side Differential Pressure for ROSA-IV/LSTF

S B -C L -1 8 ................................................................................................ 4 -3 2

Figure 4.1-29 Loop-B SG U-Tube Downhill Side Differential Pressure for ROSA-IV/LSTF

S B -C L -1 8 ................................................................................................ 4 -3 2

Figure 4.1-30 Loop-A Crossover Leg Downhill Side Differential Pressure for

R O SA -IV/LSTF S B-C L-18 ....................................................................... 4-33

Figure 4.1-31 Loop-B Crossover Leg Downhill Side Differential Pressure for

R O SA -IV/LSTF S B-C L-18 ....................................................................... 4-33

Figure 4.1-32 Loop-A Crossover Leg Uphill Side Differential Pressure for ROSA-IV/LSTF

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S B -C L -1 8 ................................................................................................ 4 -3 4

Figure 4.1-33 Loop-B Crossover Leg Uphill Side Differential Pressure for ROSA-IV/LSTF

S B -C L -1 8 ................................................................................................ 4 -3 4

Figure 4.1-34 Downcomer Differential Pressure for ROSA-IV/LSTF SB-CL-18 ............. 4-35

Figure 4.1-35 Loop-A Accumulator Injection Flowrate for ROSA-IV/LSTF SB-CL-18 .... 4-35

Figure 4.1-36 Loop-B Accumulator Injection Flowrate for ROSA-IV/LSTF SB-CL-18 .... 4-36

Figure 4.1-37 Heater Rod Surface Temperature at 3.61m (Test Data) and at 3.57m

(M-RELAP5) for ROSA-IV/LSTF SB-CL-18 ............................................ 4-36







Figure 4.1-41 Heater Rod Surface Temperature at 1.02m (Test Data) and at 1.11 m
















Figure 4.1-49 Broken Loop Secondary Pressure for ROSA-IV/LSTF SB-CL-18 by Explicit

Secondary System M odel ....................................................................... 4-42

Figure 4.1-50 Intact Loop Secondary Pressure for ROSA-IV/LSTF SB-CL-18 by Explicit


Figure 4.1-51 Primary System (Pressurizer) Pressure for ROSA-IV/LSTF SB-CL-18 by

Explicit Secondary System M odel ........................................................... 4-43

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Figure 4.1-52 Integral of SG Secondary Outlet Steam Mass for ROSA-IV/LSTF SB-CL-18

by Explicit Secondary System M odel ...................................................... 4-44

Figure 4.1-53 Core Differential Pressure for ROSA-IV/LSTF SB-CL-18 by Explicit


Figure 4.2-1 Core Differential Pressure for ROSA-IV/LSTF SB-CL-09 .......................... 4-48

Figure 4.2-2 Break Flowrate for ROSA-IV/LSTF SB-CL-09 ........................................... 4-48

Figure 4.2-3 Primary Pressure for ROSA-IV/LSTF SB-CL-09 ........................................ 4-49

Figure 4.2-4 Core Power for ROSA-IV/LSTF SB-CL-09 ................................................ 4-49

Figure 4.2-5 Heater Rod Surface Temperature at 3.61 m (Test Data) and at 3.57m



(M-RELAP5) for ROSA-IV/LSTF SB-CL-09 ........................................... 4-50











Figure 4.2-12 Heater Rod Surface Temperature at 0.61 m (Test Data) and at 0.64m




Figure 4.3-1 Core Differential Pressure for ROSA-IV/LSTF SB-CL-12 .......................... 4-58

Figure 4.3-2 Break Flowrate for ROSA-IV/LSTF SB-CL-12 ........................................... 4-58

Figure 4.3-3 Primary Pressure for ROSA-IV/LSTF SB-CL-12 ........................................ 4-59

Figure 4.3-4 Core Power for ROSA-IV/LSTF SB-CL-12 ................................................ 4-59







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6AS-1 E-UAP-1 00001 (R0)M-RELAP5 Supplementary Manual Vol. III Non-Proprietary Version












(M-RELAP5) for ROSA-IV/LSTF SB-CL-12 ............................................ 4-64Figure 4.4-1 Schematic of LOFT Major Components ..................................................... 4-71Figure 4.4-2 LO FT Core I Configuration ......................................................................... 4-72Figure 4.4-3 M-RELAP5 Noding Diagram for LOFT/L3-1 Analysis ................................ 4-73

Figure 4.4-4 Break Mass Flowrate for LOFT/L3-1 .......................................................... 4-74

Figure 4.4-5 Secondary System Pressure for LOFT/L3-1 .............................................. 4-74

Figure 4.4-6 Primary System (Upper Plenum) Pressure for LOFT/L3-1 ........................ 4-75

Figure 4.4-7 Pressurizer Liquid Level for LOFT/L3-1 ..................................................... 4-75

Figure 4.4-8 Differential Pressure in Intact Loop Crossover Leg for LOFT/L3-1 (SG-Side)

....................... ........................................ ............................................ 4 -7 6

Figure 4.4-9 Differential Pressure in Intact Loop Crossover Leg for LOFT/L3-1 (RCP-Side)

................................................................................................................ 4 -7 6

Figure 4.4-10 Accumulator Tank Pressure for LOFT/L3-1 .............................................. 4-77

Figure 4.4-11 Accumulator Tank Water Level for LOFT/L3-1 ......................................... 4-77

Figure 4.4-12 Fuel Cladding Temperature for LOFT/L3-1 (Z=62-in) .............................. 4-78

Figure 4.5-1 Semiscale Mod-2C System Configuration ................................................. 4-84

Figure 4.5-2 Semiscale Mod-2C Core Heater Rod Configuration .................................. 4-85Figure 4.5-3 M-RELAP5 Nodalization Diagram for Semiscale/S-LH-1 .......................... 4-86

Figure 4.5-4 Break Mass Flowrate for Semiscale/S-LH-1 .............................................. 4-87

Figure 4.5-5 Secondary System Pressure for Semiscale/S-LH-1 .................................. 4-87

Figure 4.5-6 Primary System Pressure for Semiscale/S-LH-1 ....................................... 4-88

Figure 4.5-7 Collapsed Level in Uphill-Side of Intact Loop Crossover Leg for

S e m isca le/S -LH -1 ................................................................................... 4 -88

Figure 4.5-8 Collapsed Level in Uphill-Side of Broken Loop Crossover Leg for

S e m isca le/S -LH -1 ................................................................................... 4 -89

Figure 4.5-9 Collapsed Level in Intact Loop Hot Leg for Semiscale/S-LH-1 .................. 4-89

Mitsubishi Heavy Industries, LTD.xii


Figure 4.5-10 Collapsed Level in Broken Loop Hot Leg for Semiscale/S-LH-1 ............. 4-90

Figure 4.5-11 Core Collapsed Level for Semiscale/S-LH-1 ............................................ 4-90

Figure 4.5-12 Core Cladding Temperature at 8.3-ft (253cm) Elevation .......................... 4-91

Mitsubishi Heavy Industries, LTD.xiii


LIST OF ACRONYMS

ACC AccumulatorAM Assessment MAPAPWR Advanced Pressurized-Water ReactorBST Blowdown Suppression TankCCFL Counter-Current Flow LimitationCHF Critical Heat FluxDC DowncomerDNB Departure from Nucleate BoilingEMDAP Evaluation Model Development and Assessment ProcessECCS Emergency Core Cooling SystemFRS Fuel Rod SimulatorHPIS High Pressure Injection SystemlET Integral Effects TestINL Idaho National LaboratoryLBLOCA Large Break Loss-of-Coolant AccidentLOCA Loss-of-Coolant AccidentLOFT Loss-of-Fluid TestLP Lower plenum of reactor vesselLPIS Low Pressure Injection SystemLSTF Large Scale Test FacilityMHI Mitsubishi Heavy Industry, Ltd.PCT Peak Cladding TemperaturePIRT Phenomena Identification and Ranking TablePWR Pressurized-water ReactorPZR PressurizerRCP Reactor Coolant PumpRCS Reactor Coolant SystemROSA Rig of Safety AssessmentRV Reactor VesselSBLOCA Small Break Loss-of-Coolant AccidentSET Separate Effects TestSG Steam GeneratorSI Safety InjectionTCOLD Cold Leg TemperatureTHOT Hot Leg TemperatureUP Upper PlenumUSNRC United States Nuclear Regulatory Committee

Mitsubishi Heavy Industries, LTD.xiv

6AS-1 E-UAP-100001 (RO)M-RELAP5 Supplementary Manual Vol. III Non-Proprietary Version

1. INTRODUCTION

M-RELAP5 has been developed by Mitsubishi Heavy Industries, Ltd. (MHI) and currentlyused for the small break LOCA analyses for the US-APWR. The code is based upon theRELAP5-3D code with modifications implemented to satisfy the requirements set forth inthe Appendix K to 10 CFR 50 for conservative safety analyses. The base code of RELAP5has a long history and has been widely assessed using experimental data obtained fromvarious Separate Effects Test (SET) and Integral Effects Test (lET) facilities for theapplication to PWR SBLOCAs as a best-estimate code. Taking into account the majormodifications from the original code, the M-RELAP5 is applicable as a conservative codefor PWR's SBLOCA analyses.

The M-RELAP5 has been developed and assessed in conformance to the EvaluationModel Development and Assessment (EMDAP) by the U. S. Nuclear RegulatoryCommission. Important phenomena and processes occurring under the specified accidentand transient are identified first, and then code models, correlations, and capabilitiesrelated to the important phenomena and processes are verified and validated by usingappropriate experimental data obtained from SET and lET facilities scalable to thespecified plant. This supplementary manual describes the M-RELAP5 code assessmentstrategy adopted by MHI, and the code assessment results using the various SET and lETdata, independently from that for the base code, RELAP5-3D, reported in its originalmanual.

The M-RELAP5 code development, assessment, and application to the US-APWR plantanalyses are described in the licensing topical report of MUAP-07013-P 'Small BreakLOCA Methodology for US-APWR'. The code validations using experimental data todemonstrate the code applicability for the US-APWR SBLOCA analyses were performedand the results are documented in the topical report. This supplementary manual presentsall the code validation results using the experimental test data independently performedfrom that for RELAP5-3D developmental assessment. The scalability of each experimentaltest data is examined in the scaling analysis report for US-APWR SBLOCAs.

The code validations using SETs were performed to examine the capability of the code topredict the important phenomena and processes involved during SBLOCA transients inthe US-APWR. The important phenomena and processes are, among others, the voidprofile and mixture level, uncovered-bundle heat transfer, reflood, counter-current flowlimit (CCFL), loop seal, and advanced accumulator operational behavior. TheROSA-IV/LSTF and ORNL/THTF facilities provided experimental data for the void profile,mixture level, and uncovered-bundle heat transfer. Experimental tests for thehigh-pressure reflooding unique to SBLOCAs were also conducted in the ORNL/THTFfacility. In addition, the code validation using the forced reflood test data obtained in theFLECHT-SEASET were performed to demonstrate the code ability under lower pressureranges. The CCFL model was validated using the reflux flooding experimental data fromthe Dukler air-water and UPTF tests. Regarding the advanced accumulator, which is aspecific improved safety feature of the US-APWR, the flow damper model has beendeveloped based on the data obtained from a scaled test conducted by MHI. The modelverification results are presented in this report.

The code capability to predict the plant responses during SBLOCAs were assessed usingthe experimental data obtained from lET facilities. The ROSA-IV/LSTF facility provided

Mitsubishi Heavy Industries, LTD.1-1


well instrumented lET data for the M-RELAP5 assessment with various break sizes. Inaddition, the LOFT/L3-1 and Semiscale/S-LH-1 tests are implemented into the M-RELAP5code assessment matrix. Therefore, M-RELAP5 has been thoroughly assessed usingvarious lET data from the different scaled-test facilities.



2. CODE QUALIFICATION STRATEGY

The present section briefly describes the M-RELAP5 code assessment processes. Asdescribed in the topical report 2-1, the important phenomena and processes were identifiedand the PIRT (Phenomena Identification and Ranking Table) for the US-APWR SBLOCAshas been developed as the initial step. Afterwards, the code assessment matrix wasformulated based on the developed PIRT. The code assessment matrix shall address allthe important phenomena and processes in the PIRT. Since M-RELAP5 has been appliedfor the US-APWR SBLOCA analyses as a conservative Appendix K code, it is necessaryto demonstrate that the code is capable to predict the important phenomena andprocesses conservatively.

2.1 Phenomena Identification and Ranking Table

Prior to developing the Phenomena Identification and Ranking Table (PIRT), theUS-APWR system and constitutive components were examined such that the specifiedaccident can be correctly modeled by using the analysis code. Next, the specifiedSBLOCA transient was divided into several phases to identify the thermal-hydraulicphenomena, behaviors and processes that are dominant in each phase. By implementingthis procedure, the US-APWR SBLOCA PIRT has been developed and established. Thedetails of the US-APWR PIRT are described in the topical report2-1.

Table 2.1-1 shows the US-APWR SBLOCA PIRT, which contains only the high-rankedphenomena and processes. Several phenomena and processes listed in the table are notvalidated using the experimental test data, because they are treated with conservativeassumptions for the plant safety analysis. The decay heat and local power for the fuel rodregion, the three-dimensional power distribution for the core region, and the DVI/SI waterflow rate for the downcomer/lower plenum regions are categorized into theaforementioned. Similarly, the break critical flow and enthalpy are not directly examinedwith SET data, since the spectrum calculations are performed for the break size andorientation (circumferential break location of the pipe) to determine the limiting accidentcase for the plant safety evaluation. Uncertainties contained in the related models can beexcluded via the spectrum calculations. Furthermore, the Moody critical flow model, whichevaluates the break flowrate conservatively, is implemented and used for the plant safetyanalysis in conformance to the requirement of Appendix K to 10 CFR 50.

The code models related to the remaining important phenomena and processes areverified and validated using appropriate experimental test data from various SET and lETfacilities. The present report describes the M-RELAP5 code assessment results using thetest data.



Table 2.1-1 PIRT for US-APWR SBLOCA (High Rank)

High-Ranked Phenomena IPhases in Small Break LOCA

in the main components Natural Loop SealBlowdown Circulation Clearance Boil-Off Recovery

Fuel RodDecay HeatLocal Power

CoreCHF/DryoutUncovered Core Heat TransferRewet (Heat Transfer Recovery)Mixture Level3-D Power Distribution

Steam GeneratorWater Hold-Up in SG Inlet PlenumWater Hold-Up in U-Tube Uphill SidePrimary Side Heat TransferSecondary Side Heat Transfer

Cross-over LegWater Level in SG Outlet PipingLoop Seal Formation and Clearance

Downcomer/Lower PlenumMixture Level/Void DistributionDVI/Sl Water Flowrate

BreakCritical FlowBreak Flow Enthalpy



2.2 Code Assessment Matrix

The important phenomena and processes identified in the PIRT shall be validated usingexperimental test data which are scaled to the specified power plant and accident. Thephenomena and processes to be assessed are listed in Table 2.2-1. As described in thepreceded section, some phenomena and processes are excluded from the assessmentmatrix. The models related to the remaining phenomena and processes are validatedusing SET and lET experimental data as follows:

(1) CHF/DryoutCHF/dryout is likely to occur during the loop seal phase and/or the boil-off phaseunder US-APWR SBLOCAs. Therefore, the related model in M-RELAP5 is validatedusing the steady-state uncovered-bundle heat transfer test data obtained from theORNL/THTF facility 2-2.

(2) Uncovered Core Heat TransferCore uncovery is likely to occur during the loop seal, boil-off and core recovery phaseunder US-APWR SBLOCAs. Along with the CHF/dryout assessment, the relatedmodel in M-RELAP5 is validated using the uncovered-bundle heat transfer test datafrom the ORNL/THTF facility 2 2.

(3) RewetRewetting occurs during the reflooding process following the loop seal clearance andin the core recovery phase. The ORNL/THTF facility provided the high-pressurereflooding test data2- 3 to validate the code models. In addition, M-RELAP5 isvalidated using the test data from the FLECHT-SEASET facility to enhance the

2-4experimental data range to the lower pressure

(4) Core Mixture LevelTwo-phase mixture level swell can be related with the void distribution under the lowcoolant flowrate. The M-RELAP5 void model, primarily the interfacial shear model, isvalidated using the test data obtained in the ROSA-IV/LSTF2- 5 2- 6 and theORNL/THTF2-2 facilities.

(5) Water Hold-Up in SG Inlet PlenumReflux flooding following the initial blowdown in the SG inlet plenum is represented bythe counter-current flow limit (CCFL) model in M-RELAP5. The CCFL model has beendeveloped based on the test data obtained in the full-scale UPTF test data2 7.

(6) Water Hold-Up in SG U-Tube Uphill SideIn M-RELAP5 modeling for US-APWR SBLOCAs, the Wallis CCFL model2-8 isapplied to the flooding in the piping with small diameter such as the SG U-tubes. Themodel is assessed using the Dukler Air-Water test data2 9.

(7) SG Primary and Secondary Heat TransferIn the M-RELAP5 code assessment, the heat transfer model has not been validatedusing the SET data representing the SG primary and secondary sides. Instead, thecode applicability has been assessed using the lET data simulating the PWRSBLOCAs obtained in the ROSA-IV/LSTF test facility2-1°.



(8) Water Level in SG Outlet PipingWater level in the SG outlet piping, namely crossover leg, is important during the loopseal formation and clearance. In M-RELAP5 code assessment, the code predictabilityfor the water level is assessed using lET data from the ROSA-IV/LSTF2 1 °' 211,212LOFT 213214, and Semiscal e2-15 facilities as well as is done for the loop seal formationand clearance. The water retention following the loop seal clearance is validatedusing SET data obtained using the UPTF full-scale test facility.

(9) Loop Seal Formation and ClearanceThe loop seal formation and clearance behavior is assessed using simulatedSBLOCA experimental data obtained in the ROSA-IV/LSTF2 -°' 2-11, 2-12. In addition, thecode is applied to the LOFT/L3-1 2-14 and Semiscale/S-LH-1 2-15 test analyses todemonstrate an enhanced capability of the code ability to predict this complicateddynamic behavior.

(10) Downcomer Mixture Level/Void DistributionSimilar to the loop seal formation and clearance, the downcomer behavior predictedby M-RELAP5 is validated using simulated SBLOCA test data from the various lETfacilities, ROSA-IV/LSTF2 1° 211.2-12, LOFT2 14 , and Semiscale2 -15 .



Table 2.2-1 M-RELAP5 Code Assessment Matrix for US-APWR SBLOCAs

- 9 ! 9 r i ! i ~ C

0

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Separate Effect Tests (SETs)

ROSA-IV/LSTF Void Profile Test X

ORNL/THTF Void Profile Test X

ORNL/THTF Uncovered Heat XTransfer TestORNLiTHTF High-Pressure X XReflood Test

FLECHT-SEASET Forced Reflood Test X X

UPTF SG Plenum CCFL Test X

Dukler Air-Water Flooding Test X

UPTF Test 5 X X

Scaled Advanced Accumulator Test

Integral Effect Tests (lETs)

ROSA-IV/LSTF SB-CL-18 (5% CLB) X X X X X X X X X X

ROSA-IV/LSTF SB-CL-09 (10% CLB) X X X X X X X X X

ROSA-IV/LSTF SB-CL-12 (0.5% CLB) X X X X X X X X X

LOFT L3-1 Test (2.5% CLB) X X X X X

Semiscale S-LH-1 Test (5% CLB) X X X X X X X X X

* Component Effects Test to determine the advanced accumulator flow model.

** Cold Leg Break


.6AS-1 E-UAP-1 00001 (RO)M-RELAP5 Supplementary Manual Vol. III Non-Proprietary Version

2.3 References

2-1 Mitsubishi Heavy Industries, Ltd., 'Small Break LOCA Methodology for US-APWR,'MUAP-07013-P (RO), July 2007.

2-2 T. M. Anklam, R. J. Miller, and M. D. White, 'Experimental Investigations ofUncovered-Bundle Heat Transfer and Two-Phase Mixture Level Swell underHigh-Pressure Low Heat-Flux Conditions,' NUREG-2456, ORNL-5848, March1982.

2-3 C. R. Hyman, T. M. Anklam, and M. D. White, 'Experimental Investigations ofBundle Boiloff and Reflood under High-Pressure Low Heat-Flux Conditions,'NUREG-2455, ORNL-5846, April 1982.

2-4 M. J. Loftus et al., 'PWR FLECHT-SEASET Unblocked Bundle, Forced and GravityReflood Task Data Report,' NUREG/CR-1532, June 1980.

2-5 The ROSA-IV Group, 'ROSA-IV Large Scale Test Facility (LSTF) SystemDescription,' JAERI-M 84-237, 1985.

2-6 Y. Anoda et al. , 'Void fraction distribution in rod bundle under high pressureconditions,' HTD-Vol.155, Am. Soc. Mech. Eng., Winter Annual Meeting, Dallas,Nov. 25-30, 1990.

2-7 P. S. Damerell et al., 'Use of Full-Scale UPTF Data to Evaluate Scaling ofDowncomer (ECC Bypass) and Hot Leg Two-Phase Flow Phenomena,'NUREG/CP-0091 Vol. 4, February 1988.

2-8 G. B. Wallis, 'One Dimensional Two Phase Flow,' McGraw-Hill, 1969.2-9 A. E. Dukler, L. Smith ," Two Phase Interactions in Counter-Current Flow : Studies

of the Flooding Mechanism, Annual Report November 1975 - October 1977,"NUREG/CR-0617, January 1979.

2-10 H. Kumamaru et al., 'ROSA-IV/LSTF 5% Cold Leg Break LOCA Experiment RunSB-CL-18 Data Report,' JAERI-M 89-027, 1989.

2-11 M. Suzuki and H. Nakamura, 'A study on ROSA/LSTF SB-CL-09 Test SimulatingPWR 10% Cold Leg Break LOCA: Loop-Seal Clearing and 3D core Heat-UpPhenomena,' JAEA-Research, October 2008.

2-12 H. Asaka et al., 'Results of 0.5% Cold-Leg Small-Break LOCA Experiments atROSA-IV/LSTF Effect of Break Orientation, Experimental Thermal and FluidScience,' Experimental Thermal and Fluid Science, Vol. 3, pp588-596, November,1990.

2-13 D. L. Reeder, 'LOFT System and Test Description (5.5-ft Nuclear Core 1 LOCEs),'NUREG/CR-0247, TREE-1208, July 1978.

2-14 P. D. Bayless et al., 'Experimental Data Report for LOFT Nuclear Small BreakExperiment L3-1,' NUREG/CR-1145, EGG-2007, January 1980.

2-15 G. G. Loomis, 'Experimental Operating Specifications for Semiscale Mod-2C 5%Small Break Loss-of-Coolant Experiment S-LH-1,' EGG-SEMI-6813, February1985.



3. SEPARATE EFFECTS TESTS

3.1 ROSA-IV/LSTF Void Profile Test

3.1.1 Test Description

3.1.1.1 Test Facility

During a small break LOCA, voiding occurs due to flashing and boiling in the core, and atwo-phase mixture level is formed. Prediction and tracking of the two-phase mixture levelin the core is important for evaluation of peak clad temperature (PCT) through the periodsof loop seal clearance, boil-off and recovery since the mixture level can eventually dropinto the core in these periods and core cooling capability is degraded. A series ofexperiments 3-1 have been performed at the ROSA-IV Large Scale Test Facility (LSTF)3-2,3-3 to measure the void fraction distribution in the simulated reactor core rod bundleunder high-pressure low-flow conditions.

The ROSA-IV/LSTF is a volumetrically-scaled (1:48) full-height model of a Westinghousedesigned 4-loop PWR. The schematic is shown in Figure 3.1-1. The facility includes apressure vessel and two symmetric loops, which consist of steam generators, coolantpumps and loop piping. The pressure vessel contains a full-length (3.66 m) bundlecomposed of 1104 rods (1008 electrically heated and 96 unheated). Table 3.1-1summarizes primary specifications for the core section, and scaling to the US-APWR. Roddiameter and pitch are of typical 17 X 17 fuel assembly. The heated rods are supported atten different elevations by grid spacers. The radial power distribution of the bundle isuniform while the axial power profile is chopped-cosine with a peaking factor 1.495.Locations of differential pressure measurements and spacers are shown with the axialpower profile in Figure 3.1-2.

3.1.1.2 Experimental Results

A series of experiments was performed at the ROSA-IV LSTF to measure the void fractiondistribution in the simulator reactor core rod bundle under high-pressure low-flowconditions. The test cases and conditions are summarized in Table 3.1-2. The tests wereconducted in the pressure range of 1.0 to17.2MPa and the rod-bundle power range of 0.5to 7.2 MW corresponding to the average heat flux range 4.5 to 62kW/m 2. For lowerpressures than 8 MPa and lower powers than 4 MW, the void fraction distributions weremeasured under steady-state reflux condensation conditions. The mixture level was keptconstant at slightly below the hot leg bottom, i.e. 2m above the top of bundle. For thehigher pressures than 8MPa or the higher powers than 4MW, the data were obtained fromthe quasi-steady conditions. In both conditions, the low inlet flow conditions into thebundle were used such that the rod-bundle entirely covered by a two-phase mixture.

The void fraction data was derived from the differential pressures along the rod-bundle,assuming negligible friction and form-loss pressure drop. The bundle-averaged voidfraction was obtained from the overall bundle differential pressure (DP1 in Figure 3.1-2).



3.1.2 Code Validation

3.1.2.1 Analysis Model

Figure 3.1-3 illustrates a schematic of M-RELAP5 noding scheme. Water is suppliedthrough the inlet (cold leg) nozzle of the pressure vessel as a boundary condition. Theflow path regions within the pressure vessel consist of the downcomer, the lower plenum,the core channel, the upper plenum, the upper head and the control rod guide tube.These regions are modeled with hydrodynamic volumes. The exit pressure at the hot legsis also specified as a boundary condition for the modeling of the experiments. Therod-bundle and the metal structures, which contact the above mentioned flow path regions,are represented with heat structure modeling. The noding scheme is essentially similar asthat used in the US-APWR M-RELAP5 small break LOCA plant model. The cells enclosedwith real lines represent hydrodynamic volumes while the one with hatched linesrepresent heat structure segments.

[

I



3.1.2.2 Analysis Results

The following 11 test cases for different three pressures are shown in Reference 3-1.

- ST-VF-01A, ST-VF-01B, ST-VF-01C, ST-VF-01D: 1.0 MPa- ST-NC-01, ST-NC-06E, SB-CL-16L: 7.3 MPa- ST-VF-01E, ST-VF-01F, ST-VF-01G, ST-VF-01H: 15.0 MPa

Among these tests, the tests at 7.3MPa were selected for analyses with M-RELAP5,because the pressure during the loop seal and core uncovery phases is around thispressure so that the void prediction at this pressure is important. In addition, codevalidation were conducted using the 1.OMPa test data, since a lower pressure conditioncan be expected during the core recovery phase with a larger break size for theUS-APWR SBLOCA.

Figure 3.1-4 shows the calculation results for axial void profile of the test ST-NC-06Ecomparing with for each test data. The calculation result of the test ST-NC-06E (7.3 MPa)shows a good agreement with the test data over the full-length. Figure 3.1-5 shows thecalculation results for overall bundle void fraction of the test cases with the pressure7.3MPa and for different bundle powers. The calculation result of the test cases with 7.3MPa shows a good agreement with the test data.

Similarly, Figure 3.1-6 compares the calculated void profile with the measurement for thetest ST-VF-01 D, and the overall bundle void fraction for the test cases with the pressure1.OMPa is shown in Figure 3.1-7. In conjunction with the ORNL/THTF void profile testanalyses, all the void fractions predicted by M-RELAP5 are compared with themeasurements in Figure 3.1-8,

It is found that [

1. However, themixture-level prediction is dependent on code accuracy not only for the void fraction itself,but also for the transition void fraction from bubbly/churn to mist flow regimes. Fromanalyses results for ORNL/THTF level swell tests with the various pressure conditionswhich are described in Section 3.2, M-RELAP5 has no significant dependency onpressure in terms of predicting the core mixture-level. Therefore, it is concluded thatM-RELAP5 is capable of predicting the mixture-level accurately.

3.1.3 Summary

The ROSA/LSTF void profile tests for the rod-bundle region were simulated usingM-RELAP5. The calculation results for the pressure 7.3 MPa test cases show a goodagreement with the measurement for both the axial void fraction profile and the averagedvoid fraction. With respect to the lower pressure condition, M-RELAP5 tends to provide alarger void fraction than the measurement. The effects on the mixture-level swell, however,are not significant as demonstrated in the code verification using the ORNL/THTF mixturelevel swell test data conducted under the various pressure conditions.



Table 3.1-1 Primary Core Section Specifications for ROSA-IV

US-APWR/Item ROSA-IV US-APWR ROSA-IV

No. of Assemblies per Core 24 257 -

Rod Array per Assembly 7x7 17x17 -

Total No. of Rods per Core 1168 74273 63.59

No. of Heated Rods per Core 1064 67848 63.77

No. of Unheated Rods per Core 104 6425 61.78

Heated-to-Unheated Rod No. Ratio 10.23 10.56 1.03

No. of Grid Spacers 9

Active Length (m) 3.66

Heated Rod Diameter (m) 0.0095 0.0095 1.00

Unheated Rod Diameter (m) 0.0122 0.0097 0.79

Rod Pitch (m) 0.0126 0.0126 1.00

Flow Area per Assembly (M2) 0.0982 [Hydraulic Diameter (m) 0.0110

Table 3.1-2 ROSA-IV/LSTF Void Profile Test Conditions

Test No. Power Heat Flux Pressure Exit Velocity Jg

(MW) (kW/m 2) (MPa) (m/s)

ST-VF-01A* 0.5 4.5 1.0 0.425ST-VF-01 B* 1.0 9.1 1.0 0.851ST-VF-01C* 2.0 148.2 1.0 1.702

ST-VF-01 D* 3.5 31.8 1.0 2.978ST-NC-08E 1.426 13.0 2.4 0.566

ST-NC-01* 3.57 30.7 7.3 0.553

ST-NC-06E* 3.95 34.0 7.3 0.612

SB-CL-16L* 5.0 43.0 7.3 0.774

ST-SG-04 7.17 61.7 7.35 1.104ST-VF-01 E 1.0 9.1 15.0 0.091

ST-VF-01 F 0.5 4.5 15.0 0.045ST-VF-01G 2.0 18.2 15.0 0.182ST-VF-01 H 4.0 36.3 15.0 0.363TR-LF-03 0.94 7.2 17.2 0.080

* Test selected for M-RELAP5 assessment.



Loop A Loop Ba I I I

Accumulator .

Flow Control ValvePressure Vessel

Coolant Pump

Figure 3.1-1 ROSA-IV/LSTF Test Facility3 2

4.0 Differential Power ProfilePressure

P4

S3.0 ._

0 .

Sace

oE 2.0 u D•

"0 1.0 1 2

0.0 *,Spacer

0.0 0.5 1.0 1.5 2.0

Power RatioFigure 3.1-2 Axial Power Profile and Location of Pressure lnstrumentations 3 1



Figure 3.1-3 M-RELAP5 Noding Scheme for ROSA-IV/LSTF Void Profile Analysis



0.8

. 0.6

LL

o 0.4

0.2

0

P=7.3MPa S ST-NC-06E DataQ=3.95MW - M-RELAP5

0 2 3 4

Elevation (m)

Figure 3.1-4 Simulated Void Fraction Profile for Test ST-NC-06E

P=7.3MPa ' LSTF DataM-RELAP5

0.8

C0

0.6

U.-o0

0.4Cm

L0.2

0

00 I 2 3 4 5 6

Bundle Power (MW)

Figure 3.1-5 Simulated Averaged Void Fraction Profile for 7.3 MPa Tests



Figure 3.1-6 Simulated Void Fraction Profile for Test ST-VF-01D

Figure 3.1-7 Simulated Averaged Void Fraction Profile for 1.0 MPa Tests



Figure 3.1-8 Comparison of Predicted and Measured Void Fractions for ROSA/LSTFand ORNLITHTF Void Profile Tests



3.2 ORNL/THTF Void Profile and Uncovered-Bundle Heat Transfer Tests



The THTF3-4 is a high-pressure-bundle thermal-hydraulics test loop. System configurationwas designed to produce thermal-hydraulic conditions similar to those expected in aSBLOCA. It contained a 64-rod electrically heated bundle with identical dimensions typicalof the 17 x 17 PWR fuel assembly.

Figure 3.2-1 is an illustration of the THTF for a small-break test configuration. Flow leavesthe main coolant pump and passes directly into the lower plenum. Flow proceeds upwardthrough the heated bundle and exits through the bundle outlet spool piece. Spool piecemeasurements include pressure, temperature, density, volumetric flow, and momentumflux. On leaving the orifice manifold, flow passes through a heat exchanger and returns tothe pump inlet. System pressure is controlled via the loop pressurizer.

The THTF test section contains a 64-rod electrically heated bundle. Figure 3.2-2 is across section of the test section. The test bundle is placed in the shroud box. Roddiameter and pitch are typical of a 17 x 17 fuel assembly. The four unheated rods aredesigned to represent control-rod guide tubes in a nuclear fuel assembly. Figure 3.2-3 isa cross section of a typical FRS (Fuel Rod Simulator). The FRS has stainless steelcladding and an Inconel heating element and the FRS is filled with boron nitride as a hightemperature insulating material.

Figure 3.2-4 is an axial profile of the THTF bundle that illustrates the positions of spacergrids and FRS thermocouples. The heated length is 3.66 m (12 ft), and a total of 25 FRSthermocouple levels are distributed over that length. The upper third of the bundle is moreheavily instrumented than the lower portion, since for most tests the two-phase mixturelevel was in the top 1/3 of the heated length. In addition to the FRS thermocouples, fluidtemperatures are measured at a number of locations. Two-phase mixture level and voidfraction profile were obtained through the use of thermocouple and differential pressurecell measurements. Figure 3.2-5 illustrates the differential pressure measurementlocations.

Table 3.2-1 summarizes the primary specifications for the ORNL/THTF test section andscaling to the US-APWR.


The two-phase mixture level swell tests 3-5 or the uncovered-bundle heat transfer testswere started by boiling off water from the bundle, which was originally filled with water.Excess volume was accumulated in the pressurizer, and nitrogen was vented from thepressurizer to maintain constant pressure. Eventually, the THTF settled into aquasi-steady state with the bundle partially uncovered and inlet flow just sufficient to makeup for the liquid being vaporized. Measurements were made at this steady state condition.The bundle power was adjusted to produce peak FRS temperatures of about 1033 K,imposed by safety limits.



The test conditions of the two-phase mixture level swell tests are listed in Table 3.2-2. Thetest bundle was uncovered for the first six tests 3.09.101-N. Three experiments were run atroughly 4MPa, and three experiments at roughly 7MPa. The three experiments at eachpressure level were designed to span a range of linear heat powers. The two-phasemixture level was not established in the bundle and the bundle was covered withtwo-phase water for the remaining six tests 3.09.10AA-FF. The pressure conditions ofthese tests were same as the first six tests.

In SBLOCA transients, the first core uncovery during the loop seal period is expected tooccur when the RCS pressure is relatively high, and the second core uncovery during theboil-off period is expected to occur when the RCS pressure is relatively low, which isabout the accumulator pressure. The THTF two-phase mixture level swell tests cover theexpected range of pressure conditions in the US-APWR SBLOCAs.

The uncovered-bundle heat transfer tests were conducted at the same time under thesame condition as the two-phase mixture level tests for the six uncovered-bundle tests3.09.101 to N.



Figure 3.2-6 shows the M-RELAP5 noding diagram for the ORNL/THTF. [

Of the uncovered-bundle tests 3.09.101 to N, tests 3.09.101 and L are not adopted asvalidation tests because they have a higher liner power/rod compared with that of theUS-APWR SBLOCA transient. For cases with low linear power, the heat loss to theenvironment from the rod bundle and housing was significant and could affect theexperimental results. [




Comparison of void fraction, FRS surface temperature, vapor temperature and heattransfer coefficient between measurement and M-RELAP5 calculation for tests 3.09.10J,K, M and N is presented in Figure 3.2-7 to Figure 3.2-22. The void profile calculatedresults are compared with the measurement in Figure 3.2-23 to Figure 3.2-28.

As indicated in the previous section, calculated void profiles reasonably agree withmeasured values. The post-CHF FRS surface heat transfer coefficient and temperaturealso, in general, reasonably agree with measured values and show slightly conservativeresults for tests 3.09.1OJ and M. The vapor temperature in the experiment was calculatedfrom an energy balance and the measured bundle exit steam temperature. The predictedsteam temperature by the M-RELAP5 code reasonably agrees with the experimental data.This means that the rod heat transfer model by vapor convection incorporated in theM-RELAP5 is adequate to predict SBLOCA core behaviors.

There are dips in rod surface temperature and leaps in heat transfer coefficientdownstream of a grid spacer for the experiments. M-RELAP5 has no mechanism toincrease the heat transfer coefficients downstream of grids 3-6 , and the calculated rodsurface temperatures show no dips.

A small dip is observed in several cases for the calculated void profile. As described inreference 3-7, this inversion occurs when the flow regime changes from bubbly/slug flowto mist flow and thereby interfacial drag coefficient becomes small. Although thisphenomenon stems from a short cell length and an increase in the vapor velocity, thecalculated rod surface temperatures show no dip and reasonably reproduce theexperimental results.

The two-phase mixture levels defined in Reference 3-5 were identified by observing theaverage temperature at the FRS thermocouple levels and were assumed to be midwaybetween the highest level where the average temperature indicated nucleate boiling andthe lowest level where the average temperature indicated CHF. The two-phase mixturelevel was defined from the experimental data such that nucleate boiling is maintained androd surface temperatures are close to the saturation temperature below this level andCHF occurs then the temperature excursion occurs above this level. The FRS surfacetemperature distributions predicted by the M-RELAP5 code match well this description.

The accurate prediction for the collapsed and two-phase mixture levels is also essentialfor a SBLOCA analysis. The comparisons of the predicted and measured collapsed andmixture levels are made in Figure 3.2-29 and Figure 3.2-30, respectively. It is noted fromthese comparisons that the M-RELAP5 code reasonably predicts the mixture level swellbeyond which the rod temperature start to increase.

3.2.3 Summary

The accurate prediction of the two-phase mixture level is important to predict the PCT in a



SBLOCA. The M-RELAP5 code was assessed by the comparison with the ORNLITHTFthe two-phase mixture level swell test and the uncovered-bundle heat transfer test. Theassessment showed that the M-RELAP5 code reasonably predicts these parameters.

The accurate prediction of the rod heat transfer above the two-phase mixture level is alsoimportant to predict the PCT in a SBLOCA. The M-RELAP5 code was assessed by thecomparison with the ORNL/THTF uncover-bundle heat transfer test. The assessmentshowed that the M-RELAP5 code reasonably predicts the rod heat transfer above thetwo-phase mixture level.

Table 3.2-1 Primary Test Section Specifications for ORNL/THTF

US-APWR/

Item THTF US-APWR THTFTHTF


Rod Array per Assembly 8x8 17x17 -

Total No. of Rods per Assembly 64 289 4.52

No. of Heated Rods per Assembly 60 264 4.40

No. of Unheated Rods per Assembly 4 25 6.25

Heated-to-Unheated Rod No. Ratio 15 10.56 0.70





Rod Pitch (m) 0.0127 0.0126 0.99

Flow Area per Assembly (M 2) 0.0062

Hydraulic Diameter (m) 0.0106



Table 3.2-2 ORNL/THTF Void Profile and Uncovered-Bundle Heat Transfer TestConditions

Inlet temperature Liner heat Fractional

Test (Subcooling) power heat loss

(MPa) (kg/s. m 2) (K) (kW/m)

3.09.101 4.50 29.76 473.0 (57.6) 2.22 0.018

3.09. 1OJ* 4.20 12.93 480.3 (46.1) 1.07 0.052

3.09.10K* 4.01 2.22 466.5 (57.2) 0.32 0.176

3.09.1OL 7.52 29.11 461.3 (102.6) 2.17 0.017

3.09. 1OM* 6.96 13.38 474.4 (84.2) 1.02 0.042

3.09.10N* 7.08 4.33 473.1 (86.7) 0.47 0.162

3.09. 1OAA* 4.04 21.15 450.9 (73.2) 1.27 0.020

3.09.1OBB* 3.86 9.44 458.2 (63.2) 0.64 0.034

3.09.10CC* 3.59 7.22 467.6 (49.6) 0.33 0.035

3.09.1ODD* 8.09 19.82 453.4 (115.5) 1.29 0.030

3.09.10EE* 7.71 11.00 455.9 (109.7) 0.64 0.039

3.09.1OFF* 7.53 4.83 451.4 (112.6) 0.32 0.092

* Test selected for M-RELAP5 assessment.



ORNK -DWG 81-7837R ETD

NITROGEN OVERPRESSURE

vo FILL LINE

_, :ýVENT LINE

Figure 3.2-1 THTF in Small-Break Test Configuration 3 5



ORNL-DWG 82-4875 ETD

SHROUD PLENUM ANNULUS(OLD DOWNCOMER)

TEST SECTION BARREL

Figure 3.2-2 Cross Section of THTF Test Section3 5



ORNL DiVG 79 4737 ETD

3i6 STAINLESS

STEEL SHEATH

INCONEL 600HEATING ELEMENT

0.05 cm TC

Figure 3.2-3 Cross Section of Typical Fuel Rod Simulator3 -5

BORONNITRIDE



osNt-OWG 81-20288 ETD

SPACR GRID T/CDES IG NATION

ROD T/CLEVEL LEVEL OlC.)

144-

GF8•

F6F5 -_

TE296X -F4 -

F3 __F2--

F1 --5

E6 ......

TE295X .. ..

K ~

-... 142- 141___ _1 138

- 136'34

-_ 139

_Q26

119- I T4

-- i112

3/4POWER

oISTRIBUTJON

I,,

€zUJ

ti

.- 108

E3 --

E2--Ef- - -.

---- 102$02.... .99

-- 95

0

A.

--. 64-- 62

....... 60TE293K

S --

U -56

-36

- f2

0-

Figure 3.2-4 Axial Location of Spacer Grids and FRS Thermocouples 35



ORNL DW(.81 20230 I-TD

*MEASUREMENTS REFERENCED TOGASKET FACE ON LOWER FLANGEOF TEST SECTION 6ARREL

Figure 3.2-5 THTF In-Bundle Pressure Instrumentation 3 5



9Figure 3.2-6 M-RELAP5 Noding Scheme for ORNL/THTF Test Analysis



-K- M-RELAP5

-0- Measured

1.0

0.8

00

CU'4-0

0.6

0.4

0.2

0.0

0 1 2 3 4

Elevation (m)

Figure 3.2-7 Comparison of Predicted and Measured Void Fraction Profiles forORNUTHTF Test 3.09.10J


3-21


--0- M-RELAP5

-0- Measured

1300

E

0,

0ý

('3

(I)V

0r

1100

900

700

500

0 1 2 3 4

Elevation (m)

Figure 3.2-8 Comparison of Predicted and Measured FRS Surface TemperatureProfiles for ORNL/THTF Test 3.09.10J



--0- M-RELAP5

-0- Measured

1300

7

OL

E

0('

1100

900

700

500

0 1 2 3 4

Elevation (m)

Figure 3.2-9 Comparison of Predicted and Measured Vapor Temperature Profilesfor ORNL/THTF Test 3.09.10J



500

0CEa)

00

"4-

400

300

200

100

-K-- M-RELAP5

-0- Measured

I I ! I I I I I I I I I I I I I

0

0 1 2 3 4

Elevation (m)

Figure 3.2-10 Comparison of Predicted and Measured Heat Transfer CoefficientProfiles for ORNLITHTF Test 3.09.10J


6AS-1 E-UAP-100001 (RO)Non-Proprietary VersionM-RELAP5 Supplementary Manual Vol. III

1.0

0.8

!0

4-

0.6

0.4

0.2

0.0

0 1 2 3 4

Elevation (m)

Figure 3.2-11 Comparison of Predicted and Measured Void Fraction Profiles forORNL/THTF Test 3.09.10K



--0- M-RELAP5

-0- Measured

1300

7

a)

E

a)

-4-0)_00

0ý

1100

900

700

500

0 1 2 3 4

Elevation (m)

Figure 3.2-12 Comparison of Predicted and Measured FRS Surface TemperatureProfiles for ORNLITHTF Test 3.09.10K



-K0- M-RELAP5

-0- Measured

1300

7

a)

L..

0..

Ea),4-0

0.0

1100

900

700

500

0 1 2 3 4

Elevation (m)

Figure 3.2-13 Comparison of Predicted and Measured Vapor Temperature Profilesfor ORNL/THTF Test 3.09.10K



-K-- M-RELAP5

-0- Measured

500

a)

2c,4

a-U)

00L..

4-

CU

a)r

400

300

200

100

0

0 1 2 3 4

Elevation (m)

Figure 3.2-14 Comparison of Predicted and Measured Heat Transfer CoefficientProfiles for ORNLITHTF Test 3.09.10K



-0- M-RELAP5

-0- Measured

1.0

0.8

0

CUI4

4-O

01

0.6

0.4

0.2

0.0

0 1 2 3 4

Elevation (m)

Figure 3.2-15 Comparison of Predicted and Measured Void Fraction Profiles forORNL/THTF Test 3.09.10M



-0-- M-RELAP5

-0- Measured

1300

7

1)

a)0--

EL)

'1:)CU,tIC,)

0

1100

900

700

500

0 1 2 3 4

Elevation (m)

Figure 3.2-16 Comparison of Predicted and Measured FRS Surface TemperatureProfiles for ORNLITHTF Test 3.09.10M



-0- M-RELAP5

-0- Measured

1300

7

a)

L.4a)

O-

E

0OCCU_

1100

900

700

500

0 1 2 3 4

Elevation (m)

Figure 3.2-17 Comparison of Predicted and Measured Vapor Temperature Profilesfor ORNL/THTF Test 3.09.1OM



500

2

4--

4

a)0

4-

CU

0

C-U

a)

4-

400

300

200

100

--0- M-RELAP5

-0- Measured

F I P [ I I ] I [ P I I I

0

0 1 2 3 4

Elevation (m)

Figure 3.2-18 Comparison of Predicted and Measured Heat Transfer CoefficientProfiles for ORNLITHTF Test 3.09.10M



-0-- M-RELAP5

-0- Measured

1.0

0.8

0.40

L..4--

O0

0.6

0.4

0.2

0.0

0 1 2 3 4

Elevation (m)

Figure 3.2-19 Comparison of Predicted and Measured Void Fraction Profiles forORNL/THTF Test 3.09.1ON



--0- M-RELAP5

-0- Measured

1300

7

CU

L,-

L..

Ea)

CU

_0

o00"

1100

900

700

500

0 1 2 3 4

Elevation (m)

Figure 3.2-20 Comparison of Predicted and Measured FRS Surface TemperatureProfiles for ORNL/THTF Test 3.09.1ON



-K>- M-RELAP5

-0- Measured

1300

EL.

0

1100

900

700

500

0 1 2 3 4

Elevation (m)

Figure 3.2-21 Comparison of Predicted and Measured Vapor Temperature Profilesfor ORNLITHTF Test 3.09.1ON



--0- M-RELAP5

-0- Measured

500

E

C-

a)000)

CU)

M

a)Ir

400

300

200

100

0

0 1 2 3 4

Elevation (m)

Figure 3.2-22 Comparison of Predicted and Measured Heat Transfer CoefficientProfiles for ORNLITHTF Test 3.09.10N



-0- M-RELAP5

-0- Measured

1.0

0.8

0

4-

0.6

0.4

0.2

0.0

0 1 2 3 4

Elevation (m)

Figure 3.2-23 Comparison of Predicted and Measured Void Fraction Profiles forORNL/THTF Test 3.09.10AA



1.0

0.8

040I-

4--

L0

0.6

0.4

0.2

0.00 1 2 3 4

Elevation (m)

Figure 3.2-24 Comparison of Predicted and Measured Void Fraction Profiles forORNL/THTF Test 3.09.1OBB



-0- M-RELAP5

-0- Measured

1.0

0.8

0.6

0.4

0C

4-.5

0.2

0.0

0 1 2 3 4

Elevation (m)

Figure 3.2-25 Comparison of Predicted and Measured Void Fraction Profiles forORNL/THTF Test 3.09.10CC



-0- M-RELAP5

-0- Measured

1.0

0.8

0L.)

'4-

O0

0.6

0.4

0.2

0.0

0 1 2 3 4

Elevation (m)

Figure 3.2-26 Comparison of Predicted and Measured Void Fraction Profiles forORNL/THTF Test 3.09.1ODD



--C- M-RELAP5

-0- Measured

1.0

0.8

00

4-a

4-

V.5

0.6

0.4

0.2

0.0

0 1 2 3 4

Elevation (m)

Figure 3.2-27 Comparison of Predicted and Measured Void Fraction Profiles forORNLITHTF Test 3.09.10EE



1.0

0.8

0

04-O.5

0.6

0.4

0.2

0.0

0 1 2 3 4

Elevation (m)

Figure 3.2-28 Comparison of Predicted and Measured Void Fraction Profiles forORNUTHTF Test 3.09.10FF



E

-Ja)

_0a)

CU

0

0~

"0

a)

4.0

3.5

3.0

2.5

2.0

1.5

1.0

0.5

0.0

0.0 0.5 1.0 1.5 2.0 2.5 3.0 3.5 4.0Measured Collapsed Level (m)

Figure 3.2-29 Comparison of Predicted and Measure Collapsed Liquid Levels forORNL/THTF Tests


M-RELAP5 Supplementary Manual Vol. III6AS-1 E-UAP-1 00001 (RO)


E

_)-JU!)

X

04

4.0

3.5

3.0

2.5

2.0

1.5

1.0

0.5

0.0

0.0 0.5 1.0 1.5 2.0 2.5 3.0 3.5 4.0Measured Mixture Level (m)

Figure 3.2-30 Comparison of Predicted and MeasureTests

Mixture Levels for ORNL/THTF



3.3 ORNLITHTF High-Pressure Reflood Test



The high-pressure reflood test3-8 was conducted using the same test facility and testsection as the uncovered-bundle heat transfer and two-phase mixture level swell test,which are already described in Section 3.2.1.1.


Initial conditions for the high-pressure reflood test were established in a manner identicalto that used in the uncovered bundle heat transfer and two-phase mixture swell test.Reflood was initiated from a configuration in which the bundle was partially uncovered andpeak cladding temperature was on the order of 1033K (1400F). Flow-power matching wassuch that 22 to 34% of the bundle heated length was initially uncovered.

To initiate reflood, the inlet flooding valve was opened to a predetermined setting. Thiscaused the test section inlet flow to increase, thus commencing bundle recovery. Bundlepower remained constant until completion of core recovery. Data were also taken untilcore recovery was complete.

The test conditions of the high-pressure reflood test are shown in Table 3.3-1. Parametricvariations include pressure, flooding rate, and linear heat rate. The initial system pressureranged from 3.9 to 7.5MPa. Average inlet flooding velocities ranged from 5.9 to 12.2cm/s.Linear heat rate ranged from 1.00 to 2.16kW/m. These test conditions cover the possiblerange of conditions for the reflood recovery during SBLOCAs. Among these tests,3.09.1OP and Q were used as validation tests because they have suitable liner power/rodfor the US-APWR SBLOCA transient.



Figure 3.3-1 shows the M-RELAP5 noding diagram for the ORNL/THTF high-pressurereflood test. Since a simulation of the high-pressure reflood tests is a transient calculation,the fuel rod simulator (FRS) is divided into a number of nodes in the radial direction, andreliable values of the thermal properties are produced from Reference 3-7 and ASMEPhysical Properties Tables (2001) for stainless steel cladding, inconel heating element,and filled boron nitride. The transients of reflood rate, inlet temperature, and pressurewere supplied as boundary conditions, which are presented in Figure 3.3-2 throughFigure 3.3-7. Theses boundary conditions were imposed by time-dependent volumecomponents and a time-dependent junction component identical to those of theuncovered-bundle heat transfer and two-phase mixture level swell test.

There are no data reported regarding heat loss in Reference 3-8. The effect of heat losswas considered insignificant and therefore not modeled in the simulation becauseexpected heat loss from a two-phase mixture-level swell test 3.09.1OJ, performed under a



similar pressure and power condition to 3.09.1OP and Q, is small, about 5%.

Prior to the initiation of reflood analysis, adequate agreement of the initial FRS surfacetemperature was established between experimental and analysis values. To do this, initialvalues of inlet flow and subcooling were adjusted in the steady state calculation prior tothe transient calculation by using values slightly shifted from those reported in Reference3-8 such that the initial conditions for the simulation best matched the experimental data.In this fashion, the uncertainty in the M-RELAP5 model initial conditions was minimized soa more accurate assessment of the M-RELAP5 high pressure reflood behavior andmodels could be assessed.

The reflood test calculation was started at a condition where the test bundle partiallyuncovered by imposing the boundary conditions shown in Figure 3.3-2 through. Figure3.3-7.


Figure 3.3-8 through Figure 3.3-15 show comparison of FRS surface temperature, fluidtemperature, collapsed liquid level, and quench level between the M-RELAP5 calculatedresults and the experimental data for tests 3.09.1OP and Q.

The collapsed liquid levels are presented in Figure 3.3-10 and Figure 3.3-14. The rate ofrise in collapsed levels is large in the early time and becomes gradually small at a latertime. This is because the axial power distributions for the tests are flat over the heatedlength; thus in the beginning of reflooding the FRS superheat is small near the two-phasemixture level and grows larger with distance from the mixture level. Consequently initialstored energy in the FRS is small near the mixture level and increases as the distancefrom the mixture level increases. As a result, the evaporation rate is small in the beginningof the reflooding and larger at later time when more stored energy is removed byquenching of the FRS. Since the inlet flow for test 3.09.10 Q (Figure 3.3-5) is smaller thantest 3.09.10 P (Figure 3.3-2), the rate of rise in collapsed level for test 3.09.10 Q is smallerthan that for test 3.09.10 P. M-RELAP5 predicts this tendency well.

Figures 8.1.3-10 and 14 indicate good agreement between the calculated results and theexperimental data. Although the oscillation of the collapsed liquid level for test 3.09.10 Qresults from rewetting at elevations representing discretized bundle volumes, the averagebehavior of the collapsed liquid level seems well simulated.

The variation of the collapsed liquid level with time relates to vapor generation under thetwo-phase mixture level and entrainment from the liquid-vapor interface. The comparisonsindicate that M-RELAP5 adequately simulates these phenomena.

The FRS surface temperatures at levels F and G (see Figure 3.3-4) for both tests arepresented in Figure 3.3-8, Figure 3.3-9, Figure 3.3-12 and Figure 3.3-13. In theexperiments the FRS surface temperature decreased gradually before the quenchoccurred and dropped to the saturation temperature in a short time once the quenchconditions were met. M-RELAP5 predicted this precursory cooling together with changesin heat transfer mode from single-phase vapor convection to saturated film boiling. In thelater time, however, the M-RELAP5-predicted FRS surface temperature did not showclear-cut quench but gradually decreased and finally reached the saturation temperaturewhen the superheat dropped below 100K and thereby the heat transfer mode changed



from transition boiling to nucleate boiling. This conservative evaluation showing a delay inquench time, resulted from the facts that the M-RELAP5 predicted heat transfer for filmboiling and transition boiling, which are dominant in the pre-quench cooling isconservatively modeled and that axial heat conduction in the heated rod surface at thequench front, also dominant effect in the quench, is not considered in the simulation.

Figure 3.3-11 and Figure 3.3-15 show comparisons of the quench level variation with timebetween experimental data and M-RELAP5 result for tests 3.09.10 P and Q, respectively.The quench times resulted from the M-RELAP5 calculation were identified by heattransfer mode change from transition boiling to nucleate boiling that occurred just beforethe FRS surface temperature reduced to the saturation temperature. These figuresindicate that the M-RELAP5 calculated quench velocities, which can be obtained bydifferentiating the quench level with respect to time, are much smaller than theexperimental results.

It is noted from the above that the M-RELAP5 heat transfer model for the reflooding phasecalculates a longer quench time than the experiment, and thus evaluates the FRS surfacetemperature conservatively.

3.3.3 Summary

M-RELAP5's reflood modeling was assessed against the ORNLFTHTF high-pressurereflood tests. It was concluded that M-RELAP5 adequately predicts fluid conditions suchas fluid collapsed level during reflood and higher FRS surface temperature, and thusM-RELAP5 conservatively. predicts the rod heat transfer behavior during reflood. Inconclusion, it is reasonable to apply M-RELAP5 to simulation of reflooding phase ofUS-APWR SBLOCA EM analysis.

Table 3.3-1 ORNL/THTF High-Pressure Reflood Test Conditions

Initial Initial inlet Initial inlet Linear FloodingTest No. Pressuremass flux temperature subcooling heat power velocity

(MPa) (kg/m 2s) (K) (K) (kW/m) (cm/s)

3.09.100 3.88 25.36 447.7 74 2.03 12.2

3.09.10P* 4.28 12.19 462.6 65 0.997 9.2

3.09.1OQ* 3.95 12.68 456.8 66 1.02 5.9

3.09. 1OR 7.34 27.64 449.2 113 2.16 11.7

3.09.10S 7.53 13.82 459.0 105 1.38 10.2* Test selected for M-RELAP5 assessment.



\IFigure 3.3-1 M-RELAP5 Noding Scheme for ORNL/THTF Reflood Test Analysis




.9

-j

20

15

10

5

0

-5

{X 10-4)

12

8

E

0_JU_

0

-10 -5 0 5 10 15 20 25 30 35 40

TIME (s)

Figure 3.3-2 Test Section Inlet Flows for ORNL/THTF Test 3.09.10P Test

ORNL-DWG 81.20343 ETD

U-0

I-

LU

w

550

500

450

400

350

300

250

200

150

550

So0

LUcc

450 2

400 w

350

100 ---10 -5 0 5 10 15 20 25 30 35 40

TIME (s)

Figure 3.3-3 Test Section Inlet Temperatures for ORNL/THTF Test 3.09.10P Test





1'1 •, 5200

Ui

LU

0~

UJU-

760

740

720

700

680

660

640

620

5000

4800 w

U,

4600 a

4400

-10 -5 0 5 10 15 20 25 30 35 40

TIME (WIFigure 3.3-4 Test Section Pressures for ORNLITHTF Test 3.09.10P Test


0 M-RELAP5 (X 10-4)1_.o FE - 3, f ,ikMR LA5 • .... 4 4

3

E

U..

0-10 -5 0 5 10 15 20 25 30 35 40 45 50

TIME (s)

Figure 3.3-5 Test Section Inlet Flows for ORNL/THTF Test 3.09.10Q Test



ORNL-DWG BI-20344 ETD

500-

480 j

I.

I ITsa

D TE-256 [A rn

LL

440-I-

420

LuS400.

(Figure 3.2-1) TOP OF BUNDLE

HEATED LENGTHQUENCHED

REFLOOD

520

w

,2500 a,-

I--

a:LU

480 Lu4-

460

INITIATED TEST SECTIONINLET -M-RELAP5TEMPERATURE -- Measured

380.

360

-10 -5 0 5 10 15 20 25 30 35 40 45 50

TIME (s)Figure 3.3-6 Test Section Inlet Temperatures for ORNL/THTF Test 3.09.10Q Test

ORNL-DWG 81.20364 ETD

0.

U.I

ILn

Enwa:CL

660

650

640

630

620

610

600

590

580

570

4500

4375

a.

4250V)Vl)

4125

4000

-10 -5 0 5 10 15 20 25 30 35 40 45 50

TIME (s)

Figure 3.3-7 Test Section Pressures for ORNLITHTF Test 3.09.10Q Test



- M-RELAP5 FRS (304.80 cm)- M-RELAP5 FLUID (304.80 cm) ORNL-DWG 81-20394 ETD

1300

1200

1100

1000wD• 900

i-L 800

700

900

800

LU700 '•

i--

600

500

600

500

400-10 -5 0 .5 10 15 20 25 30 35 40

TIME (s) Level F = 302.138 (cm)

Figure 3.3-8 FRS and Fluid Temperatures at LevelTest

F for ORNL/THTF Test 3.09.10P




L1.

cc

cr

I-

I-

1500

1400

1300

1200

1100

1000

900

800

700

600

500

1000

900 2w

800 <

W

700

600

-10 -5 0 5. 10 15 20 - 25 30 35 40

TIME (s) Level G = 362.131 (cm)

Figure 3.3-9 FRS Temperatures at Level G for ORNLITHTF Test 3.09.1OP Test



1-

0coLu

0co

-jLUU)

-J

0

0

C3LUU)10_

0_.

From ORNL-DWG 81-20399 ETD

160o IETOP OF BUNDLE E

150 - M-RELAP5 HEATED LENGTH 385.08 .j- Measured QUENCHED I I

0140 co

-. 346.30 uJ

130 0nto

120 307.52 <w

110 LU268.74 -J

100 REFLOOD 0INITIATED

90. 229.96

LOWEST UNCOVERED FRS LU.80 LEVEL (E6) QUENCHED u17o I I 191.18 o,

60 ' 1 152.40

-20 -71.0 10 20 30 40 50

TIME (s)

Figure 3.3-10 Collapsed Liquid Level for ORNL/THTF Test 3.09.10P Test



From ORNL-DWG 81-20399 ETD

rI:0M

0

LU

wO

-J

_j,_1

IC-,

z

0

160

150

140

130

120

110

100.

90

.80

70

385.08E

346.30 "J

0

307.52 W

0

268.74 .JLiJ

-J

229.96C-)z

191.18.~

152.4060.-20 -1.0 .0 10 20 30 40 50

TIME (s)

*1 The time when the heat transfer mode switches from film boiling to transition boiling.*2 The time when the heat transfer mode switches from transition boiling to nucleate boiling.

Figure 3.3-11 Quench Level for ORNL/THTF Test 3.09.10P Test



- M-RELAP5 FRS (304.80 cm)- M-RELAP5 FLUID (304.80 cm)

1200

1100

1000

900LU

I- 800

LUa 700

6U

600

O RN L-DWG 81-20395 ETD

TE-188AF -900TE-188BFTE--320AFTE-31 8CFTE-314AF - 800__ ^1.80O

9LU

700 F"

LUCL

LU

800

500

500

400

-10 -5 0 5 10 15' 20 25 30 351 40 45 50

TIME (s) Level F = 302.138 (cm)

Figure 3.3-12 FRS and Fluid Temperatures at Level F for ORNLITHTF Test 3.09.10QTest




o.TE-3..G

uL.

U.,

I-

Cr

I--

1400

1300

1200

1100

1000

900

800

700

600

500

uJCr

F-

Cr

w

1-

-10 -5 0 5 .10 15 20 25 30 35 40 45 50

TIME (s) Level G = 362.131 (cm)

Figure 3.3-13 FRS Temperatures at Level G for ORNLITHTF Test 3.09.10Q Test


M-RELAP5 Supplementary Manual Vol. III6AS-1 E-UAP-1 00001 (R0)


T-o0m

0m

uJw

LU

aJ

0

0

Co

.- I

07w

From ORNL-DWG 81.20400 ETD

160 i -E

150 - M-RELAP5 I385.08 .

- Measured "140 0

14.346.30 L

130 >

120 REFLOOD -307.52 <

1o INITATED Lu2874 "'iLu

-268.74 .- 1

100 0P OF BUNDLE

90 HEATED LENGTH 229.96.QUENCHED--._

80 LULOWEST UNCOVERED FRS 191.18

70 LEVEL (E6),b UENCH ED

60 . . . . . 152.40 0

-20 - 10 0 10 20 30 40 50 60

TIME'(s)

Figure 3.3-14 Collapsed Liquid Level for ORNL/THTF Test 3.09.10Q Test



From ORNL-DWG 81.20400 ETD

160

r

-j

0

00

03

00

"-J

-J>OzU]._

150

140

130

120

110

100,

90

80

70

E M-RELAP5A M-RELAP5

0 Measured

REFLOODINITIATED

"0

LEVEL F

0 .0

0 o0

TOP OF BUNDLEHEATED LENGTH

I QUENCHED

I -LOWEST UNCOVERED FRSLEVEL (E6) QUENCHED

I I !

A

-385.08

E

I-346,30 --

0Co

,307.52 u

000

2'8374 <.J

LU.J229.96

191.18. :

152.40Hou . . • [ . [ . . . . '• • . II =

ý20 -10 0 10 20 30 40 50 60

TIMEi'(s)

*1 The time when the heat transfer mode switches from film boiling to transition boiling.

*2 The time when the heat transfer mode switches from transition boiling to nucleate boiling.

Figure 3.3-15 Quench Level for ORNL/THTF Test 3.09.10Q Test



3.4 FLECHT-SEASET Forced-Reflood Test

The applicability of the model for the rod heat transfer and rewet phenomena during thecore recovery phase incorporated in M-RELAP5 was confirmed by the comparison withORNLITHTF high-pressure reflood test data3-8. The pressure range of the tests is greaterthan 3.9MPa. On the other hand, the pressure decreases to [ ] during thecore recovery phase in a 1.0-ft2 cold leg break case of the US-APWR. Therefore,FLECHT-SEASET data 3 9 were used to benchmark the M-RELAP5 to confirm theapplicability of the heat transfer model in the low pressure core recovery condition.



The FLECHT-SEASET (Full-Length Emergency Core Heat Transfer Separate Effects andSystem Effects Tests) program was planned and conducted to obtain the core refloodcooling experimental data expected under large break LOCA conditions. The testsconsisted of forced and gravity reflood experiments and steam cooling tests, usingelectrical heater rods to simulate current nuclear fuel arrays (similar to the Westinghouse17X17 assemblies) of PWRs. The data obtained include rod clad temperature, turnaroundand quench times, heat transfer coefficients, inlet flooding rates, overall mass balance,differential pressure and calculated void fractions in the test section, thimble wall andsteam temperatures, and exhaust steam and liquid carryover rates.

The FLECHT facility for the low flooding rate test series 3-1° was utilized as the basicconfiguration illustrated in Figure 3.4-1. The design features of the facility include thefollowing:

a) A cylindrical low mass bundle housing to minimize housing heat releaseb) Housing differential pressure cells every 12-in (0.30m) to obtain void fraction

measurements along the heated length of the bundlec) Steam probes in each of 11 thimble tubes to measure steam superheat radially

and axially across the bundled) 177 heater rod thermocouple computer channelse) Housing windows at the elevations of 36, 72, and 108-inches (0.91, 1.83, and

2.74m)

The pressurized-water is supplied by an accumulator. The injection line of the accumulatoris equipped with three rotameters and one turbine meter to measure injection flow ratesfrom 0.4-in/s (10mm/s) in forced flooding tests, and up to 14.31b/s (6.49kg/s) in gravityreflood tests.

A close-coupled carryover tank connected to the upper plenum of the test section had aminimum capacity of 1451b (65.8kg). A steam separator with a capacity of 25001b/hr(0.315kg/s) and a liquid collection tank with a volume of 211b (9.5kg) to collect liquidentrained in the exhaust line were connected to the upper plenum of the test section. Thesteam separator had a storage capacity of approximately 4251b (193kg). The exhaustpiping had a system pressure control valve and an orifice plate flowmeter to measureexhaust steam flowrate. An electric steam boiler with a capacity of 1251b/hr (0.016kg/s)



was used to set the initial pressure and temperature of the loop.

During operation, coolant flow from the 400gal (1.51 M3) capacity water-supplyaccumulator entered the test section housing through a flow redistribution skirt in thelower plenum to assure proper flow distribution. The flow was regulated manually througha series of hand valves or automatically through a hydraulic control valve or series ofsolenoid valves.

The system pressure of the rest section was initially set by the electric steam boiler, whichwas connected to the upper plenum of the test section. During the experimental run, theboiler was valved out of the system and the pressure was maintained by a pneumaticallyoperated control valve located in the exhaust line.

Liquid effluent leaving the test section was separated in the upper plenum and collected inthe close-coupled carryover tank. A baffle assembly in the upper plenum was used toimprove liquid carryout separation and minimize liquid entrainment into the exhaust vapor.An entrainment separator located in the exhaust line was used to separate any remainingentrained liquid carryout from the vapor. Dry steam flow leaving the separator wasmeasured at an orifice section before exhausting to atmosphere. To help ensure correctmeasurement of the single-phase flow, the piping upstream of the orifice section washeated to a temperature well above the saturation temperature.

The cross section of the test bundle is shown in Figure 3.4-2 in the original configuration.The bundle comprised 161 heater rods (93 non-instrumented and 68 instrumented),thimbles instrumented with wall thermocouples, 12 steam probes, 8 solid triangular fillers,and 8 grids. The triangular fillers were welded to the grids to maintain the proper gridlocation. The fillers also reduced the amount of excess flow area from 4.7 to 9.3%. Most ofthe tests were performed under a uniform radial power profile, but some tests wereperformed with radial power distribution that assumes a 1.05 peak-to-average ratio basedon a quarter section simulation of a 17X17 PWR fuel assembly. The chopped cosineprofile with a peaking of factor 1.66 was imposed on the axial power distribution. Table3.4-1 summarizes the primary specifications for the FLECHT-SEASET test section andscaling to the US-APWR.


The following is a general procedure used to establish initial test conditions and perform atypical FLECHT-SEASET unblocked bundle reflood test.

The accumulator is filled with water and heated to the desired coolant temperature of 127'F (53°C). The boiler is turned on and brought up to nominal gage pressure of 75psig(0.42MPa). The carryover vessel, entrainment separator, separator drain tank, test sectionupper plenum, and test section outlet piping (located before the entrainment separator)are heated while empty to slightly above the saturation temperature corresponding to thetest run pressure. The exhaust line between the separator and exhaust orifice is heated to500'F (260'C) nominal and the test section lower plenum is heated to the temperature ofthe coolant temperature in the accumulator.

The test section, carryover vessel, and exhaust line components are pressurized to the



desired system pressure of 20 to 60psia (0.14 to 0.41MPa) by valving the boiler into thesystem and setting the exhaust line air-operated control valve to the desired pressure.The coolant in the accumulator is pressurized to 400psia (2.76MPa). Water is theninjected into the test section lower plenum until it reaches the beginning of the heatedlength of the bundle heater rods. Coolant is circulated and drained to assure that thewater both in the lower plenum and injection line are at the specified temperature prior tothe run.

Power is then applied to the test bundle and the rods are allowed to heat up. When thetemperature in any two designated bundle thermocouples reaches the preset value of 500to 1600'F (260 to 871'C), the computer automatically initiates flood and controls powerdecay. Solenoid valves in conjunction with a hydraulic control valve control coolantinjection into the test section. The exhaust control valve regulates the system pressure atthe preset value by releasing steam to the atmosphere.

After all the designated heater rods have quenched, as indicated by the rodthermocouples, power to the heater rods is terminated, coolant injection is terminated.The entire system is depressurized by opening a control valve, and the Computer DataAcquisition System (CDAS) is deactivated. Water stored in all components is drained andweighted.

The three forced reflooding tests, Runs 31504, 31701, and 32013 are selected to assessthe applicability of M-RELAP5 for analyses under lower pressure condition. Run 31504 isthe reference experiment, while Runs 31701 and 32013 correspond to the higherreflooding and higher pressure experiments, respectively. The test conditions of theseexperiments are summarized in Table 3.4-2.



Figure 3.4-3 illustrates the M-RELAP5 noding for the FLECHT-SEASET analyses. [


Figure 3.4-4 through Figure 3.4-9 show comparisons of the heater rod temperature at 72inch and 96 inch elevations, where the peak cladding temperature occurred in theexperiments due to the chopped cosine axial power distribution. M-RELAP5 predicts apeak cladding temperature higher than the test data. In terms of the rewetting, M-RELAP5predicts a dryout of longer duration than the test data. The tendency can be recognized inany experimental case, including those with lower to higher flooding rate conditions. Thiscode validation results demonstrate that M-RELAP5 provides a conservative prediction in



terms of the fuel cladding temperature even under the low pressure reflooding. Particularly,eliminating the reflood heat transfer model contributes to the conservatism describedabove.

3.4.3 Summary

The validation of M-RELAP5 using the FLECHT-SEASET forced-reflood tests indicatesthat the code is capable to conservatively predict the reflooding behavior even under thelow pressure condition with a wide range of coolant velocity, as well as under thehigh-pressure condition as the ORNL/THTF tests. In conjunction with M-RELAP5 codeassessment using the ORNL/THTF high-pressure reflood test data, the code provides aconservative heat transfer coefficient under the reflooding states, and is applicable forconservative analysis of the SBLOCA postulated transient of the US-APWR.

Table 3.4-1 Primary Test Section Specifications for FLECHT-SEASET

FLECHT- US-APWR/Item SEST US-APWR F-

SEASET F-S


Type of Assembly 17x17 17x17 -

Total No. of Rods per Assembly 177 289 1.63

No. of Heated Rods per Assembly 161 264 1.61

No. of Unheated Rods per Assembly 16 25 1.56

Heated-to-Unheated Rod No. Ratio 10.06 10.56 1.05





Rod Pitch (m) 0.0126 0.0126 1.00

Flow Area per Assembly (Mi) 0.0156

Hydraulic Diameter (m) 0.0097

Table 3.4-2 FLECHT-SEASET Forced-Reflood Test Conditions

Inlet Max Initial Linear heat FloodingTest PressureSubcooling Temperature power velocity

(psia) ('F) (°F) (kW/ft) (in/s)

31504 40 144 1507 0.7 0.97

31701 40 141 1640 0.7 6.10

32013 60 141 1555 0.7 1.04


=Z

(-D(D1

CM-ý-c

V--1

0(

CTGo

I-IC:

0)0

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SO I F ILLER

00000000 o000 00 0000

00 00 00 000

0~ OOOM X)00000 000000000000o8oooooooooooo700000 00000000

030~00000 00

1000 THIMBL

HEATER ROD

BUNDLE STATISTICS

Housing.tnside Diameter 19A.O mm 7.625 in.Housing Wall Thickness 5.08 mm 0.200 in.Rod Diameter 9.50 n" 0.374 in.Thimble Diameter 12.3 mm o. 4B54 in;Rod Pitch 12.6 mm 0.496 in.Cross-Sectional Floi Area 151476. mm

2 23.989 in.2

Filler Dimensions 19.143 m .64mm 0.765 in. xO.3140 in.161 Heater Rods16 Thimbles8 Fillers

Figure 3.4-2 FLECHT-SEASET Bundle Cross Section 3 9



Figure 3.4-3 M-RELAP5 Noding Diagram for FLECHT-SEASET Forced-Reflood Test



Figure 3.4-4 Rod Surface Temperature at 72-in Elevation (Run 31504)





.1




3.5 UPTF CCFL Test

Heat removal by the steam generators plays an important role during a postulated smallbreak LOCA transient when the break flowrate is small and the primary pressure remainshigher than the secondary side pressure. In this situation, condensed water in the SGU-tube either accumulates in the SG U-tube and SG inlet plenum or flows back to thereactor vessel against steam flow. Counter-Current Flow Limitation (CCFL) characteristicsin the SG U-tube and the hot leg will affect core cooling through the behavior of thecondensed water in the SG U-tube.

Verification of CCFL modeling in the hot leg region by the M-RELAP5 is conducted againstthe Upper Plenum Test Facility (UPTF) hot leg CCFL experiment 3-11.



The UPTF simulates a 4-loop German PWR which is similar to a US 4-loop WestinghousePWR. The overall view of the UPTF is shown in Figure 3.5-1. The UPTF consists of afull-size reactor vessel and piping (four hot legs and four cold legs). The emergency corecooling (ECC) can be injected in the hot and/or cold legs of all four loops, or into thedowncomer. One of the four loops contains break valves which are piped to a largecontainment simulator tank. The four steam generators are simulated by four steam/waterseparators and the four reactor coolant pumps (RCPs) are simulated by four passiveadjustable resistances. The reactor vessel upper plenum internals and top-of-core arefull-scale replicas. The core is simulated by a steam/water injection system with 193nozzles, one for each active fuel assembly commonly present in a PWR. The UPTF wasoriginally designed as an integral system test facility covering the end-of-blowdown, refilland reflood phases of a large break LOCA. As discussed in Reference 3-11, it has alsoproven very useful as a full-scale separate effects facility covering both large and smallbreak LOCA phenomena. The UPTF can operate at system pressure up to 18bar(260psia) and pressure of 2200C (4280F).

Figure 3.5-2 shows the hot leg configuration. Each UPTF hot leg has a 750mm (29.5-in)inner diameter and a total lateral run from the vessel to the steam generator simulator ofabout 8m (26-ft). A 50' riser section rises 0.91 m (3.0-ft) at the end of the hot leg attachedto the steam generator simulator. In the horizontal section of hot leg, an internal ECCinjection pipe, called the "Hutze", is located along the bottom edge of the pipe, as shownin Figure 3.5-3. There was no injection through the Hutze in the CCFL tests, i.e., it is adead space in the hot leg. The Hutze blocks an area of 0.0444m 2 (0.478-ft2), about 10percent of the total pipe area. A Hutze is present in German's PWRs but not in the USPWRs. Table 3.5-1 compares the UPTF hot leg configuration with that of typicalWestinghouse and Combustion Engineering US PWRs 3-1 . The hot leg diameter is 0.75m,which is similar to that of the US-APWR.

The test was conducted using only the broken loop hot leg of the UPTF. The test wasperformed at several steady phases, each consisting of steam injection into the primaryvessel that flowed out the broken-loop hot leg, and saturated water injection in the steamgenerator simulator plenum that could either flow back down the hot leg toward the vessel



or out of the system through the steam generator simulator. Figure 3.5-4 shows theoverall flow condition of the UPTF hot leg separate effect test. The system pressure andflow conditions of the test are as follows:

System pressure • 3bar, 15barWater flowrate • 30kg/sSteam flowrate :12kg/s to 20 kg/s six conditions for 3bar

24kg/s to 40 kg/s ten conditions for 15bar

One flow condition in the 3bar test and three flow conditions in the 15bar test do not showCCFL; all injected water goes down into the reactor vessel because of the small steamflowrate.


As mentioned previously, six separate steady flows were obtained at 3bar (44psia) systempressure and 10 flows were obtained at 15bar (218psia) system pressure. In all cases,water flow was established prior to steam flow. The purpose of obtaining several flows ateach pressure was to "map out" the CCFL boundary. In addition, one of the flowconditions at 15bar simulates the condition in Westinghouse 4-loop PWR during refluxcondensation mode, which can occur during an SBLOCA311.

There is a comparison between the experimental results and the Wallis correlationdiscussed in Reference 3-11. The Wallis correlation expresses the correlation of thedimensionless gas superficial velocityjg* to void fraction as shown in Figure 3.5-5.

Figure 3.5-5 shows that close agreement is obtained between the UPTF data and theWallis correlation. The latter is based on void fraction rather than liquid flow. This indicatesthat the basic approach of introducing .g* appears correct for scaling, but thatimplementing this model to calculate liquid flows is dependent on knowing an accuratevoid fraction.

The following correlation was obtained using the Wallis form of the CCFL correlation:

I7-= 0.7955-1.1564 rjg* (3.5-1)

The data on a dimensionless i9* were calculated using

Ig Mg-g (3.5-2)PgAJ(Pt - Pg )gDh

J f = gf ;7'- (3.5-3)

p, f(Pt -pg )gDh

where: M = mass flowrate of gas or liquidA = areap = density



g = gravity

Dh = hydraulic diameter

The line drawn through the data is the "best-fit" experimental correlation to the UPTF data.

The results of the test provided direct demonstration that there is significant marginagainst hot leg CCFL during the reflux condensation phase during an SBLOCA transient.This is shown by the fact that the "typical" point is substantially above the CCFL boundaryin Figure 3.5-5. The typical point was chosen based on conservative assumptions, suchas relatively high power, steam generator inactive, and so forth. Accordingly, this resultprovides direct and convincing evidence that substantial margin to CCFL exists.

Measured hot leg level and void fraction for all of the tests are plotted against jg*, thedimensionless gas flow. These data plotted in Figure 3.5-5 are from a three-beam gammadensitometer located just on the vessel side of the hot leg riser bends. There is no "Hutze"obstructing the bottom of the hot leg in this short section of hot leg. The data clearlyindicated a stratified regime and show significant water presence in this region of the hotleg. These data appear to show that CCFL is being controlled by the hot leg (i.e., CCFLdoes not occur in the steam generator simulator), since water is present in the hot legwhen there is zero net penetration to the vessel.



The parameters of CCFL correlation were derived from the test data for the analysis.The CCFL correlation used in M-RELAP5 has the following general form3-12 .

H1 /2 + m = c (3.5-4)-•1/2

Hi = i O (3.5-5)

w= D-=Lf (3.5-6)W

L- (3.5-7)

where j is superficial velocity, p is density of fluid, g is gravity constant, Dj is junctionhydraulic diameter, L is the Laplace capillary constant and o is surface tension.

The parameters m, c, and 3 are used as input data. Suffixes g and f represent gas andliquid respectively. Generally, hydraulic diameter dependency disappears for largediameter pipe and 3 is assumed one (1.0). The constants m and c are calculated by linearregression analysis of the experimental data. Figure 3.5-6 shows results of the linearregression analysis and [ ] are obtained.

To confirm the validity of the derived CCFL correlation with M-RELAP5, analysis of theUPTF test data is performed using the CCFL correlation. The flow conditions applied to



the analyses are shown in Table 3.5-2. The nodalization applied for the analysis is shownon Figure 3.5-7.


Results of M-RELAP5 analysis with the CCFL correlation are shown on Table 3.5-3, Table3.5-4, and Figure 3.5-8. In the tables, positive water flow means downflow to the reactorvessel and negative flow means almost complete flooding condition and unstable flowoccurs. The characteristic of water downflow rate against steam upflow rate is wellreproduced by the analysis for both 3bar and 15bar conditions.

3.5.3 Summary

The CCFL parameters for a large diameter pipe were derived from the UPTF CCFL testdata. The M-RELAP5 analysis with the derived CCFL parameters well reproduced theUPTF test data. It is confirmed that M-RELAP5 with the CCFL parameters is applicable toCCFL behavior of the hot leg and the SG plenum in the US-APWR.



Table 3.5-1 Comparison of UPTF Hot Leg Configuration with Typical Westinghouseand Combustion Engineering (CE) PWRs

Parameters UPTF Westinghouse CE PWRValue PWR Value Value

Diameter, m (in) 0.750(29.5) 0.737(29) 1.07(42)Hydraulic Diameter, m (in) 0.639(25.2) 0.737(29) 1.07(42)Flow Area, m2 (ft2) 0.397(4.28)* 0.427(4.59) 0.894(9.62)

*0.4418mz within diameter minus 0.0444mz blocked by "Hutze" -

Table 3.5-2 Flowrate Conditions for UPTF CCFL Test Analysis

System Flowrate of injected steam (kg/s)pressure

15 bar 24. 25. 28. 30. 31. 32. 33. 34.35.37.39.40

3 bar 12.13.14.15.16.17.18.20

Table 3.5-3 Results of UPTF CCFL Test Analysis (15bar)

Table 3.5-4 Results of UPTF CCFL Test Analysis (3bar)

Steam mass flow [kg/s] 12 1 13 14 15 16 J 17 1 18 20Water mass flow[kg/s] 25.9 25.2 13.8 12.3 7.7 -0.9 -1.1 -1.4



I Trt Venal 3c Draing Veal for Hot Log2 Steam Genatr .smola= r 3d Drulgo VWI fo Cold Log

0nInc Loop) 4 Pan* Simulator3a Stua m Generator Simulated 5 Break Val"- (Net L.

Wowy SeparatorBmre iLop Ho Sb Break Valve (Cold Lag

3b Water Separao 6 Contamert SkriLater

(flike Loop Cowd 60g

(0 ECCIn.jae-in Nozzles (Cold Lat

(D ECC.niaedon Nozaio (out Lag

@® Cc- Simalato, lajoetio o d

@ TVD.ranage Nmaui

( Steum laj~rm Nozaul

® Drainage.

Figure 3.5-1 Overall View of UPTF Test Facility3 11

mft)

Note: These dimensions are for the UPTF broken loop hot leg, which was the only hotleg used in the Hot Leg Separate Effects Test. In the intact loops, thesetwo dimensions are slightly larger (3.86 m and 1.34 m).

Figure 3.5-2 UPTF Hot Leg Configuration 311



WI. A-. - 0.0a244

, (0. fti)

CM FlO Arta - 0.397W4 (4.Z78 ft2)

rfdroulc Dlmeta* •0" . (2.10 ft)

(lisozel

r0- A-,

(0.63 to)

h. 13-(0.51 t.)

wsmc" A-A

Figure 3.5-3 Configuration of International ECC Injection Pipe (Hutze)3 11

Section AA

Section B-BDmhwp P-

Ai

smm VWaWnm -0-ft syd

Hothlg Ecc*~cdmo pipe 'Huba(Ohtacthratedo A hat log

Figure 3.5-4 UPTF Hot Leg Separate Effect Test Overall Flow Conditions 3 11



UA 1.1 1-

I - _________ 0.91.

V2t M" It.r IIn 1

* . TypIC&I P110 SELOU Rfla

:1:"--LI• Mo•.€•d F~tqýtS wtftenrly

I j0 l. r

UPTF LOOP 4 NOT LEO 0.7

•~ ~ ~ 1 " SrIT Ont,q:0.4 -/FDt

0.0

0.0 0.1 a.7 o.o5 .0.Dl~sloles Stptf~calSt.-e Veloity. p

Figure 3.5-5 UPTF Hot Leg Separate Effect Test Comparison of UPTF Hot Leg VoidFractions to Wallis Correlation 3 11



\I- __I

Figure 3.5-6 Regression Analysis Results of UPTF CCFL Parameters f8, c and m forM-RELAP5



Figure 3.5-7 Nodalization Scheme for UPTF CCFL Test Analysis



Figure 3.5-8 UPTF CCFL Test Analysis, Comparison of Flooding Curves of AnalysisResults and Test Results



3.6 Dukler Air-Water Flooding Test

Heat removal by the steam generator plays an important role in the small break LOCAwhen the break flowrate is small and the primary pressure remains higher than thesecondary side pressure. In this situation, condensed water in the SG U-tube eitheraccumulates in the SG U-tube and SG inlet plenum or flows back to the reactor vesselagainst steam flow. Counter-Current Flow Limitation (CCFL) characteristics in the SGU-tube and the hot leg will affect core cooling through the behavior of the condensedwater in the SG U-tube. Verification of CCFL modeling in a relatively small diameter pipe,like a SG U-tube, by the M-RELAP5 is conducted against the Dukler Air-Water FloodingTest 3-13.

The tests were carried out at the four liquid flow conditions of 100, 250, 500, and 1000Ibm/hr at low pressure and atmospheric temperature. Air flowrate was varied at each liquidflowrate condition and steady condition was established. Water downflow rate wasmeasured at each air flowrate.



Figure 3.6-1 shows a schematic of the Dukler air-water flooding test facility. The testfacility consists of a "calming" section which separates the falling liquid film and guides airinto the test section (flow length: 1.524m, ID= 0.0508m), a Test section (flow length: 3.96m, ID=0.0508 m), an air-inlet section and an exit section which separates upflow waterfilm and air. Entrained water is separated in the separator further downstream before airexhaust. The input flowrates, pressure and liquid film flowrates were measured duringeach test. The flowrate of the rising liquid film and the flowrate of entrainment weredetermined by the weight change of a collection tank. The entire system was supported bya unistrut structure and all air and water connections were by flexible tygon tubing toeliminate external vibrations.

A schematic drawing of the entrance section is shown in Figure 3.6-2. The air inlet sectionwas designed to remove the liquid film falling counter to the air flow and to provide asmooth entrance for the air. The entrance consisted of a 12-inch I.D. section of plexi-glasspipe containing a smooth flange at the top connecting to the test section and at the bottoma 2-inch I.D. section of pipe which could be moved vertically various distances from thesmooth flange. In order to prevent undesirable pressure fluctuation in the entrance theliquid level had to be maintained constant. This was accomplished through the use of aliquid-level control system consisting of two metal electrodes, a relay, and a solenoid valve.The falling liquid film passed over an expansion nozzle which caused the outer diameterof the liquid sheet to expand from 2.00" to approximately 5.0". After leaving the nozzle theliquid sheet spread still farther thus avoiding interaction with the rising air jet. Visualobservations indicated this was an excellent method of separating the film without creatingentrainment. Liquid flowing out of this section was either weighed to determine the amountof downflow or recirculated to the system.

A sketch of liquid entrance device is shown in Figure 3.6-3. The device was constructedto provide a smooth liquid film at the liquid entrance. Water entered an annulus whose



inside wall was made of porous sintered metal and passed through the porous metal toform a smooth film. The inner wall of the porous metal was sized to coincide with the testsection I.D. to prevent discontinuities.

A sketch of the exit section is shown in Figure 3.6-4. The exit section consisted of asmooth flange over which the liquid film flowing up was separated from the air steamcontaining entrained droplets. In a manner similar to that of the air entrance, the sheet ofliquid was expanded in diameter across an expansion nozzle. The film passed around theair removal pipe and fell to the liquid surface in the tank. The air and entrained dropsflowed out of the separator. In this way it was possible to distinguish between liquid upflowin the film and that which was entrained. However, under churn flow conditions, a portionof the continuous liquid phase could be captured across the outlet pipe. Similarly, underlower liquid flow conditions, some of the air could be expected to flow into the reservoirbefore leaving in the air line. Thus the entrainment measurement must be considered onlyan approximate measure of the entrainment actually existing under upflow conditions inthe test section. This section, like the entrance, was equipped with a liquid level controller.


Experimental data were obtained which described the major features of the system beforeand during flooding. Results included the following:

Liquid film upflowLiquid film downflow

Verification of the experimental results was carried out using a general correlation withdimensionless parameters in Reference 3-14. Dimensionless groups, which are related tomomentum fluxes, are shown as follows:

jig = jgP1/2 [gD(pf - pg)]-1/ 2 (3.6-1)jf= fp/[gD(pf - pg)]-l/ (3.6-2)

Correlations for flooding in vertical tubes may be expressed in the general form.

j*1/2 + mj;12C (3.6-3)

For turbulent flow m is equal to unity. The value of C is found to depend on the design ofthe ends of the tubes and the way in which the liquid and gas are added and extracted.For tubes with sharp-edged flanges, C = 0.725, whereas when end effects are minimized,C lies between 0.88 and 1.0. Table 3.6-1 shows water and air flowrate conditions of thetest selected for the verification analysis and also resultant water downflow rate incomparison to analysis result. Figure 3.6-5 shows the relation about the floodingvelocities for air and water in vertical tubes designed to minimize end effects. InReference 3-13, it is mentioned that Hewitt and Wallis found that for an air-water systemthe flooding velocities could be correlated by the equation

j*gl2 + j*112 = 0.88 (3.6-4)





Figure 3.6-6 shows the nodalization diagram used for this analytical study. The model issummarized as follows: [

]

The parameters of CCFL correlation proposed by Hewit and Wallis 3-14 are used for thisanalysis.

/8= 0.0c = 0.88 (3.6-5)m= 1.0

Jg 0 5 + jf = 0.88 (3.6-6)

.. = jg og" (3.6-7)Ig =IJggD(pf-pg)

j; = jf gD((3.6-8

where jg is the vapor/gas superficial velocity, j] is the liquid superficial velocity, pg is thevapor/gas density, pf is the liquid density, g is the gravitational acceleration, and D is thehydraulic diameter.


Figure 3.6-7 shows characteristics of water downflow rate against air upflow rate incomparison to the test data. The liquid downflow rate is somewhat underestimated, butthe overall agreement with the test data is good. The calculated water downflow rate is30% smaller than the test data in average. Figure 3.6-8 shows the same characteristicswith non-dimensional form. In both figures, the M-RELAP5 calculations agree well with thesolid line, which comes directly from the CCFL correlation.

3.6.3 Summary

Analysis of the Dukler Air-Water Flooding Test was conducted using M-RELAP5 withCCFL parameters proposed by Hewitt and Wallis. The analysis results show good



agreement with the test data. This verification analysis demonstrates that M-RELAP5 withthe CCFL parameters proposed by Hewitt and Wallis is applicable to simulation of CCFLbehavior of a small diameter pipe, such as a SG U-tube in the US-APWR.

Table 3.6-1 Four Different Input Liquid Flowrates for Dukler Air-Water Flooding Test

WL(lb/h) 100 250 500 1000

ReL 310 776 1552 3105

(From Reference 3-13)

Table 3.6-2 Experimental Results for Dukler Air-Water Flooding Test

Input Liquid Experimental ResultsPoint Flowrate Input Gas Liquid Flow

(Ibm/hr) Flowrate Down(Ibm/hr) (Ibm/hr)

1 100 250 1002 100 265 503 100 277 104 250 218 2505 250 232 1806 250 247 1057 250 262 558 250 278 109 250 292 010 500 192 49011 500 214 31012 500 231 20013 500 251 10514 500 269 4515 1000 133 100016 1000 159 72017 1000 185 52518 1000 210 37019 1000 229 20520 1000 262 60




AIR-WATER SEPARATOR 9

INDICATES MEASURINGSTATIONSINDICATES DISCHARGETO WEIGH TANK

UPFLOWENTRAINMENT

FLOWMETERS

a1

WATER TANK

ORIFICE METERSI • AIR INLET

Figure 3.6-1 Schematic of Dukler Air-Water Test Facility3 13



Figure 3.6-2 Air Inlet Section of Dukler Air-Water Test Facility3 13

Figure 3.6-3 Liquid Entrance Device of Dukler Air-Water Test Facility3 13



Figure 3.6-4 Exit Section of Dukler Air-Water Test Facility3-1 3



1.07N+

V\\% I

U.8

0.6

0.4

0.2

VA0 +1

N

0

ax

S-a----- O-

0

"X,+

\+x0c= 0.88 '\ 0+ 0

\ I

I 10.6 0.8

0 L0 0.2 ' 0.4

o I"-dic. (Nicklin and Davidson 2 8)well-rounded air inlet design.

x I"-dia. (Nicklin and Davidson28)well-rounded air inlet design.

"short column"

"main column"

1 /4"-dia. (Hewitt and Wallis 21) water iniectionand extraction through porous walls. Pointswhere flooding starts with increasing flow.

3 1'/&"-dia. (Hewitt and Wollis 2 ') points whereflooding stops with decreasing flow.

+ 3/4"-dia. (Wallis, Steen, and Brenner 29)

flooding starts.

v 3/4"-dia. (Wallis, Steen, and Brenner2)flooding stops.

Figure 3.6-5 Flooding Velocities for Air and Water in Vertical Tubes Designed toMinimize End Effects (Atmospheric Pressure)3 14



Figure 3.6-6 Noding Scheme for Dukler Air-Water Flooding Test Analysis



Figure 3.6-7 Comparison of Calculated and Measured Results using Wallis FloodingCorrelation Constants (C=0.88 and m=1.0)

Figure 3.6-8 Comparison of Calculated and Measured Results using Wallis FloodingCorrelation Constants (C=0.88 and m=1.0)



3.7 UPTF Test 5

Loop seal clearing is one of the identified phases during a postulated small-break LOCAtransient in the US-APWR. The phase occurs subsequent to the natural circulation phaseand prior to the core level boil-off phase. The loop seal clearing phase is important since ithas substantial influence on the increase of fuel cladding temperature or core heat-up.Experiments to investigate the loop seal clearing mechanism were performed using theGerman Upper Plenum Test Facility (UPTF). The UPTF Test 53.15 was a separate effecttest carried out to study loop seal clearing phenomena and to provide data for improvingthe performance of computer codes for LOCA analysis. The purpose of the UPTF Test 5was to study the amount of water remaining in the loop seal under different steamflowrates.



As the described in Section 3.5, the UPTF is a full scale simulation of the primary systemof 1300-MW PWR from Siemens-KWU. Figure 3.7-1 shows a general view of the primarysystem. The upper plenum of the test vessel, including original reactor internals, thedowncomer and the four connected loops are represented on a 1:1 scale. The thermalhydraulic response of the reactor core is simulated by controlled injection of steam andwater. All four loops are equipped with valves to simulate the overflow edge and the flowresistance of the RCPs and steam-water separators representing the steam generators.The hot and cold legs of the broken loop lead through steam-water separators and breakvalves to the containment simulator. Breaks of variable size can be simulated in the hot orin the cold leg accordingly. The maximum operating system pressure of the UPTF is20bar.

In the UPTF, all four loop seals are instrumented with differential and absolute pressuretransducers as well as thermocouples for fluid temperature detection. The water levelmeasurement is based on differential pressure. The steam flow through the loop seal ismeasured with an orifice.

The separate effect tests were performed in a single-loop operation using the loop seal inloop-2. A schematic of this loop seal, including the instrumentation is shown in Figure3.7-2. A calibrated turbine meter is used to measure the steam flow through the loop seal.Water, injected into the downward vertical part of the loop seal (simulation of thecondensate from the reflux condenser operation, shown in Figure 3.7-1) is detected by anorifice in the injection line. The fluid temperature in the loop seal of loop-2 is measuredwith thermocouples (TC).


The separate effect tests were performed in a single-loop seal connected to the full-scale1:1 primary system to investigate the residual water levels, the flow patterns in and thepressure drops across the loop seal during clearing at system pressures of 3 and 15 bar.Loops 1, 3 and 4 were sealed off by closing the pump simulators. The boundary and initialconditions are shown in Table 3.7-1 adopted from Reference 3-15.



The tests were initiated by injecting steam into the test vessel, which was controlled sothat a constant steam flow through the loop seal always existed. Most of the tests wereperformed starting with simultaneous steam and water injection ramped up to a specifiedvalue. Shortly after the water had reached a constant level in the loop seal, the residualwater level was estimated and the water injection was switched off to maintain thepresence of steam phase only. Residual water levels for steam flow were taken whenthere was zero droplet entrainment from the horizontal loop seal section. At the end of thetest period, the steam flow was reduced to zero to obtain data for mass balance purposes.

Figure 3.7-3 shows the comparison of residual loop seal water levels from the UPTF Test5 and experimental results given in the Reference 3-15. The mean residual water levels ofthe UPTF tests at 3 and 15 bar with steam flow alone are compared with results from theliterature (Noailly, 1980; Tuomisto and Kajanto, 1988; Junghans et al., 1990. Only the IVOtests (Tuomisto and Kajanto, 1988) were performed in a full-sclae single-loop seal (scale1.13:1 compared to UPTF) with air flow at atmospheric pressure. The diameter of the IVOloop seal was 0.850 m (UPTF: d=0.750m, ratio L/d=2.3). Figure 3.7-3 reveals a widescattering of data and different residual water levels for a given gas Wallis parameter.Higher water levels were measured at 15 bar than at 3 bar in the UPTF. The IVO data at 1bar show lower values than the UPTF results at 3 bar3-15 . Thus, the experiments indicatethat the residual water level increases as the pressure increases.



Verification of M-RELAP5 applicability using UPTF Test 5 data was performed. For theanalysis parameter, as described in Reference 3-15 the Wallis parameterlg* for steam andwater is the most appropriate dimensionless number to ensure similarity of controllingphenomena like water entrainment, flow pattern, steam-water counter-current flow andbreak flow. Therefore, the steam Wallis parameter has been chosen as the basis forpressure scaling of steam mass flows in the UPTF:

L/) = s (3.7-1)gD=pýg O- pJ ) gD(p., - pJ

As described in Reference 3-15, the steam mass flowrates in the UPTF can be scaled tothose in a PWR by:

Js.UPTF (1 5bar) = j*,WR (8Obar) (3.7-2)

Figure 3.7-4 displays the nodalization scheme and injection procedure for UPTF testanalysis. The base nodalization (a) corresponds to that used in the US-APWR and a finernodalization (b) is developed for the sensitivity analysis. [




Figure 3.7-5 compares the results of M-RELAP5 calculation against the UPTF Test 5 data.M-RELAP5 predicts well the qualitative relation between the residual amount of water andthe steam flowrate, but generally overestimates the amount of residual water. Thisquantitative tendency means that less water being supplied from the crossover-leg intothe core region, which corresponds to a conservative prediction of core thermal-hydraulicsduring the core boil-off and core recovery phases.

3.7.3 Summary

Based on the verification of M-RELAP5 calculation using the UPTF Test 5 experimentaldata, the M-RELAP5 has a tendency to overestimate the residual amount of water in thecrossover-leg. It means that there is less water being supplied from the crossover leg intothe core region. This overestimation shows conservatism of M-RELAP5 in the predictionof core thermal-hydraulics behavior during core boil-off and core recovery phases.

Table 3.7-1 Boundary and Initial Conditions for UPTF Test 5

System pressure (bar) 3 or 15Steam flows through single loop seal (kg/s) 3 to 34 in steps in about 1Injected water mass flowrates (kg/s) 8, 15, 19 and 22Flow resistance of the pump simulator in the Simulation of blocked pump:investigated Loop-2 4Ps = 18 referred to the flow

cross-section area of the main coolantline (AMcL = 0.4418 M2 )

Flow resistance of the pump simulator in loops Completely closed1, 3 and 4Initial water inventory in the loop seal Partly water-filled horizontal section




I TetVe 3c Drainage Venalr Hot Log

2 Stem Generator S.amlatm 3d Drainago Vasel to Cold Leg(ormact ioot4 4 Pump Simulator

3a Steam Generator Sinuliod 5 Break Vales (Not LagWatr Sepoat(Broken Lop Hot Lt Sb Break Voav (Cold Leg)

3b Water Separator 6 Contaieet ShIa

Vlrakes Loop Cold! Wa

) ECCc-4actian Nozzles Cold W

(D ECC~lrjection Nodle, (Hot LaaJ

@ Cars StUINNta 1r1ei- Nozzle® TY.Dai Nzcle

@ Steam It•o imo Nozzie

@ Drainage Naze

Figure 3.7-1 Overall View of UPTF Test Facility 3 "13

Steam flow fromsteam generator

simulator

L= 1734 -- •-- 3330

Note: All dimensions are in mm

Figure 3.7-2 Loop Seal Portion and Instrumentation Configuration of UPTF3 15




UPTF:Sc•ala 1:1'1Lid 2.3steanbwaterpmsswe: 3j5 bar

4 -0.75m r

Amgha.•s et &I. (1990) U,, = 16,8swale 1-7.5 arkwawer

prssre t bar

40.l0M

tto ts d A (U14):LUd = 5.1Scae1.13:1 *MawT

pm~Ssff: I a

0.65 m

Noally (1980): airkder

scalel3 presure: Ib

... ,0125m ,,,

I

cu

0.8

0.6

0.4

a.2

00 0.1 0.2 0.3 0.4

Wallis pawmeti ior gas JX0

.5

Figure 3.7-3 Comparison of Residual Loop Seal Water Levels fromExperimental Results 3 15

UPTF Test 5 and



K

Figure 3.7-4 Noding Scheme and Injection Procedure for Vapor and Water againstUPTF Test 5



-I

Figure 3.7-5 Assessment Results for Residual Water Amount in UPTF Test 5



3.8 Advanced Accumulator Test

An advanced accumulator design 3-16 is used in the US-APWR. The unique feature of theadvanced accumulator design is to be able to control the injection flowrate using a flowdamper. The advanced accumulator is designed to initially inject a large amount of coolantjust after activation that compensates for the loss of coolant from the LOCA. After theinitial high flow period, the advanced accumulator will inject water at a small flowrate forlonger-term cooling after the initial high flow injection.

The present section describes validation for the advanced accumulator modelimplemented into M-RELAP5 using the scaled experimental data conducted by MHI. It isnoted that the advanced accumulator model is explained in the M-RELAP5 topicalreport3> 17 and/or Volumes I and IV of the M-RELAP5 supplementary manuals 3-18,3-19.



MHI has conducted several scaled tests for development of the advanced accumulator.The full-height 1/2-scaled experimental data are selected for the present code verification.Details of the experimental facility are given in Reference 3-16.


The full-height 1/2-scaled test cases selected for validation analysis simulate ECCSperformance during a large LOCA and are shown in Table 3.8-1. The following four caseswere tested on initial tank pressure that reflects the Accumulator operating conditionsduring a large LOCA. The pressure of the exhaust tank corresponds to RCS pressure.

Case 1: The initial test tank pressure was 586 psig (4.04 MPa [gage]) simulating thecondition for ECCS performance during a large LOCA.

Case 2: The initial test tank pressure was 657 psig (4.53 MPa [gage]) to obtain datafor high pressure design.

Case 3: The initial tank pressure was 758 psig (5.23 MPa [gage]) to obtain data forhigh pressure design.

Case 4: The initial tank pressure was the same as Case 1. However, the pressurein the exhaust tank was maintained at 71 psig (0.49 MPa [gage]) to obtaindata for high backpressure.

The pressure in the exhaust tank was 14 psig (0.098 MPa [gage]) for Case 1, 2, and 3.Since the pressure of the exhaust tank becomes the same as the pressure of thecontainment vessel (CN) after the blowdown phase during a large LOCA, and ECCSperformance analysis uses approximately 14 psig (0.098 MPa [gage]), the backpressurewas set at 14 psig (0.098 MPa [gage]). The experimental results are described inReference 3-16.





The noding diagram is shown in Figure 3.8-1. [

Pressure, water level, temperature, etc. were supplied by input data as the initialconditions of the accumulator tank, based on the test data. However, because gas andliquid phase were assumed to be in equilibrium in M-RELAP5, the gas temperature wasset to the same value as that of the coolant. The heat transfer between the accumulatortank wall and nitrogen gas was simulated here because this is a test analysis.


After input data describing the test system were prepared, the analysis carried out until170 seconds. Figure 3.8-2 through Figure 3.8-13 show the analysis results of theinjection volumetric flowrate, the tank pressure and the tank water level for four cases incomparison with the test results. In each case the analysis results are in good agreementwith the test results, and it is shown that the injection characteristic is well simulated bythe advanced accumulator model. In particular, the analysis results reproduce the testresults very well with regard to the tank water level, which is the integration value of theinjection volumetric flowrate.

3.8.3 Summary

The advanced accumulator model implemented into M-RELAP5 are verified and validatedby using the test data obtained in the full-height 1/2-scaled test facility. M-RELAP5 iscapable of reproducing the injection flowrate, tank pressure and water level behaviorsaccurately.

In applying the advanced accumulator model to the safety analysis, the total uncertaintyconcerning the flow damper is addressed in Appendix D of the M-RELAP5 topicalreport3-17, including discussion about the uncertainty of the flow resistance and the waterlevel that switches flow resistance. The uncertainties are quantified based on the fullheight 1/2 scale test data. The uncertainties of the flow damper resistance and the flowswitching level are considered deterministically for the US-APWR SBLOCA analysis.



Table 3.8-1 Full-Height 1/2-Scaled Advanced Accumulator Test Conditions

Test Exhaust Initial Injection Water

Tank Tank Gas Volume

Pressure Pressure Volume Large Small ObjectiveFlow Flow

psig psig ft3 ft3 ft3

[MPa [gage]] [MPa [gage]] [M3] [M3] [M3]Obtain flow characteristics for

Case 1 586 14 ECCS performance(4.04) (0.098) evaluation during a large

LOCA

657 14 Obtain flow characteristics forCase 2 (4.53) (0.098) high pressure design

758 14 Obtain flow characteristics forCase 3 (5.23) (0.098) large differential pressure

586 71 Obtain flow characteristics forCase 4 (4.04) (0.49) small differential pressure

(404 0.9

Figure 3.8-1 Noding Scheme for Advanced Accumulator Test



-I

Figure 3.8-2 Injection Volumetric Flowrate for Advanced Accumulator Test (Case 1)

Figure 3.8-3 Tank Pressure for Advanced Accumulator Test (Case 1)



Figure 3.8-4 Tank Water Level for Advanced Accumulator Test (Case 1)





Figure 3.8-13 Tank Water Level for Advanced Accumulator Test (Case 4)K



3.9 References

3-1 Y. Anoda et al. , 'Void fraction distribution in rod bundle under high pressureconditions,' HTD-VoI.155, Am. Soc. Mech. Eng., Winter Annual Meeting, Dallas,Nov. 25-30, 1990.

3-2 The ROSA-IV Group, 'ROSA-IV Large Scale Test Facility (LSTF) SystemDescription,' JAERI-M 84-237, January 1985.

3-3 The ROSA-IV Group, 'ROSA-IV Large Scale Test Facility (LSTF) SystemDescription for Second Simulated Fuel Assembly,' JAERI-M 90-176, October 1990.

3-4 D. K. Felde et al., "Facility Description - THTF MOD 3 ORNL PWR BDHTSeparate-Effects Program," NUREG/CR-2640, ORNL/TM-7842, September 1982.

3-5 T. M. Anklam, R. J. Miller, and M. D. White, 'Experimental Investigations ofUncovered-Bundle Heat Transfer and Two-Phase Mixture Level Swell underHigh-Pressure Low Heat-Flux Conditions,' NUREG-2456, ORNL-5848, March1982.

3-6 USNRC, Compendium of ECCS Research for Realistic LOCA Analysis,NUREG-1230 Revision 4, 1988.

3-7 D. G. Morris, C. B. Mullins, and G. L. Yoder, "An Analysis of Transient Film BoilingOf High-Pressure Water In A Rod Bundle," NUREG/CR-2469, ORNL/NUREG-85Rev.2, March 1982.

3-8 C. R. Hyman, T. M. Anklam, and M. D. White, 'Experimental Investigations ofBundle Boiloff and Reflood under High-Pressure Low Heat-Flux Conditions,'NUREG-2455, ORNL-5846, April 1982.

3-9 M. J. Loftus et al., PWR FLECHT-SEASET Unblocked Bundle, Forced and GravityReflood Task Data Report, NUREG/CR-1532, June 1980.

3-10 E. R. Rosal et al., 'FLECHT Low Flooding Rate Skewed Test Series Data Report,'WCAP-9108, May 1977.

3-11 P. S. Damerell, N. E. Ehrich, K. A. Wolfe, 'Use of Full-Scale UPTF Data to EvaluateScaling of Downcomer (ECC Bypass) and Hot Leg Two-Phase Flow Phenomena,NUREG/CP-0091 Vol.4, CONF-8710111-Vol.4.

3-12 'RELAP5-3D Code Manual Volume I1: User's Guide and Input Requirements,'INEEL-EXT-98-00834, Revision 2.3, April 2005.

3-13 A. E. Dukler and L. Smith, 'Two Phase Interactions in Counter-Current FlowStudies of the Flooding Mechanism,' Annual Report November 1975 - October1977, NUREG/CR-0617, January 1979

3-14 G. B. Wallis,'One dimensional Two Phase Flow,' McGraw-Hill, 19693-15 J. Liebert and E. Emerling, 'UPTF Experiment Flow Phenomena during Full-Scale

Loop Seal Clearing of a PWR,' Nuclear Engineering and Design 179, pp51-64,1998.

3-16 Mitsubishi Heavy Industry, Ltd., 'The Advanced Accumulator,' MUAP-07001-P (R2),September 2008.


3-18 Mitsubishi Heavy Industries, Ltd., 'M-RELAP5 Code Supplementary ManualVolume I: Code Structure, System Models and Solution Methods,'6AS-1 E-UAP-090047 (RO), August 2009 (Proprietary).

3-19 Mitsubishi Heavy Industries, Ltd., 'M-RELAP5 Code Supplementary ManualVolume IV: Models and Correlations,' 6AS-1E-UAP-090071 (RO), December 2009(Proprietary).



4. INTEGRAL EFFECTS TESTS

4.1 ROSA-IV/LSTF SB-CL-18 Test

The purpose of ROSA-IV/LSTF4-1 ,4-2 calculation is validation of M-RELAP5 codeperformance to predict phenomena ranked high importance in PIRT for US-APWRSBLOCA. There are 5% break tests, 10% break test, 0.5% break tests, 2.5% break tests,and so on in the series of SBLOCA tests. The test SB-CL-18 4

-3 (5% break) was selected

as the baseline lET for the M-RELAP5 assessment because both loop seal phenomenaand boil off phenomena considered important for SBLOCA were observed in theexperiment. The scaling to US-APWR SBLOCAs is well investigated in Reference 4-4.



The ROSA-IV/LSTF is a 1/48 volumetrically scaled model of a Westinghouse-type3423MWt four loop PWR. The LSTF has the same major component elevations as thereference PWR to simulate the natural circulation phenomena, and large loop pipes (hotand cold legs of 207mm in diameter) to simulate the two-phase flow regimes andphenomena of significance in an actual plant. The LSTF equipment can be controlled inthe same way as that of the reference PWR to simulate long term operational transients.Furthermore, the LSTF is designed to be operated at the same high pressures andtemperatures as the reference PWR.

Figure 4.1-1 and Table 4.1-1 show the structure and major dimensions of the LSTF,respectively. Figure 4.1-2 shows the pressure vessel assembly. The four primary loops ofthe reference PWR are represented by two equal-volume loops. The overall facility scalingfactor is 1/48. The hot and cold legs were sized to conserve the volume scaling and ratioof the length to the square root of pipe diameter, L/f-D for. the reference PWR inexpectation that the flow regime transitions in the primary loops can be simulatedappropriately by taking this scaling approach.

The core power is scaled by 1/48 at core powers equal to or less than 14% of the scaledreference PWR rated power. The LSTF rated and steady-state power is 10 MWt, i.e., 14%of the rated reference PWR core power scaled by 1/48.

Fuel assembly dimensions, i.e., fuel rod diameter, pitch and length, guide thimblediameter pitch and length, and ratio of number of fuel rods to number of guide thimbles,designed to be the same as the 17 x 17 fuel assembly of the reference PWR to preservethe heat transfer characteristics of the core. The total number of rods was scaled by 1/48and is 1064 for feasted and 104 for unheated rods. The primary specifications are given inTable 3.1-1 of Section 3.1.


The major initial conditions of the LSTF 5% cold log break test, SB-CL-18, are shown inTable 4.1-2. Both the initial steady state conditions and the test procedures weredesigned to minimize the effects of LSTF scaling compromises on the transients duringthe test.



The most important design scaling compromise is the, 10MW maximum core powerlimitation, 14% of the scaled reference PWR rated power. The steady-state condition isrestricted to a core mass flowrate that is 14% of the scaled value to simulate the referencePWR temperature distribution in the primary loop. The desired primary coolant flowratewas established by reducing the pump speed with the flow control valves (FCVs) in thecross-over legs fully open. The primary loop flowrate was then increased at the time ofbreak to improve the similarity of the LSTF to the reference PWR by increasing the pumpspeed.

The primary-to-secondary heat transfer must also be maintained at 10MW, i.e., 14% of thescaled value. Since the LSTF steam generators (SGs) are geometrically scaled to thereference PWR, the 14% primary-to-secondary heat transfer rate is established by raisingthe secondary temperature such that the primary pressure and temperature isrepresentative of the reference PWR.

Major operational setpoints and conditions including emergency core cooling system(ECCS) actuation logic for this test are summarized in Table 4.1-3 and Table 4.1-4. Afterthe break occurred at time zero, the primary system depressurizes quickly. At apressurizer pressure of 12.97MPa, the reactor scrams. Loss of offsite power concurrentwith the reactor scram is assumed and the primary coolant pumps are tripped to begincoastdown and the core power begins to decrease along the pre-programmed decaycurve. The power decay curve used in the test takes into account the fission products andactinides decay powers, and delayed neutron fission power. The decay power historyused for the present test gives a slower decrease than the ANS standard. The core powerdecay curve used in the test is tabulated in Table 4.1-5. The SG auxiliary feedwater isassumed to fail to simplify the transient.

At a pressurizer pressure of 12.27 MPa, the safety injection signal is sent that trips ECCSto be actuated at the respective pressure setpoints. However, the high pressure chargingsystem and high pressure injection system (HPIS) are assumed to fail in the test. TheECCS conditions are summarized in Table 4.1-4. The accumulator (ACC) system and thelow pressure injection system (LPIS) are specified to initiate coolant injection into theprimary system at pressures of 4.51 and 1.29 MPa, respectively. The accumulator-cold(ACC-Cold) system simulates ACC injection flow to the cold leg A and the accumulator-hot(ACC-Hot) system simulates ACC injection flow to the cold leg B. The water temperaturesof ACC-Cold and ACC-Hot tanks are the same and the ratio of ACC injection flowrate tocold leg A and cold leg B is 3:1. This injection method is adopted for good simulation ofACC injection flowrate to each clod leg in the LSTF.

The break point was located in the B-loop (loop without a pressurizer) cold leg betweenthe reactor coolant pump and the reactor pressure vessel. The break orientation washorizontal. The break assembly and break orifice are shown schematically in Figure 4.1-3and Figure 4.1-4, respectively.

The chronology of events for the SB-CL-18 is shown in Table 4.1-6. The experiment wasinitiated by opening the break valve at time zero. The reactor scram signal was sent at apressurizer pressure of 12.97MPa, 10 seconds after the break initiation, and this signalclosed the turbine throttle valve. The turbine bypass system was inactive due to theassumption of loss-of-offsite power occurring concurrently with scram. The loss of offsitepower terminated the main feedwater, and also tripped the reactor coolant pumps to



initiate coastdown. The reactor coolant pumps completely stopped at about 265 secondsafter the break.

The safety injection signal was sent at a pressurizer pressure of 12.27MPa, about 12seconds after the break initiation. However, the high pressure charging and high pressuresafety injection systems were not activated because of the failure assumptions. Thesecondary pressure increased after the closure of the turbine throttle valve, but wasmaintained at approximately 8MPa due to the SG relief valve operation.

The core was temporarily uncovered from 120 to 155 seconds, and the heater rods inmost of the core experienced superheating of up to about 190K. This temporary coreuncovery occurred during loop seal clearing. The core liquid level was depressedconcurrently with the level drop in the cross-over leg downflow sides. The core level dropwas amplified by the manometric effect caused by an asymmetric coolant holdup in theSG upflow and downflow sides. At about 140 seconds after the break, loop seal clearingoccurred in both loops and the core liquid level recovered rapidly. After the loop sealscleared, the break flow changed from low quality to high quality two-phase flow, and thedepressurization of the primary loop was accelerated. By about 180 seconds after thebreak initiation, the primary loop pressure decreased below the SG secondary sidepressure. Thereafter, the steam generators no longer served as heat sinks and the energyremoval from the primary system was through the discharge of coolant from the break. Itis noted that the loop seal clearing occurred before the reversal in primary and secondarypressures.

The core was uncovered again around 420 seconds due to vessel coolant boil-off, and theheater rods in the upper part of the core showed superheating of up to about 80K. Due todepressurization of the primary system, the accumulators were automatically actuated at455 seconds to fill the system with the emergency core cooling (ECC) water. The corewas covered with two-phase mixture again after about 540 seconds by the ACC waterinjection. The peak cladding temperature in the test was approximately 740K, observedduring the temporary core uncovery just before the loop seal clearing.



The M-RELAP5 noding diagrams for the LSTF uses a similar amount of detail in thevessel, steam generators, and loops as used in the actual PWR plant model. Figure 4.1-5shows the M-RELAP5 noding diagram of a cold leg break LOCA for the LSTF. Figure4.1-6 through Figure 4.1-9 show the M-RELAP5 noding diagrams of the LSTF pressurevessel, hot leg, steam generator (SG), cross-over leg and cold leg, respectively.



I

I

Core power curve (Figure 4.1-10), pump coastdown data (Figure 4.1-11 and Figure4.1-12), SG secondary pressure (Figure 4.1-13 and Figure 4.1-14), feedwater stop timing,and main steam line valve close timing uses time table boundary condition data madefrom experimental data.

To ensure the calculated break flowrate equal to test data, critical flow model was notused and velocity boundary condition was used. [

] Break flow velocity iscalculated from break mass flowrate and M-RELAP5 calculation result of two-phasemixture density at break nozzle. The comparison of break flow in test data and M-RELAP5calculation is shown in Figure 4.1-15.

[

I




Table 4.1-2 summarizes initial conditions before break. 2000-second steady-statesimulation was performed. At the end of this 2000-second simulation, predicted andmeasured flow parameters were compared to ensure reasonably good agreement by themodel.

Transient behaviors predicted by M-RELAP5 are shown and compared with themeasurements in Figure 4.1-16 through Figure 4.1-42.

(a) Blowdown PhaseThe rapid decrease in the RCS pressure following the break initiation is well simulatedby M-RELAP5 (Figure 4.1-16). Similarly, the loop flowrates of M-RELAP5 agree withthe measurements (Figure 4.1-18 and Figure 4.1-19). In the present analysis, thebreak flowrate and the SG secondary side pressure are given as boundary conditionsso as to exclude uncertainties due to these parameters. In addition, the break flowrateduring the latter portion of transient, such as during the boil-off phase, are adjusted toobtain a better agreement in the RCS pressure. However, no artificial calibration istaken into account for the blowdown phase. Therefore, it is concluded that M-RELAP5is capable of predicting the blowdown behaviors observed in the SB-CL-18 testaccurately.

(b) Natural Circulation PhaseAfter blowdown period, the core heat was removed by the natural circulation flow,which is driven by the heat transfer between SG primary and secondary side. Theliquid velocity at the top of SG U-tube stopped at about 95 sec, and then the liquidcondensed in the U-tubes is accumulated in the SG inlet plenum and SG U-tube uphillside because of the CCFL phenomena. Both loops flowrate at crossover leg ofM-RELAP5 calculation agree with these of the test data during natural circulationperiod (Figure 4.1-18 and Figure 4.1-19). As a result, M-RELAP5 capability to predictthe natural circulation behavior is good. It is noted that the M-RELAP5 ability to predictthe SG heat transfer will be independently assessed by using the analysis modelwhere the mechanical motions of secondary system valves are explicitly simulated, asdiscussed in Section 4.1.3.

(c) Loop Seal Clearance PhaseBreak flowrate of M-RELAP5 calculation is adjusted to test data (Figure 4.1-15), as aresult, primary pressure drop behavior agrees with test data excellently (Figure 4.1-16).Signal timings agree with test data (Table 4.1-6). Secondary pressures are alsoadjusted to test data (Figure 4.1-13 and Figure 4.1-14). Primary pressure andsecondary pressures of M-RELAP5 calculation agree with test data, as a result,M-RELAP5 capability to predict SG primary and secondary heat transfer is good.



The timing and depth of core level drop caused by loop seal of M-RELAP5 resultagree with test data (Figure 4.1-17). The timing of loop seal clearing of M-RELAP5result also agrees with test data excellently. After loop seal clearing, core water levelrecovered to 15kPa, equivalent to test data. As a result, M-RELAP5 can predict thephenomena of core mixture level and loop seal formation and clearance.

The downcomer water level of M-RELAP5 result agrees with that of test data at about150 sec (Figure 4.1-34). As a result, M-RELAP5 capability to predict downcomermixture level is good during loop seal period. Similarly, the water levels in the SGU-tube downhill side and crossover leg downhill side by M-RELAP5 agree with theseof test data (Figure 4.1-28 through Figure 4.1-31). The timing when crossover legdownhill side water level dropped to bottom (Figure 4.1-30 and Figure 4.1-31), corewater level began to recover (Figure 4.1-17), and break flow switched from two-phasemixture flow to single phase vapor flow (Figure 4.1-15). As a result, M-RELAP5 canpredict loop seal formation and clearance phenomena.

Differential pressure from hot leg to SG U-tube top of M-RELAP5 calculation resultagrees with that of test data (Figure 4.1-20 and Figure 4.1-21). As a result M-RELAP5can predict important phenomena during loop seal period. To check further detailsabout water held up in hot leg, SG inlet plenum and SG U-tube uphill side, theirdifferential pressures of M-RELAP5 calculation are shown in Figure 4.1-22 throughFigure 4.1-27, respectively. These figures present that water was held up in both SGU-tube uphill side and SG inlet plenum. In the test, the amount of water held up in 141SG U-tubes were different each other because of multi-dimensional effect (Figure4.1-26 and Figure 4.1-27). In M-RELAP5, water is easily held up in SG U-tube,because SG U-tube is modeled with one pipe. As a result, M-RELAP5 capability topredict water hold up in SG primary side and condensation drainage to inlet plenum isgood.

Core collapsed water level drop agree with test data at the timing of loop sealformulation (Figure 4.1-17). Regarding the core fuel cladding heat-up, the surfacetemperature of high-power rod heated up at 124 seconds after break in the test(Figure 4.1-37). In the M-RELAP5 result, heater surface temperature heated up at 134seconds after the break initiation (Figure 4.1-38). As a result, M-RELAP5 capability topredict core dryout phenomenon is a little conservative in hot assembly.

In the test data, the heater surface temperature reached 739 K at 147 seconds andthen cooled down (Figure 4.1-37 and Figure 4.1-42), so that this is the maximum rodsurface temperature during loop seal period. In the M-RELAP5 result, the heatersurface temperature reached 730 K at 158 seconds and then cooled down (Figure4.1-38 and Figure 4.1-41), so that this is the maximum rod surface temperature duringloop seal period. As a result, M-RELAP5 capability to predict Post-CHF heat transfer isgood in hot assembly.

The heater surface temperature reached saturated temperature and then all of rodswere rewetted before 155 seconds in the test (Figure 4.1-37). In the M-RELAP5 result,the heater surface temperature reached saturated temperature and then rewetted at212 seconds after break (Figure 4.1-38)' As a result, M-RELAP5 capability to predictrewet is conservative.

As a result, M-RELAP5 can predict core dryout, Post-CHF heat transfer, and rewet



during loop seal formulation and clearing period.

(d) Boil-off PhasePressure drop to equal to data of test data, break flowrate is adjusted when the breakflow become to single-phase vapor flow (Figure 4.1-15 and Figure 4.1-16 and 16).Core water level was dropping at about 300 seconds in M-RELAP5, while core waterlevel did not drop until 400 seconds in test data (Figure 4.1-17). Downcomer waterlevel drop timing is also earlier than the test data (Figure 4.1-34). As a result, the watermass inventory in the vessel of M-RELAP5 was smaller than that of the test data.

Loop seal clearing was occurred in both loops in M-RELAP5 calculation (Figure 4.1-18and Figure 4.1-19). But much liquid remain in the cross-over leg uphill side (Figure4.1-32 and Figure 4.1-33). It is insufficient to predict the liquid distribution after loopseal clearing and the core liquid level of M-RELAP5 is conservative.

Heater rod surface temperature heat-up timing is earlier than test data (Figure 4.1-37and Figure 4.1-38) because onset of core collapsed liquid level depression is earlierthan test data (Figure 4.1-17). In M-RELAP5, the heater rod surface temperatureheated up when core collapsed liquid level was about 15kPa, while the heater rodsurface temperature heated up when core collapsed liquid level was about 16kPa inthe test. This difference in pressure drop is negligibly small, after converting to thecollapsed liquid level and comparing with the node height employed for the corenodalization.

In the test data, the surface temperature of high-power rod reached 620 K at 497seconds and then cooled down, so that this is the maximum rod surface temperatureduring boil off period (Figure 4.1-38). In the M-RELAP5 result, the heater surfacetemperature reached 761 K at 472 seconds and then cooled down (Figure 4.1-38). InM-RELAP5, peak temperature was higher because core collapsed water level waslower (Figure 4.1-17). It indicates M-RELAP5 results are conservative enough.

Heat-up rate in M-RELAP5 was greater than that in test data (Figure 4.1-37 andFigure 4.1-38). The reason is that core power was higher because of the early corewater level drop timing. M-RELAP5 capability to predict Post-CHF heat transfer isgood.

(e) Core Recovery PhaseACC flow was initiated at 455 seconds (test data) or 445 seconds (M-RELAP5), andthen core water level started to recover (Figure 4.1-35 and Figure 4.1-36). Anddowncomer water level also started to recover. In test data, heater rod surfacetemperature reached saturated temperature and then rewetted at 538 seconds (Figure4.1-37). In M-RELAP5 result, heater rod surface temperature reached saturatedtemperature and then rewetted at 554 seconds (Figure 4.1-38). The difference of corecollapsed liquid level between M-RELAP5 and test data is very small (Figure 4.1-17),so that the final rewet time of M-RELAP5 agrees with test data.

The temperature behavior of heater rod depends on the core liquid level profile.M-RELAP5 can validly predict these phenomena as shown above. It is concluded thatM-RELAP5 predicts the rewet phenomena conservatively.

As described above, the base case calculation is slightly insufficient to predict the dryout



phenomena at the upper portion (above about 2.45m) of the heater rod during the loopseal period. This discrepancy is due to the nominal analysis condition of M-RELAP5 whichpurpose is to estimate the average thermal hydraulic behavior. There are some heaterrods heated up at such an upper portion in the experimental data. It indicates that theaccumulated water on the upper plenum region partly flow down to the core region. This isconsistent with the view that a spatially non-uniform liquid distribution exists in the coreregion. Such liquid distribution effects are already modeled as CHF multiplier inM-RELAP5. But, this base case result indicates that the CHF multiplier is not enough forROSA-IV/LSTF SBLOCA analysis.

It is concluded that M-RELAP5 under the nominal conditions can simulate the overallhydraulic behavior during the loop seal period very well and can predict the heat-upbehavior of the average heater rod due to core uncovery, and M-RELAP5 under theconservative conditions for dryout can also predict the heat-up behavior at the upperportion.

4.1.3 Steam Generator Heat Transfer Validation

In the M-RELAP5 analysis model examined in Section 4.1.2, the SG secondary sidepressure is imposed as a boundary condition so as to preclude an uncertainty due to thesecondary system behavior. On the other hand, the SG U-tube heat transfer affects thepressure behaviors both for the primary and secondary systems. This means that thecode assessment in the preceded section is insufficient to validate the SG heat transfermodel. Therefore, an additional code validation for the SG heat transfer is performedusing the SB-CL-18 test data in the following.


In the present calculation, mechanical motions of the main steam isolation valves (turbinethrottle valves) and the relief valves in the SG secondary system are explicitly modeled byusing the VALVE component in M-RELAP5. The main steam isolation valve is closedwhen the reactor scram signal occurs, and then the pressure control system regulates theSG secondary system pressure by opening and closing the relief valves.

The noding scheme for the ROSA-IV/LSTF test facility used in the present calculation isjust identical to that used in the preceded M-RELAP5 code calculations.




Dynamic behaviors in terms of the SG secondary side pressures, primary systempressure are compared between calculations and measurements as shown in Figure4.1-49 through Figure 4.1-51. The figures also contain the results obtained with the basemodel, where the secondary system pressure was given as a boundary condition. Theresults show that the model mechanistically simulating the SG secondary system is ableto reproduce the measured SG secondary side pressures with reasonable accuracy, andthere is no significant difference in the predicted primary system pressure compared withthe base model. The predicted integral of SG outlet steam mass agrees with themeasurement within 10% as shown in Figure 4.1-52. Furthermore, the calculated coredifferential pressure is comparable with that by the base model (Figure 4.1-53), both ofwhich predict the core liquid level depression conservatively in comparison with themeasurement.

4.1.4 Summary

M-RELAP5 predicted 5 period phenomena specific to SBLOCA. In addition, M-RELAP5predicted excellently the following important parameters: water hold up in SG primary side,condensation drainage to inlet plenum, SG primary and secondary heat transfer, waterlevel in SG outlet piping, and loop seal formation and clearance. And M-RELAP5predicted the following important parameters excellently in the loop seal period andconservatively in the boil off period: core dryout, Post-CHF heat transfer, rewet, coremixture level, and downcomer mixture level. Particularly, the loop seal in both loops aregradually cleared in M-RELAP5. It makes the results conservative.

Consequently, it is concluded that M-RELAP5 predicts conservatively SBLOCA from theresult of ROSA-IV SB-CL-18 simulation.



Table 4.1-1 Primary Test Facility Specification for ROSA-IV/LSTF

Parameters/Components US-APWR ROSA-ITV Ratio1 LSTFPrimary volume (i) 6.000Initial pressurizer pressure (MPa) 15.5 15.5 1.0Initial hot leg temp (K) 598.15 599 0.999

Initial cold leg temp (K) 561.25 A:563,B:564 A:0.997,3B:0 995

Initial RCS flowrate (kg/s) 48.7

Initial core bypass flowrate (kg/s) _ N/AInitial core power (MW) 4451 10 445.1

Reactor VesselInside diameter (m) 0.640Core height (m) 3.66Lower plenum max height (m) 2.361

avg. height from volume (m) 1.901Upper plenum height (m) 2.126Upper head max height (m) 2.126

avg. height from volume (m) 1.585Downcomer gap (m) 0.053

0.1134Core heated flow area (mn2) (below spacer) _Core bypass flow area (M

2) N/A

Downcomer flow area (m2 ) 0.09774

Hot LegsInner diameter (m) 0.207 +Length (m) 3.686in

Cold Legs

Inner diameter (m) 0.207Length (m) 3.438



Table 4.1-2 Steady-State Parameters for ROSA-IV/LSTF SB-CL-18

Parameter Target Predicted

Pressurizer pressure (MPa) 15.5 15.48

Hot leg fluid temperature (K) 599 / 599 599.2 / 599.2

Cold leg fluid temperature (K) 563 / 564 563.7 / 563.7

Core power (MW) 10.0 10.0

Core inlet flowrate (kg/s) 48.7 48.7

Pressurizer water level (m) 2.7 2.74

Pump speed (rpm) 769 /796 769 /796

Steam generator secondary pressure (MPa) 7.3 / 7.4 7.31 / 7.31

Steam generator secondary level (m) 10.8/10.6 10.6/10.6

Steam generator feedwater temperature (K) 494 494

Steam generator feedwater flowrate (kg/s) 2.6-2.8 2.7 / 2.8

Steam generator steam flowrate (kg/s) 2.6-2.8 2.7 / 2.8

Table 4.1-3 Operational Setpoints for ROSA-IV/LSTF SB-CL-18

Event Setpoint

Reactor scram signal (MPa) 12.97

Initiation of RCP coastdown With reactor scram

Safety injection signal (MPa) 12.27

High pressure charging not actuated

Safety injection not actuated

Accumulator injection (MPa) 4.51

Low pressure injection (MPa) 1.29

Main feedwater termination With reactor scram

Turbine throttle valve closure With reactor scram

Auxiliary feedwater initiation not actuated



Table 4.1-4 ECCS Conditions for ROSA-IVILSTF SB-CL-18

ECCS Specification

High pressure charging systemPump shut-off headDelay time from SI signal not actuatedFlowrateFluid temperatureInjection location (ratio)

High pressure injection systemPump shut-off headDelay time from SI signal not actuatedFlowrateFluid temperatureInjection location (ratio)

Low pressure injection systemPump shut-off head 1.29MPaDelay time from SI signal 17 sFlowrate scaled full capacityFluid temperature 310 KInjection location (ratio) CLA, CLB (3:1)

Acc systemPressure setpoint 4.51 MPaWater temperature 320 KInjection location (ratio) CLA, CLB (3:1)Initial tank level

to loop-A : ACC-Cold 5.76 mto loop-B : ACC-Hot 6.43 m

Terminal tank levelto loop-A : ACC-Cold 3.38 mto loop-B : ACC-Hot 5.64 m



Table 4.1-5 Core Power Decay Curve for ROSA-IV/LSTF

Time Power Time Power(sec) (MW) (sec) (MW)

0.000 10.000 100.000 5.2001.000 10.000 150.000 3.6322.000 10.000 200.000 2.8483.000 10.000 400.000 1.7764.000 10.000 600.000 1.5685.000 10.000 800.000 1.4886.000 10.000 1000.000 1.4247.000 10.000 1500.000 1.2808.000 10.000 2000.000 1.200

10.000 10.000 4000.000 .99215.000 10.000 6000.000 .84820.000 10.000 7980.000 .78429.000 10.000 10020.000 .78440.000 8.912 19980.000 .59260.000 7.344 60000.000 .46480.000 6.128 100020.000 .368



Table 4.1-6 Primary Test Chronology for ROSA-IV/LSTF SB-CL-18

Event Experiment M-RELAP5

Break (s) 0 0

Reactor trip (s) -10 13.0

Safety injection signal (s) -12 16.3

Main steam line valve close (s) 14 14

SG feedwater stop (s) 16 16

High pressure charging injection (s)

High pressure safety injection (s)

Auxiliary Feedwater ON (s)

First Core Uncovery (s) 120-155 95-175

Loop Seal Clearing (Loop A/B) (s) ~140 130

Primary / Secondary Pressure Reversal (s) ~180 172

Reactor Coolant Pumps (PC-A/B) stop (s) 265 265

Second Core Uncovery (s) 420 - 540 300 - 540

Accumulator Injection ON (s) 5



1 SfrQtd. I¢•.19 2 Io99= r, p..$wre -asset

U JS pepn

11. Accol. 1.ato .

SA.CCplF•• Ig11 P511 Piping

I SHR pipingI Pretsrii control Vent vil.e

IS Break loItlatl .-Salve

1 reled s.te r 11.0

zO ie~d wsatr pSumZ1 At codenserL

2 Cooling tower lineZ4 Steen generator Lobe

Break flaw Storage tik

Figure 4.1-1 General Structure of ROSA-IV/LSTF Facility4 '



___1 ~Py qh~.11 fu l . 46L

Central Red Mold. teloi lenp

Care Sopp,1 Plorte U4,6170.2

110t. Cold LVX EL4•3S2.8

Core IPlate 1:03S9U

- Se..... .54 ......Boe. It- 13864.5

511~m 19 EV17•+9|Specer On nnpao

.Spctr 07 '102293

Spacer t6 ZL-23•8

$041-1 1% Tftg~2f)

1.... .Sp.. r 5.) U#1it)

Spacer 12 EI-S24

L+m Me= p~acer 15 EL-?LtO

- e r5 . - L -I

- 0-•- to-t:;.'

- -L Sp.ce, 12 li-1701

Looer Peav.Spoce It 11-2031

=jF_

HLI h'te PP ottcoo CL-23613/

I j ra.re bottoam refers to the bot tom atJ heot*d care.

Figure 4.1-2 Pressure Vessel Assembly of ROSA-IV/LSTF Facility4 1



BREAK 'UNIT

DV.- 64'.53f

Do - 22 .

Figure 4.1-3 Break Assembly of ROSA-IV/LSTF Facility4 "1

24.0 ;c

-" x

4 So

7e. FIow

Figure 4.1-4 Break Orifice of ROSA-IV/LSTF Facility4 1


Cn

(D

(-

C"

I-

-4

01

Cn"0

3CD

0)

0)

:3

--b

C)

C,)0<

0 C

MO-

K JFigure 4.1-5 M-RELAP5 Noding Scheme for ROSA-IV/LSTF SBLOCA Test Analysis


Figure 4.1-6 M-RELAP5 Vessel Noding for ROSA-IV/LSTF SBLOCA Test Analysis



Figure 4.1-7 M-RELAP5 Hot Leg Noding for ROSA-IV/LSTF SBLOCA Test Analysis



Figure 4.1-8 M-RELAP5 Steam Generator Noding for ROSA-IV/LSTF SBLOCA TestAnalysis



Figure 4.1-9 M-RELAP5 Crossover Leg Noding for ROSA-IV/LSTF SBLOCA TestAnalysis



TIME (sec)

Figure 4.1-10 Total Core Power for ROSA-IV/LSTF SB-CL-18

E

600TIME (sec)

Figure 4.1-11 RCP Rotation Speed in Primary Loop-A for ROSA-IVILSTF SB-CL-18



1500

1000

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

---- -------------------

---------- --------------

------------------ -----------

------------------ ---------------

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

M-RELAPS(BASE)I---TEST DATA

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

E

500,L

0

-50CI i _______ L1 i______ I ______ i _______ _______0 100 200 300 400 500 600

TIME (sec)

Figure 4.1-12 RCP Rotation Speed in Primary Loop-B for ROSA-IVILSTF SB-CL-18

0 . r-____________ ____________-____________ _________________________

8

6

4

2

------------------------- ------------------------- -------------------------

-------------------- ------------------------- ------------------------- --------------------------

- ------------------- -- ------------------------- ---------------------------------------------------

I . M-RELAP5 (BASETEST DATA

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

~6oItl I Ii ....

0 100 200 300 400 500

TIME (sec)

Figure 4.1-13 SG-A Steam Dome Pressure for ROSA-IVILSTF SB-CL-1 8



BM-RELAP5(BE)TEST DATA

-------------------- -------------------------- --------L----------------. . . .. . .. . . .. . .. . .. . . .. . .. . .. . . . . . .. . .. . . .. . .. . . . . . .. . .. . .. . .

100 200 300

TIME (sec)

400 500 6110

Figure 4.1-14 SG-B Steam Dome Pressure for ROSA-IV/LSTF SB-CL-18

45

M-RELAP5 (BASE)TEST DATA I

3 0 .3 ... ..... ------. -------------------------. -------------------------.-. --------------------------.I.-------------------------

300TIME (sec)

Figure 4.1-15 Break Flowrate for ROSA-IV/LSTF SB-CL-18



I M-RELAP5(BASE)TEST DATA

15

'O----------------- ------------------------- 1 ------------------------- t -------------------

---------- ----------------

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

'0 100 200 300 400 500 600TIME (sec)

Figure 4.1-16 Pressurizer Pressure for ROSA-IV/LSTF SB-CL-18

TIME (sec)

Figure 4.1-17 Core Differential Pressure for ROSA-IVILSTF SB-CL-18



0C,

bfl

TIME (sec)

Figure 4.1-18 Loop-A Crossover Leg Flowrate for ROSA-IV/LSTF SB-CL-18

0CD

bZ

300TIME (sec)

600

Figure 4.1-19 Loop-B Crossover Leg Flowrate for ROSA-IV/LSTF SB-CL-18



M

TIME (sec)

Figure 4.1-20 Loop-A Hot Leg to U-Tube Top Differential Pressure for ROSA-IV/LSTFSB-CL 18

100

SM-RELAP5(BASE)TEST DATA

40

-20C L ______ _____ i ______I______0100 200 300 400 500 600

TIME (sec)

Figure 4.1-21 Loop-B Hot Leg to U-Tube Top Differential Pressure for ROSA-I VILSTFSB-CL-18



TIME (sec)

Figure 4.1-22 Loop-A Hot Leg to SG Inlet Plenum Bottom Differential Pressure forROSA-IV/LSTF SB-CL-18

•A1if', - _______________ - ________________ - ____________________________________________________

8

6

M-RELAP5 (BASE)TEST DATA

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

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

M~

4

400 500 600

TIME (sec)

Figure 4.1-23 Loop-B Hot Leg to SG Inlet Plenum Bottom Differential Pressure forROSA-IV/LSTF SB-CL-18



M~

TIME (sec)

Figure 4.1-24 Loop-A SG Inlet Plenum Differential Pressure for ROSA-IV/LSTFSB-CL-1 8

w~

600

TIME (sec)

Figure 4.1-25 Loop-B SG Inlet Plenum Differential Pressure for ROSA-IV/LSTFSB-CL-1 8



w

TIME (sec)

Figure 4.1-26 Loop-A SG U-Tube Uphill Side Differential Pressure for ROSA-IV/LSTFSB-CL-1 8

100

M-REL (BASE)TEST DATA

60

420 iE

80 .................. 4 ; ---------................. -- ......._ ...____ ____....-- - - -___. . .- -- ____-- - - - - - - -

00 100 200 300 400 500 600

TIME (sec)

Figure 4.1-27 Loop-B SG U-Tube Uphill Side Differential Pressure for ROSA-IV/LSTFSB-CL-18



300TIME (sec)

Figure 4.1-28 Loop-A SG U-Tube Downhill Side Differential Pressure forROSA-IV/LSTF SB-CL-18

0~

600

TIME (sec)

Figure 4.1-29 Loop-B SG U-Tube Downhill Side DifferentialROSA-IV/LSTF SB-CL-18

Pressure for



M2

TIME (sec)

Figure 4.1-30 Loop-A Crossover Leg Downhill Side Differential Pressure forROSA-IV/LSTF SB-CL-1 8

ca

300

TIME (sec)

600

Figure 4.1-31 Loop-B Crossover Leg Downhill Side Differential Pressure forROSA-IVILSTF SB-CL-18



#_

TIME (sec)

Figure 4.1-32 Loop-A Crossover Leg Uphill Side Differential Pressure forROSA-IV/LSTF SB-CL-1 8

40

M-REL (BASE)TEST DATA

3 0 - -- ----------- -- -- -- --- -- -- -- --- -- -- -- --- -- ---. -- --------- -------------.. .. . .. . .. . .. . .. . .. . .. . .. . -- -----------------------------.. .. ... .. .. ... .. ..

00 20L 00 30 40 0 0

T0. . . (sec)0 100 200 300 400 500 600

Figure 4.1-33 Loop-B Crossover Leg Uphill Side Differential Pressure forROSA-IV/LSTF SB-CL-18



co

TIME (sec)

Figure 4.1-34 Downcomer Differential Pressure for ROSA-IV/LSTF SB-CL-18

M-RELAP5(BASE)i. TEST DATA _ _ _

6 ............ -................................................................................................... --------------

________________I _____ L ________

- 100 200 300 400 500 600

TIME (sec)

Figure 4.1-35 Loop-A Accumulator Injection Flowrate for ROSA-IV/LSTF SB-CL-18



00,

TIME (sec)

Figure 4.1-36 Loop-B Accumulator Injection Flowrate for ROSA-IVILSTF SB-CL-18

600

TIME (sec)

Figure 4.1-37 Heater Rod Surface Temperature at 3.61m (Test Data) and at 3.57m(M-RELAP5) for ROSA-IV/LSTF SB-CL-18



TIME (sec)


TIME (sec)




MOUU- ___________-____________-_____________________________________

900 -

800 -

-- -- --- -- --- -- ---- --- ---- -- ----- --- ---- -- --- -- --- -

------------- ----------------------------------------700

M-RELAP5 (SEI---TEST DATA I

---- ---- - --- -600 -

500 -

-0- 100 200 300

TIME (sec)

400 500 600


Figure 4.1-43 Heater Rod Surface Temperature at 3.61m (Test Data) and at 3.57m(M-RELAP5) for ROSA-I V/LSTF SB-CL-18



Figure 4.1-44 Heater Rod Surface Temperature at 3.05m (Test Data) and at 3.17m(M-RELAP5) for ROSA-IVILSTF SB-CL-18



6AS-1 E-UAP-1 00001 (R0)Non-Proprietary VersionM-RELAP5 Supplementary Manual Vol. III


Figure 4.1-47 Heater Rod Surface Temperature at 1.02m (Test Data) and at 1.11m(M-RELAP5) for ROSA-I V/LSTF SB-CL-18

-1)




10

8 =

¢0

0.

M.

CF

M-RELAP5(Base Case)M-RELAP5 (MSRV Mode I)TEST DATA

------- - -- --- --- --- -- - --- -- ---- --

4

2ý

_0 100 200 300 400 500 600Time (sec)

Figure 4.1-49 Broken Loop Secondary Pressure for ROSA-IV/LSTF SB-CL-18 byExplicit Secondary System Model



10

8

c-

C,_

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oF---------------------

---- ------------ ----- ----------

------------------------ ----------------

M-RELAP5(Base Case)- M-RELAP5(MSRV Model)----- TEST DATA

.. . .. . . . . . . . . . . . . . . . . . . . .......... o. ..................... -- -- --- -- -------------------..

4

2F

10 100 200 300Time (sec)

400 500 600

Figure 4.1-50 Intact Loop Secondary Pressure for ROSA-IV/LSTF SB-CL-18 byExplicit Secondary System Model

20

-_M-RELAP5(Base Case) IM-RELAP5(MSRV Model)I

----- TEST DATA

lE_

1-

o0 100 200 360 400 500 600Time (sec)

Figure 4.1-51 Primary System (Pressurizer) Pressure for ROSA-IV/LSTF SB-CL-18by Explicit Secondary System Model



-Test DataM-RELAP5 (MSRV Model)600

41 500

0 400U-

E 300E

CDC

2004--0

100a)4-'

0

I I i i i i

- - --- -- - -!- - - - - - - - - - - - - - - - - - - -

- - - - - - - - - - - - - - -

0 20 40 60 80 100 120Time (sec)

140 160 180 200

Figure 4.1-52 Integral of SG Secondary Outlet Steam Mass for ROSA-IV/LSTFSB-CL-18 by Explicit Secondary System Model

40 -

-M-RELAP5 (Base Case)M-RELAP5(MSRV Model) I

-- TEST DATAC 3C

z: 3 0 ... ......... ... ... ....... .. ... ... ...-.-.---.-.-- --- --- --

C,,

C,

-2CD

4-

0 100 200 300 400 500 600

Time (sec)

Figure 4.1-53 Core Differential Pressure for ROSA-IV/LSTF SB-CL-18 by ExplicitSecondary System Model




The SB-CL-09 test simulated a 10% cold leg break SBLOCA. This test analysis wasperformed to clarify the capability of M-RELAP5 to predict the behavior during SBLOCAswith break sizes larger than that of the SB-CL-1 8 (5% break).



The content is the same as that in Section 4.1 1.1.


Most of the parameters at the initial condition of the LSTF 10% cold leg break test,SB-CL-09 8, are almost identical to the specified in Table 4.1-2. Major operational setpointsand conditions including the actuation logic of the emergency core cooling system (ECCS)for this test are summarized in Table 4.1-3 and Table 4.1-4, respectively.

The chronology of events observed in Run SB-CL-09 is shown in Table 4.2-1. Theexperiment was initiated by opening the break valve at time zero. The reactor scramoccurred at a pressurizer pressure of 12.97 MPa and the turbine throttle valve was closed.The turbine bypass system was unavailable due to the assumption of loss-of-offsite power.Also at the scram, the main feedwater was terminated and the reactor coolant pumpswere tripped to initiate coastdown. The safety injection signal occurred at a pressurizerpressure of 12.27MPa, however, no water was injected from the high pressure chargingsystem and HPIS.

The secondary pressure increased to and remained at approximately 8MPa due to the SGrelief valve operation. No auxiliary feedwater supplied to the steam generators. Thecoolant in both hot legs became saturated immediately following a break, and then thetwo-phase circulation started. The coolant in cold-legs A and B saturated at 49 and 55s,respectively. The two-phase circulation terminates in the loops A and B at 68 and 55 s,respectively. After the termination of the two-phase circulation, drainage started in theupflow-side and downflow-side SG U-tubes. Thereafter, the downflow side of thecrossover leg became empty and the loop seal clearing occurred in both loops. After theloop seal clearing, steam generated in the core reached the cold leg via the SG U-tubesand began to leave from the break hole. In the stage of the drainage in the SG U-tubesuntil the loop seal clearing, the drainage in the upflow-side SG U-tubes was slower than inthe downflow-side SG U-tubes, which elevated the hot-leg pressure higher than the coldleg pressure. Thus, the elevated pressure in the hot leg resulted in the core liquid leveldepression lower than the loop seal bottom elevation. This core liquid level depressioncaused the excursion of heater rod temperatures. After the loop seal clearing, the coreliquid level recovered, however, it was not complete and the core was still partiallyuncovered since some liquid remains in the upflow-side SG U-tubes and the SG inletplenum including the bending portion between the SG inlet plenum and the hot leg.Therefore, the heater rod temperature continued to increase. Finally, the mechanism toprotect the heater rods from physical damage was automatically actuated by tripping offthe power supply to the heater rod. After the power tripped-off, the core liquid levelrecovered and the heater rods were rewetted.





This section is same as Section 4.1.2.1.


The results of the M-RELAP5 analysis of SB-CL-09 are shown in Figure 4.2-1 to Figure4.2-13. The core uncovery associated with loop seal clearance starts about 60s after thebreak, and the loop seal clears at about 70s (Figure 4.2-1). After the loop seal clearance,the core water level did not fully recover (Figure 4.2-1). M-RELAP5 well simulates thetiming of core uncovery, loop seal clearance, and the subsequent core water levelincrease. At 111s, core power is tripped off, the core water level increases again, and thecore is fully quenched by about 180s (Figure 4.2-1, and Figure 4.2-5 to Figure 4.2-13).M-RELAP5 simulates the SB-CL-09 test qualitatively well. The core water level drop islarger than in the test at 50s, and core heat up starts because of CHF rather than atop-down dryout. There are no boil-off and subsequent core recovery periods because thecore power was tripped.

4.2.3 Summary

M-RELAP5 is capable of predicting the primary plant responses in the ROSA-IV/LSTFSB-CL-09 test (10% break). Transient evolutions faster than that in the SB-CL-18 test (5%break), are accurately reproduced by M-RELAP5 particularly for the core liquid levelbehavior. The important point is that the code predicts the heater rod heat-up behaviorsaccurately or conservatively in comparison with the measurements. Therefore, it isconcluded that M-RELAP5 is sufficiently applicable to the safety analysis for SBLOCAswith larger break sizes.




Event Experiment M-RELAP5

Break (s) 0 0

Reactor trip (s) 8 8.9

Main steam line valve closed (s) 11 11

Safety injection signal (s) 11 13.3

SG feedwater stop (s) 15 15

Core heat-up started (s) 67 44

Loop seal clearing 74 68

Primary/secondary pressure reversal 97 78

Core power tripped 111 111

Accumulator injection on (to Loops A & B) 195 182



CL

4-

4--

Time (sec)

Figure 4.2-1 Core Differential Pressure for ROSA-IV/LSTF SB-CL-09




Ic

U,U,

10-

Time (sec)

Figure 4.2-3 Primary Pressure for ROSA-IV/LSTF SB-CL-09

16I M-RELAP5

1 4 ~ ~ ~ ~ ~ ~ ~ ---- -- -- -- -- .. .... . ... ..... .... ... .... . ... .. .. .... . ... .... .. .....--" T E S T D A T M I 7 . .

10

12 ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~~~~~----- ----------------------------..................................... ..........................................

0 6 ............... ..... .------------ --------------------- .............................. ..... ------------------ ------------------

100Time (sec)

Figure 4.2-4 Core Power for ROSA-IV/LSTF SB-CL-09



1:

I

2

Ca,

100Time (sec)


1:

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41

Caw

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cc

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Time (sec)




12001

1000

-I-,

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M -RELAP5TEST DATA TW3291

...... ... ............................. -------- ................................................................800

6001

W I)0

Time (sec)

Figure 4.2-7 Heater Rod Surface Temperature at 2.64m (Test Data) and at 2.68m(M-RELAPS) for ROSA-IV/LSTF SB-CL-09

1200

M-RELAP5 TW328...TEST DATA T38

1000ý

.4-

OV

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------------------------------------- --- ---------- ------ - -- ------- 7 -

----------- -- --- - ----- ----------------- -------------

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600

40CL0 50 100

Time (sec)150 200




1200

M-RELAP5 TTEST DATA TW333

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. .............600

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50 100 150 200Time (sec)


1200

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M-RELAP5 T36...TEST DATA TW2

800

600

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150 200




1200r

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---- --- --- ---- ---- --- ------ --- --- --- ---- ---- -- -- ------- -- -- -- ----- -- -- -- -- --800

600

4000-50 100

Time (sec)150 200


1 20 0 1

1000

0

4-,

CLE

Ca

C-3

SM-RELAP5 T35TEST DATA TW325

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

800

600

a ll[ i50 100 150 200

Time (see)

Figure 4.2-12 Heater Rod Surface Temperature at O.61m (Test Data) and at 0.64m(M-RELAP5) for ROSA-IVILSTF SB-CL-09



12001

1000

2

E

-0

C_>

800

M-RELAP5... TEST DATA TW3551

... .. . ... .. ....... .... .. . ... ... ----.......... .... ............ ............ ..... .......... .......... ............ ........... . ........................600

/al I I

50 100Time (sec)

150 200





The SB-CL-12 simulated a 0.5% cold leg break SBLOCA. This test analysis wasperformed to clarify the capability of M-RELAP5 to predict the behavior during SBLOCAswith break sizes smaller than that of SB-CL-1 8 (5% break).



The content is the same as that in Section 4.1.1.1.


[

I



I I



The content is the same as that in Section 4.1.2.1.


[

4.3.3 Summary

The applicability of M-RELAP5 code to analyze SBLOCA transients with smaller breaksizes has been assessed using experimental test data obtained from the ROSA-IV/LSTFSB-CL-12 test (0.5% break). Transient evolutions slower than that of SB-CL-18 test (5%break) were observed in the present test, although the accident scenario and fundamentalplant behaviors are comparable between the two experiments. The code validation resultsshow that M-RELAP5 reproduces the plant behaviors, as also the case for the otherROSA SBLOCA tests, although the code tends to predict the loop seal formation andclearance, and the onset of boil-off phase earlier than the experiment. M-RELAP5provides conservative heater rod heat-up than the measurements. Thus, it is concludedthat M-RELAP5 is applicable for the SBLOCA analysis with smaller break sizes.



Figure 4.3-1 Core Differential Pressure for ROSA-IV/LSTF SB-CL-12




Figure 4.3-3 Primary Pressure for ROSA-IV/LSTF SB-CL-12

Figure 4.3-4 Core Power for ROSA-IV/LSTF SB-CL-12



Figure 4.3-5 Heater Rod Surface Temperature at 3.61m (Test Data) and at 3.57m(M-RELAP51 for ROSA-IV/LSTF SB-CL-12




4.4 LOFT L3-1 Test



The Loss-of-Fluid Test (LOFT) reactor system4-9 , in particular the primary coolant systemand reactor core, is a fully operational, scaled representation of a commercial pressurizedwater reactor (PWR). Details of the test facility scaling are given in Reference 4-10. Assuch, transients resulting from accident initiating events are representative in complexityand nature of those accidents which may occur in commercial PWRs. The experimentalassembly comprises five major subsystems which have been instrument such that systemvariables can be measured and recorded during the test. The subsystems include a) thereactor vessel, b) the intact loop, c) the broken loop, d) blowdown suppression system,and e) the emergency core cooling system (ECCS). The LOFT major components areshown in Figure 4.4-1.

The LOFT reactor vessel, which simulates the reactor vessel of a commercial PWR, hasan annular downcomer, a lower plenum, lower core support plates, a nuclear core, and anupper plenum. The downcomer is connected the cold legs of the intact and broken loopsand contains two instrument stalks. The upper plenum is connected the hot legs of theintact and broken loops. The core contains 1300 unpressurized nuclear fuel rods arrangedin five square (15x15 fuel assemblies) and four triangular fuel modules located at thecorner, shown in Figure 4.4-2. The fuel rods have an active length of 1.67-m and anoutside diameter of 10.72-mm. The fuel consists of U0 2 sintered pellets with an averageenrichment of 4.0 st% fissile uranium (U235) and with a density that is 93% of theoreticaldensity. Fuel pellet diameter and length are 9.29 and 15.24-mm, respectively. Both endsof the pellets are dished with the total dish volume equal to 2% of the pellet volume.Cladding material is Zircaloy-4. Cladding inside and outside diameters are 9.48 and10.72-mm, respectively. The details are given in Reference 4-11.

The intact loop simulates three loops of a commercial four-loop PWR and contains asteam generator (SG), two primary coolant pumps in parallel, a pressurizer, a venturi flowmeter, and connecting piping. The broken loop consists of a hot leg and a cold leg that areconnected to the reactor vessel and the blowdown suppression tank (BST) header. Eachleg consists of a break plane orifice, a quick-opening blowdown valve (QOBV), arecirculation line, an isolation valve, and connecting piping. The break for Experiment L3-1is located in the broken loop cold leg. The recirculation lines establish a small flow fromthe broken loop to the intact loop and are used to warm up the broken loop. The brokenloop hot leg also contains a simulated steam generator and simulated pump. Thesesimulators have hydraulic orifice plate assemblies which have similar resistances to flowas an active steam generator and a pump.

The blowdown suppression system is comprised of the BST header, the BST, the nitrogenpressurization system, and the BST spray system. The blowdown header is connected tothe suppression tank downcomers which extend inside the tank below the water level. Theheader is also directly connected to the BST vapor space to allow pressure equilibration.The nitrogen pressurization system is supplied by the LOFT inert gas system and uses aremote controlled pressure regulator to establish and maintain the specified BST initialpressure. The spray system consists of a centrifugal pump that discharges through a



heat-up exchanger and any of three spray headers or a pump recirculation line thatcontains a cool-down heat exchanger. The spray pump suction can be aligned to eitherthe BST or the borated water storage tank. The three spray headers have flowratecapacities of 1.3, 3.8 and 13.9 Us, respectively, and are located in the BST along theupper centerline.

T he LOFT ECCS simulates that of a commercial PWR, which consists of twoaccumulators, a high-pressure injection system (HPIS), and a low-pressure injectionsystem (LPIS). Each system is arranged to inject scaled flowrates of emergency corecoolant directly into the primary coolant system. The accumulator, HPIS, and LPIS wereused during the L3-1 test. Each system was arranged to inject scaled flowrates of ECCdirectly into the primary coolant system (RCS) cold leg. To provide these scaled flowrates,accumulator ACC-A, HPIS Pump A, and LPIS Pump A were utilized. Accumulator ACC-Awas preset to inject the ECC at a system pressure of 4.22 MPa. HPIS Pump A was set toinitiate injection at a system pressure of 13.16 MPa. The pressure setpoint for automaticLPIS injection was 0.98 MPa.

Details of the LOFT system are described in Reference 4-12.


Important results from the experiment are discussed in Reference 4-13 and summarizedbelow.

LOFT/L3-1 was the first nuclear powered SBLOCA experiment. The test was designed tosimulate a 4-in diameter equivalent (2.5%) single-ended break in the cold leg of a PWR.Coolant from the accumulator, HPIS and LPIS was injected into the intact loop cold leg.The reactor was scrammed manually at 2 seconds prior to the break initiation (defined tooccur at time zero) when the cold leg blowdown valve was opened. The pumps weretripped at the break initiation and coasted down in about 19 seconds. The HPIS flowinitiated automatically at about 5 seconds. The pressurizer was empty by 17 seconds andthe upper plenum fluid was saturated by 25 seconds.

Natural circulation began as the pumps completed their coastdown and continued until390 seconds when the primary system pressure dropped below the secondary pressureand the steam generator was no longer a heat sink. The break flow was sufficient,however, to remove the decay heat and to continue system depressurization. At about630 seconds the accumulator started injecting the ECC. The accumulator emptied ofwater and nitrogen entered the system at about 1750 seconds. The LPIS setpoint waspurposely lowered from a normal pressure of 2.12 MPa to 0.98 MPa to assure nitrogeninjection from the accumulator to the RCS. No effects of the nitrogen on the RCSresponse were observed in the measurements.

The pump inlet loop seal did not clear during the transient as expected because of thelarge core bypass paths from the upper plenum to the cold leg which allowed pressureequalization between the hot and cold legs.

At about 3600 seconds, secondary bleed and feed was initiated by the operator action,which imposed a 38.8 to 50 K/hr cool-down rate on the secondary system. This procedurehad no effect on the primary system pressure because the primary and secondary



systems were thermally decoupled.

The mass inventory in the reactor vessel was sufficient at all times to keep the corecompletely covered, consequently the core remained cooled with the clad temperaturesfollowing the coolant saturation temperature.



The LOFT/L3-1 test was simulated by M-RELAP5 such that the code ability to predict theSBLOCA test was examined as well as done for the other lET analyses, the ROSA/LSTFSB-CL-18 and the Semiscale/S-LH-1. The M-RELAP5 LOFT model used here is based onthe input model developed by INL4 13. However, the noding scheme and thethermal-hydraulic model options have been modified so as to conform to the modelsapplied to the US-APWR SBLOCA analysis. The noding diagram is shown in Figure4.4-3.

The M-RELAP5 LOFT model primarily consists of the a) reactor vessel, b) pressurizer, c)steam generator, d) intact loop, e) broken loop, f) ECCS, and g) break assembly. [

]

Heat conduction in the nuclear core fuel rods and the reactor component structures aretaken into account. [

The counter-current flow limitation (CCFL) occurring in the piping with a smaller diameteris taken into account for the calculation. The CCFL in the SG U-tubes is modeled usingthe Wallis correlation4-14 , where 3=0.0, c=0.88, and m=1.0 are applied. This modeling isidentical to that for the US-APWR plant calculation, because the geometric scaling of theSG U-tubes is almost identical between the LOFT and US-APWR. [



The post-test analysis report 4-13 states that the steam control valve of the SG secondarysystem did not seat 100% nor did it seat the same each closure although the valve beganto close at 5%/s during the transient test. The actual steam leakage from the secondarywas not measured directly. In the present calculation, therefore, the secondary systempressure is imposed as a boundary condition based on the measurement.

The break flow history is imposed as a boundary condition which is specified by the inputdata based on the measurement. The Moody critical flow model4- 18 has beenimplemented into M-RELAP5 for the plant safety analyses 4-16 , in conformance to therequirement prescribed in Appendix K to 10 CFR 50. The Moody critical flow model isknown as a model which maximizes the break flowrate. In the framework for theM-RELAP5 code assessment using the lET data, therefore, MHI practically employed anapproach to impose the measured break flowrate data as a boundary condition, excludingexcessive conservatism and distortions caused by applying the Moody critical flow model.It is noted that the experimental test report 4-9 mentions that the uncertainty for themeasured break flowrate was ±15%. [ ]

The core fission power and decay power history are also given through the input datatable for the present calculation. Although Reference 4-12 mentions that no significanteffects of noncondensable gas from the accumulator were observed after the accumulatoremptied, the noncondensable gas model simulating the nitrogen entering the RCS wasapplied in the present calculation.

The M-RELAP5 transient calculation simulated the experiment from the break initiationuntil shortly before the operators manually initiated the steam bleed of the secondarycoolant system (SCS). The latter portion of the experiment was not simulated because thebehavior of the LOFT facility after the onset of the steam bleed is not relevant to thebehavior of the US-APWR.


The steady-state calculation was performed by M-RELAP5. The converged plantparameters are listed in Table 4.4-1, in which the calculation results are compared withthe measurements. The table shows that M-RELAP5 accurately reproduces thesteady-state condition prior to the transient test for the LOFT/L3-1.

The chronology during the LOFT/L3-1 test is listed in Table 4.4-2, where the experimentaland calculated results are compared. The transient calculation was initiated by thesimulated break flow data shown in Figure 4.4-4. The measured secondary systempressure was also given by a boundary condition as shown in Figure 4.4-5. [

],agood agreement was obtained for the primary system pressure as shown in Figure 4.4-6.It is noted that the M-RELAP5 accuracy for the SG heat transfer has been validated usingthe ROSA-IV/LSTF SB-CL-18 test data. Following the break initiation, the RCS rapidlydecreases to the secondary system pressure during the blowdown phase. The temporalchange of pressurizer liquid level was well reproduced by M-RELAP5 as shown in Figure4.4-7. The natural circulation begins as the pumps complete their coastdown, and then theprimary and secondary pressures equivalently decrease. Around 400 seconds after thebreak initiation, the primary system pressure falls below the secondary system pressure,



which is the end of the natural circulation phase. After that, the SG no longer behaves asa heat sink.

Calculated differential pressures in terms of the crossover leg downhill-side and uphill-sideare compared with the measurements in Figure 4.4-8 and Figure 4.4-9, respectively. Thedifferential pressure is essentially due to the liquid level after the natural circulation periodends. In the experiment, the loop seal formed in the intact loop crossover leg was notcleared because the steam generated in the core was able to be vented through thebypass paths. Reference 4-12 describes that the core bypass fractions were 3.6% ofprimary loop flow for the lower plenum to upper plenum path, 6.6% for the inlet annulus(downcomer) to upper plenum path, and 1.3% for the reflood assist bypass valve at thetest initiation. It was also noted that the valve leakage area for the reflood assist bypasschanged with the pressure difference across the valve. Similar to the measurement, theM-RELAP5 calculation predicts that the loop seal in the intact loop crossover leg does notclear throughout the transient as shown in Figure 4.4-8 and Figure 4.4-9.

The accumulator started injecting the safety coolant as the RCS pressure fell below theinitial accumulator pressure around 640 seconds. The nitrogen gas in the accumulatortank expands and ejects the safety coolant to the RCS. The accumulator emptied of thewater and the nitrogen began to enter the RCS at about 1750 seconds. These behaviorsare well simulated in the M-RELAP5 calculation as shown in Figure 4.4-10 for the tankpressure, and in Figure 4.4-11 for the tank level, respectively. This validates theaccumulator model implemented in M-RELAP5.

No fuel cladding heat-up was observed in the LOFT/L3-1 test or calculated withM-RELAP5 as shown in Figure 4.4-12.

4.4.3 Summary

The LOFT/L3-1 experiment was simulated by using M-RELAP5 so as to validate the codeability to predict the plant response occurring under SBLOCAs. The LOFT facility is notexactly scalable to the US-APWR from the viewpoint of the fuel rod array, reactor height,and loop configuration, however, the experiment is useful to assess the M-RELAP5models and noding scheme which are applied to the plant analysis.

The break flowrate was optimized to obtain a better agreement in terms of the RCSpressure behavior. In plant safety analyses, the uncertainty in the break flowrate isaddressed by performing the break size spectrum calculations. M-RELAP5 showed thatthe code is capable of reproducing primary plant responses, particularly for the loop sealbehavior observed in the measurement. Both for the measurement and calculation, theloop seal formed in the crossover leg of the intact loop was not cleared throughout thetransient, since a large core bypass flow paths allowed generated steam to vent from thecore without the loop seal clearance. In addition, M-RELAP5 predicted no claddingheat-up as same as observed in the measurement.

Hence, it can be concluded that M-RELAP5 is able to reproduce the transient behavior,phenomena and processes of interest during the LOFT/L3-1 SBLOCA test.



Table 4.4-1 Steady-State Parameters for LOFT/L3-1

Parameter Experiment 9'13 M-RELAP5

Primary system pressure [MPa] 14.81 ± 0.04 14.82Primary system mass flowrate [kg/s] 484.0 ± 6.3 484.0

Cold leg temperature [K] 554.0 ± 3 554.0Hot leg temperature [K] 574.0 ± 1 573.0Steam generator pressure [MPa] 5.43 + 0.11 5.38

Steam generator mass flowrate [kg/s] 25.0 ± 0.4 25.0Pressurizer level [m] 1.164) + 0.01 1.16Core bypass fraction (LP to UP)1) [%] 3.6 3.45Core bypass fraction (DC to UP) 2) [%] 6.6 6.62Core bypass fraction (RABV)3 ) [%] 1.3 1.30

Core power [MW] 48.9 ± 1.0 48.91) Core bypass fraction from lower plenum to upper plenum.2) Core bypass fraction from downcomer to upper plenum.3) Core bypass fraction through the reflood assist bypass valve.4) Including the instrumentation elevation difference.

Table 4.4-2 Primary Test Chronology for LOFT/L3-1

Event Experiment 9 M-RELAP5

(sec) (sec)

Reactor scram -2.15 -2.15LOCA initiated 0.0 0.0Primary coolant pumps tripped 0.04 ± 0.01 0.04

Scaled HPIS initiated 4.6 ± 0.5 0.951)Pressurizer empty 17.0 ± 1 23Pump coastdown complete 19.0 ± 1 312)

Accumulator injection initiated 633.6 ± 0.5 655.85Accumulator liquid level below standpipe 1570.0 ± 1 1558Accumulator line empty of fluid 1741.0 ± 1 1690

SCS steam bleed initiated 3622.5 ± 1 -

LPIS injection initiated 4240.0 ± 1 -Experiment completed 4368.0 ± 1

1) Determined when the RCS pressure is less than 13.07 MPa.2) Determined when the RCP head is less than 0.0m.


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500 1000 1500 2000 2500

Time (s)

Figure 4.4-4 Break Mass Flowrate for LOFT/L3-1

3000

a_

Q)

U)P1

a_

0 500 1000 1500 2000 2500 3000Time (s)

Figure 4.4-5 Secondary System Pressure for LOFT/L3-1



M-RELAP5Measurementi

W

U)

10 ................

N•J

II I

0 500 1000 1500 2000 2500 3000

Time (s)

Figure 4.4-6 Primary System (Upper Plenum) Pressure for LOFT/L3-1

1.

-j

E

2'_J

0..

0.

-0..100

Time (s)

Figure 4.4-7 Pressurizer Liquid Level for LOFT/L3-1



0)

UW)

1• 0 ...... .. ............. ............................ .. .. .. .. ... .. .. .. .. .. .................... .... ...................... .... ....................... .

C.

-20i

0 500 1000 1500 2000 2500 3000

Time(s)

Figure 4.4-8 Differential Pressure in Intact Loop Crossover Leg for LOFTIL3-1(SG-Side)

20 1

U)()

ci)0.

cv

--

10

0

-10

_20C

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

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

.1 -. , JL

E M-RELAP5-MeasurementI

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

,-r--

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

L

0 500 1000 1500 2000 2500 3000Time(s)

Figure 4.4-9 Differential Pressure in Intact Loop Crossover Leg for LOFT/L3-1(RCP-Side)



Cu

0v

0)

0 500 1000 1500 2000 2500 3000

Time (s)

Figure 4.4-10 Accumulator Tank Pressure for LOFT/L3-1

20. I

1.5

M-RE me nMeasurementI

_j-J-o

.-J

1.0

0.05 .. . .. .... .............

0 500 1000 1500 2000 2500 3000

Time (s)

Figure 4.4-11 Accumulator Tank Water Level for LOFT/L3-1



2,

(V.

CL

EI-

1500Time (s)

3000

Figure 4.4-12 Fuel Cladding Temperature for LOFT/L3-1 (Z=62-in)



4.5 SEMISCALE S-LH-1 Test



The Semiscale Program was a part of the Water Reactor Research Test Program Divisionof EG&G Idaho, Inc., which conducted research of the thermal-hydraulic phenomenaassociated with simulated accident conditions in a PWR. The Semiscale Mod-2C systemas structured during the S-LH-1 and S-LH-2 experiments simulated centerline cold legsmall break loss-of-coolant accidents (5% SBLOCAs) 4-19.

Semiscale Mod-2C is a scaled model representation of a PWR plant, with a fluid volumeof about 1/1705 of a PWR (Figure 4.5-1). The modified-volume scaling philosophyfollowed in the design of the Mod-2C system preserves most of the first-order effectsthought important for SBLOCA transients. Most notably, the 1:1 elevation scaling of theSemiscale system is an important criterion for preserving the factors influencing signatureresponse to a SBLOCA. Details of the scaling principle and the scaling results aredescribed in Reference 4-20.

The Mod-2C system'-" consists of a pressure vessel with external downcomer andsimulated reactor internals: an "intact loop," with a shell and inverted U-tube active steamgenerator (SG), pressurizer, and pump; and a "broken loop," including an active pump,active SG, and associated piping to allow break simulations. The intact loop simulatesthree "unaffected loops" of a four-loop PWR, and the broken loop simulates an "affectedloop" in which the small break is assumed to occur. The break simulates a 5% cold-leg,centerline, communicative break in the loop piping between the pump and vessel. Theintact loop SG consists of six inverted U-tubes, and the broken loop SG consists of twoinverted U-tubes. Vessel internals include a simulated core, consisting of a 5 x 5 array ofinternally heated electric rods, of which 23 were powered as shown in Figure 4.5-2. Therods are geometrically similar to nuclear rods, with a heated length of 3.66 m (12 ft) andan outside diameter of 1.072 cm (0.42 in.).


The Semiscale S-LH-1 test exhibited phenomena not generally observed during previousSBLOCA transients. In particular, a severe core uncovery occurred prior to clearing of theloop seals at the suction of the reactor coolant pumps. Following the initialdepressurization, significant CCFL (counter-current flow limit) occurred in the hot legpiping and uphill-side of the SG U-tubes, particularly in the broken loop which consisted ofsmaller piping than that of the intact loop. During the loop seal phase, the core liquid levelwas depressed by the liquid holdup in the hot legs and uphill-side of the U-tubes. Theupper portion of the core uncovered and heater rod cladding started heating up. Theheat-up was terminated by an increase in the core liquid level following the loop sealclearance.

Since a large flow resistance occurred at the spray nozzle between the upper head anddowncomer, relatively little steam was vented from the core, resulting in the significantcore liquid level depression prior to loop seal clearance. Early seal clearance wasobserved in the intact loop because a larger amount of liquid flooded at the hot leg and



U-tubes in the broken loop compared to the intact loop.



The Semiscale Mod-2C system is numerically represented by the noding diagramillustrated in Figure 4.5-3. The primary feature is that the system is nodalized with thesame manner as for US-APWR SBLOCA calculations. The M-RELAP5 SemiscaleMod-2C model primarily .consists of the a) reactor vessel, b) downcomer pipe, c)pressurizer, d) steam generator, e) intact loop, f) broken loop, and g) ECCS.

] The CCFL inthe SG U-tubes is modeled by using the Wallis correlation4 -14 , wheref3=0.0, c=0.88, andm=1.0 are applied. This modeling is identical to that for the US-APWR plant calculation,because the geometric scaling of the SG U-tubes is almost identical between theSemiscale and US-APWR. [

The break flow history is imposed as a boundary condition which is specified by the inputdata based on the measurement so as to exclude uncertainties and distortions which arecaused by applying the conservative Moody critical flow model as was done previously forthe other lET calculations. The core power and secondary pressure are also given throughinput data tables for the present calculation.




The steady-state calculation was performed by M-RELAP5. The converged plantparameters are listed in Table 4.5-1 in which the calculation results are compared with themeasurements. The table shows that M-RELAP5 accurately reproduces the steady-statecondition prior to the transient test for the Semiscale/S-LH-1.

The chronology during the Semiscale/S-LH-1 test is listed in Table 4.5-2, where theexperimental and calculated results are compared. The transient calculation was initiatedby the simulated break flow data shown in Figure 4.5-4. The measured secondary systempressure was also given by a boundary condition as shown in Figure 4.5-5. M-RELAP5well simulates the primary system pressure response as shown in Figure 4.5-6, indicatingthat the system mass and energy balances during the test are well reproduced by theM-RELAP5 calculation.

A complicated loop seal behavior was observed in the Semiscale/S-LH-1 test, where thecoolant seal in the intact loop cleared first and the broken loop seal cleared about 90 slater. This loop seal behavior can be simulated by M-RELAP5 as shown in Figure 4.5-7and Figure 4.5-8, which demonstrate the code's ability to predict the loop seal behaviorduring SBLOCAs. It is noted that M-RELAP5 predicts transient decrease in the collapsedliquid level for the broken loop crossover leg as core liquid level depression during theloop seal period, while not observed in the measurement. However, the resultant coreliquid level depression predicted by M-RELAP5 is deeper than the measurement,indicating that conservative prediction with respect to the loop seal PCT.

In addition, a severe reflux flooding occurred in the hot leg piping and SG-U-tubes in theS-LH-1 test and the core liquid level was significantly depressed during the loop sealphase. This was primary caused by the small core bypass flow fraction between the upperhead and downcomer, which prevented the steam from venting from the core. This causewas experimentally validated by comparing the two tests, S-LH-1 (0.9% bypass) andS-LH-2 (3.0% bypass) of the Semiscale Program4 ' 9 . M-RELAP5 results are comparedwith the measurements from Figure 4.5-9 to Figure 4.5-11 in terms of the hot leg for theintact and broken loops, and the core liquid level, respectively. The severe flooding andcore liquid depression can be well simulated by M-RELAP5.

As a result of the core liquid depression, the heater rod experienced the dryout andheat-up during the loop seal phase. This temperature excursion was terminated byincrease in the core liquid level after the loop seal cleared as shown in Figure 4.5-11.Histories of the measured and calculated heater rod surface temperature are compared inFigure 4.5-12 and the peak values are listed in Table 4.5-3. M-RELAP5 is capable ofpredicting the heater rod temperature behavior accurately. [

I

Figure 4.5-7 through Figure 4.5-11 show collapsed liquid levels. The approach used hereis consistent with that used in Reference 4-19, where the calculated values were obtainedby integrating liquid volume fraction distributions and the measured values were obtainedfrom differential pressure measurements. Differences between the two methods affectthe level comparisons before 90 s, when the pumps finish coasting down, because the



measured differential pressures are affected by flow. Also, the difference between thecalculated and measured cladding temperatures shown in Figure 4.5-12 prior to scram iscaused by the comparison of a calculated surface temperature with a measuredtemperature inside the heater rod.

4.5.3 Summary

The simulation of Semiscale S-LH-1 was performed by M-RELAP5. The resultsdemonstrate that M-RELAP5 well predicted the complicated plant responses, includingthe loop seal behavior. In particular, the severe core depression and heater rodtemperature excursion during the loop seal phase were well reproduced by M-RELAP5,showing its high applicability to the PWR SBLOCA safety analysis.



Table 4.5-1 Steady-State Parameters for Semiscale/S-LH-1

Parameter Experiment19 M-RELAP5

Pressurizer pressure [MPa] 15.47 ± 0.14 15.47Core AT 37.65 +1.5/-0.6 37.52Intact loop flowrate 7/13 7.11Broken loop flowrate 2.35 2.34Intact loop cold leg temperature [K] 562.12 ± 2 562.06Broken loop cold temperature [K] 564.05 ± 2 564.07Intact loop Steam generator pressure [MPa] 5.72 ± 0.07 5.72Broken loop Steam generator pressure [MPa] 6.08 ± 0.07 6.08Pressurizer level [cm] 395 ± 14 394.95Core bypass fraction [%] 0.9 0.9Core power [MW] 2014.75 ± 0.15 2014.75

Table 4.5-2 Primary Test Chronology for Semiscale/S-LH-1

Experiment19 M-RELAP5Event (sec) (sec)

Pressurizer at 12.6MPa (trip level) 14.67 22.10Core scram 19.57 26.70Pump coastdown initiated

Intact loop 21.35 26.80Broken loop 20.76 26.45

HPIS initiatedIntact loop 41.60 48.40Broken loop 40.98 48.40

Minimum core liquid level reached 172.6 179Intact loop pump suction cleared 171.4 183Broken loop pump suction cleared 262.3 262

Table 4.5-3 Summary of PCTs during Loop Seal for Semiscale/S-LH-1

Time (s) PCT (K)Measured PCT 182.4 624.4M-RELAP5 191.5 634.1



Broken loopatmosphere dumpvalve (ADV) 6

A Intact loop maljsteam isolationvalve (MSIV)

Intact loopType II steamgenerator

I r

measuring tank 4 o4o2

Figure 4.5-1 Semiscale Mod-2C System Configuration 4 21



A B

Kt143

-cmL

T f(._ •• /, \.2533,13 182 16• 53

01.072m 160oO D4 19

C

1180"136

29

0 5918231-354196

0

208

226157

289 137

E

10

320K 134

2 53r- 16 6

2

Density beam throughthis channel at 12. 113.183. 253. and 342 cmabove bottom of core

P, 3

319 11

L 135q4~~l165 181 40192O 2

319 135 6 124

166 251Q 181 13?

209 158

230(~7~)138

290

137

C167

Cold leg centerline

207156

7437

208

229 5

7 5229 136

22:*25

® Unpowered rod

Thermocouple azimuthal location

\.J ;)s 2 - Elevation above core bottom

40e40

Figure 4.5-2 Semiscale Mod-2C Core Heater Rod Configuration4 21


(D

m

c(D

O-o

c'n

0)-o

Z3

o0

CD?

CmCo',

Figure 4.5-3 M-RELAP5 Nodalization Diagram for Semiscale/S-LH-1


1.

0.

0.to

S0.

0.

0.

Time (sec)

Figure 4.5-4 Break Mass Flowrate for Semiscale/S-LH-1

8.01

7.5--

7.0IC.

0)

CoCo0)

........----6.5F-

M-RELAP5- Measurement

/Intact Loop

............ B R O K EN... ......Loop ---.- .. ..........................

-- - - - - - - -- - - - - - - ------.. . . . . . . . . . . • . . . . . . . . . . . .... .. . .--- -- --- --- -- --- --- --- -- - -- --- -- --- --36.0

5.51-

0'-100 200 300 400

Time (sec)

Figure 4.5-5 Secondary System Pressure for Semiscale/S-LH-1



1

1

CD

a-

0_

Time (sec)

Figure 4.5-6 Primary System Pressure for Semiscale/S-LH-1

-100

-.5

400

Figure 4.5-7 Collapsed Level in Uphill-Side of Intact Loop Crossover Leg forSemiscale/S-LH-1



Eý

C.-

-J

_j

Time (sec)

Figure 4.5-8 Collapsed Level in Uphill-Side of Broken Loop Crossover Leg forSemiscale/S-LH-1

EC-,

a)

a)-J

0*

-J

400

Time (sec)

Figure 4.5-9 Collapsed Level in Intact Loop Hot Leg for Semiscale/S-LH-1



C.

WJ

Time (sec)

Figure 4.5-10 Collapsed Level in Broken Loop Hot Leg for Semiscale/S-LH-1

-- 200

* -300or

-500' 0 2__ 0 3000 100 200 300 400

Time (sec)

Figure 4.5-11 Core Collapsed Level for Semiscale/S-LH-1



4-

0.

E4-,

.-,

CD_

CD

Time (sec)

Figure 4.5-12 Core Cladding Temperature at 8.3-ft (253cm) Elevation



4.6 References

4-1 The ROSA-IV Group, 'ROSA-IV Large Scale Test Facility (LSTF) SystemDescription,' JAERI-M 84-237, January 1985.

4-2 The ROSA-IV Group, 'Supplemental Description of ROSA-IV/LSTF with No.1Simulated Fuel-Rod Assembly,' JAERI-M 89-113, September 1989.

4-3 H. Kumamaru et al., 'ROSA-IV/LSTF 5% Cold Leg Break LOCA Experiment RunSB-CL-18 Data Report,' JAERI-M 89-027, 1989.

4-4 Mitsubishi Heavy Industries, Ltd., 'Scaling Analysis for US-APWR Small BreakLOCAs,' UAP-HF-09568, December 2009.


4-6 Mitsubishi Heavy Industries, Ltd., 'M-RELAP5 Code -Supplementary ManualVolume I: Code Structure, System Models and Solution Methods,'6AS-1 E-UAP-090047 (RO), August 2009 (Proprietary).

4-7 Mitsubishi Heavy Industries, Ltd., 'M-RELAP5 Code Supplementary ManualVolume IV: Models and Correlations,' 6AS-1E-UAP-090071 (RO), December 2009(Proprietary).

4-8 M. Suzuki and H. Nakamura, 'A study on ROSA/LSTF SB-CL-09 Test SimulatingPWR 10% Cold Leg Break LOCA: Loop-Seal Clearing and 3D core Heat-UpPhenomena,' JAEA-Research, October 2008.

4-9 P. D. Bayless et al., 'Experimental Data Report for LOFT Nuclear Small BreakExperiment L3-1,' NUREG/CR-1145, EGG-2007, January 1980.

4-10 L. J. Ybarrando et al., 'Examination of LOFT Scaling,' ASME Annual Meeting,74-WA/HT-53, New York, NY, November 17-22, 1974.

4-11 M. L. Russell, 'LOFT Fuel Modules Design, Characterization, and FabricationProgram,' TREE-NUREG-1131, June 1977.

4-12 D. L. Reeder, 'LOFT System and Test Description (5.5-ft Nuclear Core 1 LOCEs),'NUREG/CR-0247, TREE-1208, July 1978.

4-13 K. G. Condie et al., 'Four-Inch Break Loss-of-Coolant Experiments: PosttestAnalysis of LOFT Experiment L3-1, L3-5 (Pumps Off), and L3-6 (Pumps On),'EGG-LOFT-5480, October 1981.

4-14 G. B. Wallis, 'One Dimensional Two Phase Flow,' McGraw-Hill, 1969.4-15 C. L. Tien, et al., 'Flooding in Two-Phase Countercurrent Flows,' EPRI NP-1283,

December 1979.4-16 Mitsubishi Heavy Industries, Ltd., 'Small Break LOCA Methodology for US-APWR,'

MUAP-07013-P (RO), July 2007.4-17 P. S. Damerell et al., 'Use of Full-Scale UPTF Data to Evaluate Scaling of

Downcomer (ECC Bypass) and Hot Leg Two-Phase Flow Phenomena,'NUREG-CP-0091 vol. 4.

4-18 F. J. Moody, 'Maximum Flow Rate of a Single Component, Two-Phase Mixture,' J.of Heat Transfer, Trans. ASME, Series C, Vol. 87, No. 1, pp134-142, February1965.

4-19 G, G. Loomis and J. E. Streit, 'Results of Semiscale Mod-2C Small-Break (5%)Loss-of-Coolant Accident Experiments S-LH-1 and S-LH-2,' NUREG/CR-4438,EGG-2424, R2, November 1985.

4-20 R. A. Shaw et al., 'A Description of the Semiscale Mod-2C Facility, IncludingScaling Principle and Current Measurement Capabilities,' EGG-M-11485, January1985.

4-21 G. G. Loomis, 'Experimental Operating Specifications for Semiscale Mod-2C 5%


6AS-1 E-UAP-1 00001 (RO). Non-Proprietary VersionM-RELAP5 Supplementary Manual Vol. III

Small Break Loss-of-Coolant Experiment S-LH-1,' EGG-SEMI-6813, February1985.

4-22 H. T. Kim and J. C. No, 'Assessment of RELAP5/MOD3.2.2y against floodingdatabase in horizontal-to-inclined pipes,' Annals of Nuclear Energy, Vol. 29, Issue 7,May 2002.



5. REVIEW FOR CODE ASSESSMENT RESULTS

M-RELAP5 has been validated using a wide set of experimental data obtained fromvarious SET and lET facilities as listed in Table 2.2-1. The results of the code modelsassessment and validation demonstrates that the code is capable to predict all theimportant phenomena and processes expected during the postulated SBLOCA transientsof the US-APWR, as identified in the PIRT (Table 2.1-1).

The SET facilities used for the M-RELAP5 assessment provides well scaled test data notonly from the viewpoint of the geometrical condition but also the experimental conditionrepresenting US-APWR SBLOCAs. Through the various lET analyses, M-RELAP5 hasbeen assessed using experimental test data obtained from test facilities with differentscales and break sizes, demonstrating that the M-RELAP5 is applicable to predictUS-APWR plant response during SBLOCA transients. In particular, the scalability of theROSA-IV/LSTF SB-CL-18 test, which is the most important aspect in the M-RELAP5assessment, is quantitatively investigated and ensured by the hierarchical two-tieredscaling methodology. The scaling analysis results are given in the separate reportdeveloped by MHI.

The followings summarize the primary results obtained from the M-RELAP5 codeassessment and validation.

(1) CHF/DryoutCHF/dryout model was assessed using the ORNL/THTF uncovered-bundle heattransfer test from SETs, and the ROSA-IV/LSTF and Semiscale SBLOCA tests fromlETs. M-RELAP5 accurately or conservatively predicts the onset of dryout

(2) Uncovered Core Heat TransferUncovered core heat transfer model was assessed using the same experimental testas was done for the CHF/dryout model. The code validation indicates that theM-RELAP5 calculates conservative heat transfer coefficient resulting in a higher PCTcompared with the measurement.

(3) RewetThe ORNLITHTF high-pressure reflood test and the FLECHT-SEASET forced refloodtest provided the code assessment data for the rewet model. In addition, the modelwas validated by applying the SBLOCA lET analyses obtained in the ROSA-IV/LSTFand Semiscale facilities. The code validation demonstrates that M-RELAP5 predicts alonger period of dryout duration in comparison with the measurement.

(4) Core Mixture LevelPrediction for the core mixture level is related to the interfacial drag model whichaffects the void distribution. The model was assessed using the ROSA-IV/LSTF voidprofile test and the ORNL/THTF mixture level swell test from SETs. In addition, theROSA-IV/LSTF and Semiscale SBLOCA lETs provided the well instrumented mixturelevel data in addition to the collapsed level data. M-RELAP5 is capable to accuratelyand conservatively predict the mixture level and dryout region, compared with themeasurements.

(5) Water Hold-Up in SG Inlet Plenum



The CCFL model for the reflux flooding at the SG inlet plenum was developed andassessed using the experimental data obtained in the full-scale test facility, UPTF.The code calculation accurately reproduces the CCFL behavior.

(6) Water Hold-Up in SG U-Tube Uphill SideThe CCFL in the SG U-tubes are simulated by the Wallis model for the US-APWRSBLOCA analysis by M-RELAP5. The model was validated by using the DuklerAir-Water experimental data, and M-RELAP5 provides a good agreement with themeasurements in terms of the flooding in the narrow piping.

(7) SG Primary and Secondary Heat TransferIn the M-RELAP5 code assessment, the SG heat transfer model is validated usinglET test data obtained in the ROSA-IV/LSTF SB-CL-18 test. In the code calculation,mechanical motions of the main steam isolation and relief valves and the feedwaterbehavior are explicitly modeled to simulate the SG secondary system pressure aswell as are done for actual plant calculations. M-RELAP5 well predicts the plantresponses both for the primary and secondary systems, indicating that the SG heattransfer model in M-RELAP5 is valid.

(8) Water Level in SG Outlet PipingWater level in SG outlet piping is important in predicting the loop seal behavior.M-RELAP5 was assessed using experimental data obtained in various lET facilities,ROSA-IV/LSTF, LOFT, and Semiscale. In addition, water retention after the loop sealcleared was validated using SET data from the full-scale UPTF test facility. Thesecode validations demonstrated that M-RELAP5 is capable of predicting the waterlevel behavior during SBLOCAs accurately.

(9) Loop Seal Formation and ClearanceAs described in the preceding item, M-RELAP5 ability to predict the loop sealbehavior was assessed using various SET and lET experimental data. Particularly,M-RELAP5 is capable to well reproduce various modes in terms of the clearingbehavior observed in the various lET facilities with different scales: a) all loops are notcleared, b) all loops are cleared, and c) loop seal clearance occurs partially orheterogeneously. These code validation results shows that M-RELAP5 is reliablyapplicable to predict the loop seal behaviors.

(10) Downcomer Mixture LevelNoid DistributionSimilar to the loop seal formation and clearance, M-RELAP5 was assessed for thedowncomer behaviors using various lET data. The ROSA-IV/LSTF test, particularly,provided well instrumented experimental data for the code assessment, andM-RELAP5 well reproduced the downcomer behavior observed in the lET.



6. CONCLUSIONS

The M-RELAP5 code has been developed by the Mitsubishi Heavy Industries, Ltd. (MHI)and currently used for the safety analyses of US-APWR plant in responding to a set ofpostulated small break LOCA transients. The present supplementary manual describesthe strategy or approach adopted by MHI for M-RELAP5 code assessment and presentsthe assessment results. In the code assessment, the important phenomena andprocesses involved in US-APWR during SBLOCA are identified and formulated in theUS-APWR SBLOCA's PIRT, which is the basis of the code assessment matrix forM-RELAP5. The code assessment matrix consists of various SETs and lETs.

The SET assessment consists of (1) ROSA-IV/LSTF void profile test, (2) ORNL/THTF voidprofile and uncovered-bundle heat transfer tests, (3) ORNLITHTF high-pressure refloodtest, (4) FLECHT-SEASET forced reflood test, (5) UPTF CCFL test, (6) Dukler air-waterflooding test, (7) UPTF loop seal test, and (8) MHI's scaled advanced accumulator test.

The lET assessment comprises (1) ROSA-IV/LSTF SB-CL-18 test (5% break), (2)SB-CL-09 test (10% break), (3) SB-CL-12 (0.5% break), (4) LOFT L3-1 test (2.5% break),and (5) Semiscale S-LH-1 test (5% break).

The code assessment results show that the M-RELAP5 is capable to well predicts all theimportant phenomena and processes identified for the US-APWR SBLOCA. Particularly,the code is capable to reproduce the primary transient behaviors observed in various lETfacilities with different scales, and to predict the fuel cladding temperature excursionsaccurately or conservatively. Therefore, it can be concluded that M-RELAP5 is sufficientlyapplicable to the safety analyses for the US-APWR SBLOCA transients.


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