ML18120A150

Enclosure 4 - WCAP-17938-NP, Revision 3, AP1000 In-Containment Cables and Non-Metallic Insulation Debris Integrated Assessment (Non-Proprietary)

ML18120A150
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Site:	05200006
Issue date:	04/30/2018
From:	Westinghouse
To:	Office of New Reactors
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ML18121A237	List: ML18120A149 ML18120A150
References
APP-GW-GSR-013, AW-18-4734, DCP _NRC_003332 WCAP-17938-NP, Rev 3
Download: ML18120A150 (164)

Text

ENCLOSURE 4 to AW-18-4734 WCAP-17938-NP, Revision 3 AP1000 In-Containment Cables and Non-Metallic Insulation Debris Integrated Assessment (Non-Proprietary)

Westinghouse Non-Proprietary Class 3 WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3 AP1000 In-Containment Cables and Non-Metallic Insulation Debris Integrated Assessment

Westinghouse Non-Proprietary Class 3 WCAP-17938-NP Revision 3 APP-GW-GSR-013 Shayantan Sinha U.S. Regulatory Support Kevin McNamee Systems & Equipment Engineering I April 2018 Reviewer: Robert Pinkston AP1000 Plant Technical Services and Integration Approved: Jason Eisenhauer Manager, AP1000 Plant Technical Services and Integration

• Electronically approved records are authenticated in the electronic document management system.

Westinghouse Electric Company LLC 1000 Westinghouse Drive Cranberry Township, PA 16066, USA

© 2018 Westinghouse Electric Company LLC All Rights Reserved WCAP-17938-NP

Westinghouse Non-Proprietary Class 3 ii RECORD OF REVISIONS Revision Description 0 Initial issue.

1 All changes in the document are denoted by a left column change bar, and are summarized below:

Changes in the Executive Summary, Section 1, Section 3, and Section 7 have been made to reflect the additional neutron shielding block (containing non-metallic insulation NMI) jet impingement testing performed at [*****]a,b,c.

The locations to which the suitable equivalency applies have been clarified throughout the document.

The request for approval of the onset of incipient damage for cables at

[*******]a,b,c was removed.

Section 3.5.4 replaced Section 3.4.4.4 and calculates the predicted impingement pressure on the NMI blocks.

Full NMI submergence test results were included in Subsection 3.6.

A detailed discussion of break scenario selection is included in Subsection 4.4.1.

Additional information on the AP1000 plant chemical effects model is included in Subsection 5.1.3.

Additional pictures are included throughout Section 5 to clarify orientation.

Editorial and formatting changes were made throughout the document and are not marked with change bars.

2 This revision incorporates E&DCR: APP-CA31-GEF-024, Modifications to CA31 Shield Blocks.

This revision also incorporates responses and clarifications related to Nuclear Regulatory Commission (NRC) Requests for Additional Information (RAIs) issued on Revision 1.

All changes in this document, with the exception of editorial, grammatical, and format changes, are denoted by a left column change bar and are summarized below:

Changes in the List of Acronyms, Executive Summary, Section 1, Section 2, Section 3, Section 5, and Section 7 have been made to reflect a change in the components containing non-metallic insulation (NMI) and a change in the neutron absorbing material contained in the CA31 module.

Section 5.1.3 was added to assess the thermal expansion of [ ]a,c The locations to which the requirement for suitable equivalency applies have been clarified throughout the document.

The upper neutron shield (UNS) has been redesigned and no longer contains NMI; therefore this document will no longer seek NRC approval for the UNS NMI as a suitable equivalent to metal reflective insulation (MRI) as required by Design Control Document (DCD) Section 6.3.2.2.7.1 Item 3.

Corrected error in plant pressure distribution integration (now in sub section 3.3.5.5) which was identified in CAPAL 100473253.

WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 iii Revision Description 3 This revision also incorporates responses and clarifications related to Supplemental Revisions issued after Revision 2.

All changes in this document, with the exception of editorial, grammatical, and format changes, are denoted by a left column change bar and are summarized below:

Removed Proprietary markings from 4D ZOI when discussing the applicable ZOI for AP1000 In-Containment cables.

Incorporated changes presented in APP-GW-GLY-131/-132 to Sections 3.5, 3.6, 5.1.3, and 5.2. Added new Section 5.1.2.1 to Section 5.1.2.

Incorporated changes presented in APP-GW-GLY-152 to Appendix A.

Corrected copy errors in Sections 4.4.3.1.1, 4.4.3.2.2, 4.4.3.2.3, and mislabeled text in Section 5.1.1.1.5.2 Corrected location error for red dot marking AP1000 R/D in Figure 3-79.

Additional changes to Appendix A for consistency and clarification.

WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 iv TABLE OF CONTENTS RECORD OF REVISIONS .......................................................................................................................... ii LIST OF TABLES ....................................................................................................................................... vi LIST OF FIGURES .................................................................................................................................... vii LIST OF ACRONYMS, ABBREVIATIONS, AND TRADEMARKS ....................................................... xi EXECUTIVE

SUMMARY

........................................................................................................................ xiii 1 INTRODUCTION ........................................................................................................................ 1-1 1.1 PURPOSE ........................................................................................................................ 1-1 1.2 LIMITS OF APPLICABILITY........................................................................................ 1-2

1.3 REFERENCES

................................................................................................................ 1-2 2 POTENTIAL SOURCES OF ADDITIONAL DEBRIS ............................................................... 2-1 2.1 CABLES .......................................................................................................................... 2-1 2.1.1 Description of the Cables ................................................................................ 2-1 2.2 REACTOR VESSEL CAVITY NON-METALLIC INSULATION................................. 2-2 2.2.1 Description of the RVIS Water Inlet Doors ..................................................... 2-3 2.2.2 Description of the RVIS Lower Neutron Shielding ......................................... 2-3 2.2.3 Description of the CA31 Neutron Shielding ................................................... 2-4

2.3 REFERENCES

................................................................................................................ 2-6 3 GSI-191 TEST PROGRAM

SUMMARY

.................................................................................... 3-1 3.1 JET IMPINGEMENT TEST BACKGROUND ............................................................... 3-1 3.2 JET IMPINGEMENT TEST FACILITY ......................................................................... 3-3 3.2.1 Comparison of PWROG Facility with AP1000 Plant Facility ........................ 3-4 3.2.2 Comparison of PWROG Facility Data with AP1000 Facility Data............... 3-10 3.2.3 Comparison of AP1000 Plant licensing basis with AP1000 Plant facility data . 3-18 3.3 JET IMPINGEMENT TESTS........................................................................................ 3-21 3.3.1 Jet Impingement Test Objectives ................................................................... 3-21 3.3.2 Jet Impingement Test Specimens................................................................... 3-21 3.3.3 Cable Jet Impingement Test Summary .......................................................... 3-26 3.3.4 Neutron Shield Block Jet Impingement Test Summary................................. 3-39 3.3.5 Considerations Resulting from Confined Jet Behavior ................................. 3-55 3.4 NEUTRON SHIELD BLOCK SUBMERGENCE TEST

SUMMARY

AND OBJECTIVES ................................................................................................................ 3-77 3.4.1 Neutron Shield Blocks Submergence Test Summary .................................... 3-77 3.4.2 Submergence Test Specimens........................................................................ 3-77 3.5 CHARACTERIZATION OF CA31 NEUTRON SHIELDING MATERIAL ............... 3-81 3.5.1 Impact of CA31 Neutron Shielding Material on Chemical Effects ............... 3-81 3.5.2 Impact on Containment Recirculation Screens.............................................. 3-82

3.6 REFERENCES

.............................................................................................................. 3-82 4 DEBRIS GENERATION BREAK SIZE DETERMINATION .................................................... 4-1 4.1 DEBRIS GENERATION BREAK SIZE ......................................................................... 4-1 WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 v 4.2 REGION I ANALYSIS .................................................................................................... 4-2 4.3 REGION II ANALYSIS................................................................................................... 4-5 4.3.1 Full Separation Break ...................................................................................... 4-6 4.3.2 Limited Separation Breaks .............................................................................. 4-6 4.4 RCS MAIN LOOP PIPING DISPLACEMENT ANALYSIS .......................................... 4-9 4.4.1 Analysis Overview .......................................................................................... 4-9 4.4.2 Pipe Displacement Results ............................................................................ 4-14 4.4.3 Application of Results to Region II Analyses................................................ 4-15

4.5 REFERENCES

.............................................................................................................. 4-21 5 NON-METALLIC INSULATION SUITABLE EQUIVALENCY ............................................... 5-1 5.1.1 Debris Generation ............................................................................................ 5-1 5.1.2 Aging Effects ................................................................................................. 5-14 5.1.3 Thermal Expansion of LNS Neutron Shielding Material .............................. 5-15 5.1.4 Additional Conservatisms.............................................................................. 5-16

5.2 REFERENCES

.............................................................................................................. 5-17 6 REGULATORY IMPACTS .......................................................................................................... 6-1 6.1 LICENSING BASIS CHANGES .................................................................................... 6-1 6.2 REFERENCE................................................................................................................... 6-1 7 CONCLUSIONS .......................................................................................................................... 7-1

7.1 REFERENCES

................................................................................................................ 7-2 APPENDIX A AP1000 DCD REVISION 19 MARKUPS ..................................................................... A-1 WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 vi LIST OF TABLES Table 2-1. Cable Jet Impingement and Characterization Test Program Samples (Figure 2-1) ................. 2-2 Table 3-1. Blowdown Facility Components.............................................................................................. 3-5 Table 3-2. AP1000 Plant Conditions vs. Test Conditions ....................................................................... 3-18 Table 3-3. Cable Jet Impingement Test Matrix ....................................................................................... 3-26 Table 3-4. Cable Arrangement in Test Fixture ........................................................................................ 3-30 Table 3-5. Cable Jet Impingement Test Results ...................................................................................... 3-31 Table 3-6. Large Cable Damage Summary (percent loss of material) .................................................... 3-33 Table 3-7. Small Cable Damage Summary (percent loss of material) .................................................... 3-34 Table 3-8. Neutron Shield Block Jet Impingement Test Results Summary ............................................ 3-40 Table 3-9. Key NMI Dimensions for Assessing Confined Jet Behavior ................................................. 3-59 Table 3-10. SI units of Cold Leg Temperature, Pressure, and Subcooling for use in developing impingement pressures from Figure 3-76 (Reference 3-6) ............................................ 3-74 Table 3-11. NTS Stagnation Pressures on Neutron Shield Blocks.......................................................... 3-75 Table 3-12. Integrated Elemental Release............................................................................................... 3-80 Table 4-1. Comparison of Proposed ZOI to Volume-Equivalent ZOI ...................................................... 4-4 Table 4-2. Comparison of Jet Model Benchmark to Table I-3 of Reference 4-2 ...................................... 4-5 Table 4-3. Results of Pipe Displacement Analysis - APP-PL01-PL0C-003 (Reference 4-7) ................ 4-15 Table 5-1. NEI 04-07 Comparison of Computed Spherical ZOI Radii from Independent Evaluations of the ANSI Jet Model.......................................................................................................... 5-2 Table 5-2. Chemical Debris Generation for AP1000 Plant Licensing Basis Case .................................. 5-16 WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 vii LIST OF FIGURES Figure 2-1. Cables used in the Cable Jet Impingement Test Program ...................................................... 2-2 Figure 2-2. CA31 Neutron Shield Block (typical) .................................................................................... 2-5 Figure 2-3. CA31 Supplemental Neutron Shield Block (typical) ............................................................. 2-5 Figure 2-4. CA31 Shield Block System (typical) ..................................................................................... 2-6 Figure 3-1. Schematic of NTS High Flow Blowdown Facility ................................................................ 3-6 Figure 3-2. As-Built AP1000 Plant High Flow Test Facility Configuration ............................................. 3-7 Figure 3-3. PWROG As-Built Facility...................................................................................................... 3-8 Figure 3-4. AP1000 Plant High Flow Test Facility Configuration for Jet Impingement Testing.............. 3-9 Figure 3-5. PWROG Jet Impingement Test Program A3 Tank Pressure ................................................ 3-10 Figure 3-6. PWROG Jet Impingement Test Program Reducer Pressure................................................. 3-11 Figure 3-7. PWROG Jet Impingement Test Program Mass Flow Rate................................................... 3-11 Figure 3-8. PWROG Jet Impingement Test Program Exit Nozzle Pressure ........................................... 3-12 Figure 3-9. PWROG Jet Impingement Test Program Reducer Temperature .......................................... 3-12 Figure 3-10. PWROG/AP1000 Plant Jet Impingement Test Program A3 Tank Pressure ....................... 3-13 Figure 3-11. PWROG/AP1000 Plant Jet Impingement Test Program Reducer Pressure ....................... 3-14 Figure 3-12. PWROG/AP1000 Plant Jet Impingement Test Program Mass Flow Rate ......................... 3-14 Figure 3-13. PWROG/AP1000 Plant Jet Impingement Test Program Exit Nozzle Pressure.................. 3-15 Figure 3-14. PWROG/AP1000 Plant Jet Impingement Test Program Reducer Temperature ................. 3-15 Figure 3-15. Stagnation Pressure at [ ]a,b,c .................................................................... 3-16 Figure 3-16. Stagnation Pressure at [ ]a,b,c .................................................................... 3-17 Figure 3-17. Stagnation Pressure at [ ]a,b,c .................................................................... 3-17 Figure 3-18. Center Line Stagnation Pressure at the Target.................................................................... 3-18 Figure 3-19. Comparison of AP1000 Plant Licensing Basis and the AP1000 Plant Jet Impingement Test Program Stagnation Pressure .............................................................................. 3-19 Figure 3-20. Comparison of AP1000 Plant Licensing Basis and the AP1000 Plant Jet Impingement Test Program Mass Flux ............................................................................................. 3-20 Figure 3-21. Comparison of AP1000 Plant Licensing Basis and the AP1000 Plant Jet Impingement Test Program Reservoir Pressure ................................................................................ 3-20 Figure 3-22. Small Cables....................................................................................................................... 3-22 Figure 3-23. Large Cables....................................................................................................................... 3-23 Figure 3-24. Typical Neutron Shield Block Construction....................................................................... 3-24 Figure 3-25. Neutron Shield Block (Type I) ........................................................................................... 3-24 Figure 3-26. [************]a,c Neutron Shield Block (Type II) ........................................................... 3-25 Figure 3-27. [****************]a,c Neutron Shield Block (Type III) .................................................. 3-25 Figure 3-28. Original Cable Test Fixture Design - Cables Behind Fixture.......................................... 3-28 WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 viii Figure 3-29. Original Cable Test Fixture Design (Cable Clamp Placed Behind the Holding Plate) .... 3-28 Figure 3-30. Cable Damage from Cable/Fixture Interaction (Equivalent Damage at Both Top and Bottom of Window) .................................................................................................................... 3-29 Figure 3-31. Revised Cable Test Fixture that Eliminates Cable/Fixture Interaction by Placing Cable Clamp in Front of the Holding Plate ........................................................................... 3-29 Figure 3-32. Large and Small Cable Arrangements in the Test Fixture .................................................. 3-30 Figure 3-33. Large Cable [************************]a,b,c ................................................................ 3-35 Figure 3-34. Small Cable [************************]a,b,c ................................................................ 3-35 Figure 3-35. Large Cable [****************]a,b,c ................................................................................ 3-36 Figure 3-36. Small Cable [****************]a,b,c ................................................................................ 3-36 Figure 3-37. Large Cables at [*****************]a,b,c ........................................................................ 3-37 Figure 3-38. Small Cables at [*****************]a,b,c ........................................................................ 3-37 Figure 3-39. [************************]a,b,c (Cable Test CT9FT12) ............................................... 3-38 Figure 3-40. [************************]a,b,c (Cable CT10FT13) ..................................................... 3-38 Figure 3-41. [************]a,b,c Type I Neutron Shield Block at [******]a,b,c ..................................... 3-41 Figure 3-42. [************]a,b,c Type II Neutron Shield Block at [*****]a,b,c ...................................... 3-41 Figure 3-43. [ *]a,b,c Type III Neutron Shield Block at [******]a,b,c............................ 3-42 Figure 3-44. [* *** *********]a,b,c Type III Neutron Shield Block with [

• • • • • • • • • • • • • • • • • ]a,b,c ............................................................................................... 3-42 Figure 3-45. [****************]a,b,c Type III Neutron Shield Block at [*****]a,b,c............................. 3-43 Figure 3-46. Neutron Shielding Block Test 6 at [********************************]a,b,c ............. 3-44 Figure 3-47. Post-shot Neutron Shield Block Test 6 - [ *******************

• • • • ]a,b,c ......................................................................................................................... 3-45 Figure 3-48. Neutron Shield Block Test 6 [********************************************

• • • • • • ]a,b,c Neutron Shielding Blocks ........................................................................... 3-46 Figure 3-49. Neutron Shield Block Test 6 [*******************************************

• • • • • • • • • • • • • • • • • • • • • • • • • • • ]a,b,c ........................................................................... 3-46 Figure 3-50. Neutron Shield Block Test 6 [********* ********************************

• • • • • • • • • • • • • • • • • • • • • • • • ]a,b,c ................................................................................. 3-47 Figure 3-51. Neutron Shield Block Test 6 [************************ **

• • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • ]a,b,c.............................................. 3-47 Figure 3-52. Neutron Shielding Block Test at [**********************]a,b,c .................................... 3-49 Figure 3-53. Neutron Shield Block Test 7 - [ *******************************************

• • • • • • • • • • • • • • • • • • • ]a,b,c ........................................................................................... 3-49 Figure 3-54. Neutron Shield Block Test 7 - [ *********

• • • • • • • • • • • • • • • • • • • ]a,b,c ........................................................................................... 3-50 Figure 3-55. A and B Neutron Shield Block Test 7 - [***************]a,b,c ....................................... 3-50 WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 ix Figure 3-56. Neutron Shield Block Test 7 [ ************************************************ x

• • ]a,b,c ............................................................................................................ 3-51 Figure 3-57. Neutron Shield Block Test 7 [**********************************************

• • • • • • • • • • • • • • • • • • • • • • • • • *]a,b,c ........................................................................... 3-52 Figure 3-58. Neutron Shield Block Test 7 [*********************************************

• • • • • • • • • • • *****]a,b,c ........................................................... 3-52 Figure 3-59. Neutron Shield Block Test 7 [888888888888888888888888888888888888888888888888 -

8888888888888888888888888*8 ]a,b,c ........................................................... 3-53 Figure 3-60. Neutron Shield Block Test 7 [************************************************

55555555555555555555555555%%%55555]a,b,c.......................................................... 3-53 Figure 3-61. Neutron Shield Block Test 7 [5555555555555555555555555555555555555555555555 5555555555555555555555555]a,b,c ............................................................................... 3-54 Figure 3-62. Neutron Shield Block Test 7 [555555555555555555555555555555555555555555555555 -

55555555555555555555555555555555555]a,b,c ........................................................... 3-54 Figure 3-63. Neutron Shield Block Test 7 [55555555555555555555555555555555555555555555555 555555555555555555555555555 5555]a,b,c....................................................... 3-55 Figure 3-64. Depiction of Impinging Jet Regions on a Flat Plate........................................................... 3-56 Figure 3-65. Pressure distribution in as a function of L/D (x/B0) and R/D (y/2b). ................................ 3-57 Figure 3-66. Key Dimensions for Assessing Confined Jet Behavior ...................................................... 3-58 Figure 3-67. Confined Jet Facility Apparatus and Comparison to AP1000 Plant Geometrical Configuration ................................................................................................................. 3-60 Figure 3-68. Reference 3-17 Data ........................................................................................................... 3-61 Figure 3-69. Reference 3-18 Test Configuration .................................................................................... 3-63 Figure 3-70. Static Pressure Fields at Impingement and Confining Plate .............................................. 3-64 Figure 3-71. Effect of Unconfined and Confined Jet Impingement on Nusselt Number ........................ 3-65 Figure 3-72. Reference 3-19 Experimental Setup ................................................................................... 3-66 Figure 3-73. Results of Reference 3-19 Testing for Confined and Unconfined (Normal) Jet Wall Pressure Distribution .................................................................................................................... 3-67 Figure 3-74. Structure of Underexpanded Jet ......................................................................................... 3-69 Figure 3-75. Reference 3-16 Shock Wave Formation within a Pipe ....................................................... 3-70 Figure 3-76. Shlieren Photograph of Impingement of an Underexpanded Jet ........................................ 3-71 Figure 3-77. Excerpt from Reference 3-32 ............................................................................................. 3-72 Figure 3-78. Dimensionless velocity profile for Region IV jet as compared to slot jet without impingement plate.......................................................................................................... 3-73 Figure 3-79. NUREG/CR-2913 Target Pressure Distributions ............................................................... 3-74 Figure 3-80. Schematic of Autoclave Test Vessel ................................................................................... 3-79 Figure 4-1. Circumferential Break with Limited Separation .................................................................... 4-7 Figure 4-2. Geometry of Circumferential Break with Limited Separation ............................................... 4-7 WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 x Figure 4-3. Primary Coolant Loop Break Locations (Reference 4-7)..................................................... 4-10 Figure 4-4. RCS Cold Leg Steam Generator Compartment Break Locations ........................................ 4-11 Figure 4-5. RCS Cold Leg Nozzle Gallery Break Locations .................................................................. 4-12 Figure 4-6. RCS Cold Leg CA01 Module Penetration Break Location.................................................. 4-13 Figure 4-7. RCS Hot Leg Break Locations ............................................................................................. 4-14 Figure 4-8. Limited Separation (A) and Full Separation (B) Break Geometries .................................... 4-16 Figure 5-1. Centerline Stagnation Pressure at [ ]a,c from the Jet Nozzle ..................................... 5-3 Figure 5-2. Plot of NEI 04-07 Spherical ZOI Radii .................................................................................. 5-4 Figure 5-3. Nozzle Gallery Plan View ...................................................................................................... 5-5 Figure 5-4. Nozzle Gallery Elevation View at the DVI Piping Centerline ............................................... 5-5 Figure 5-5. Nozzle Gallery Elevation View at the Hot Leg Piping Centerline ......................................... 5-6 Figure 5-6. Nozzle Gallery Elevation View at the Cold Leg Piping Centerline ....................................... 5-6 Figure 5-7. Elevation View of DVI Break Zone of Influence [ ]a,c ............................................ 5-7 Figure 5-8. Plan View of DVI Break Zone of Influence [ ]a,c ...................................................... 5-8 Figure 5-9. Elevation View of Hot Leg Break Zone of Influence [ ]a,c ...................................... 5-9 Figure 5-10. Plan View of Hot Leg Break Zone of Influence [ ]a,c ............................................. 5-9 Figure 5-11. Elevation View of Region I Analysis Cold Leg Break Zone of Influence [ ]a,c .... 5-10 Figure 5-12. Plan View of Region I Analysis Cold Leg Break Zone of Influence [ ]a,c ............. 5-11 Figure 5-13. CA31 Module and Neutron Blocks Top View; Top Liner Plate Removed ..................... 5-12 Figure 5-14. Elevation View of Cold Leg Break Zone of Influence [ ]a,c ................................. 5-12 Figure 5-15. Plan View of Cold Leg Break Zone of Influence [ ]a,c......................................... 5-13 WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 xi LIST OF ACRONYMS, ABBREVIATIONS, AND TRADEMARKS ADS automatic depressurization system ANS American Nuclear Society ANSI American National Standards Institute AP1000 Registered trademark of Westinghouse Electric Company LLC ASME American Society of Mechanical Engineers BWROG Boiling Water Reactor Owners Group CL cold leg COL Combined Operating License CT cable test DBA design basis accident DCD Design Control Document DECLG double-ended cold-leg guillotine DEGB double-ended guillotine break DVI direct vessel injection EQ equipment qualification FSAR Final Safety Analysis Report FT facility test GL Generic Letter GSI Generic Safety Issue HL hot leg ID inner diameter IRWST in-containment refueling water storage tank LOCA loss-of-coolant accident LNS lower neutron shielding LV low voltage MRI metal reflective insulation MV medium voltage NMI non-metallic insulation NRC Nuclear Regulatory Commission NTS National Technical Services PIRT Phenomena Identification and Ranking Table PVC polyvinylchloride PWR pressurized water reactor PWROG Pressurized Water Reactor Owners Group PWSCC primary water stress-corrosion cracking PXS passive core cooling system RAI request for additional information RCP reactor coolant pump RCS reactor coolant system RES Office of Nuclear Regulatory Research RNS normal residual heat removal system RPV reactor pressure vessel RVIS reactor vessel insulation system SE safety evaluation WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 xii LIST OF ACRONYMS, ABBREVIATIONS, AND TRADEMARKS (cont.)

SS stainless steel STC Westinghouse Science and Technology Center Type I Upper neutron shield; [***********]a,c Type II upper neutron shield; [*********************** **]a,c Type III upper neutron shield; [***********]a,c UFSAR updated Final Safety Analysis Report UNS upper neutron shield ZOI zone of influence ZOIr zone of influence radius WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 xiii EXECUTIVE

SUMMARY

The AP10001 plant safety evaluation addressing Generic Safety Issue (GSI)-191, Assessment of Debris Accumulation on PWR Sumps Performance (Reference 1), and Generic Letter (GL) 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors (Reference 2), states that fibrous debris will not be generated during a loss-of-coolant accident (LOCA). This determination is documented in the licensing basis in subsection 6.3.2.2.7.1 of APP-GW-GL-700, AP1000 Design Control Document (Reference 3), which states that a LOCA in the AP1000 does not generate fibrous debris due to damage to insulation or other materials included in the AP1000 design. The design of the AP1000 plant includes some cabling in close proximity to high-energy lines such that they may be directly impinged upon by a postulated LOCA jet.

Additionally, the AP1000 plant reactor vessel insulation system (RVIS) has neutron shield blocks surrounding the reactor pressure vessel (RPV) encapsulating non-metallic [********** ]a,c designed to be a suitable equivalent to metal reflective insulation (MRI). This report documents that the radius of the zone of influence (ZOI) for cables is defined as four inside pipe diameters, 4D, that non-metallic insulation (NMI) contained within the neutron shield blocks is a suitable equivalent to MRI, and that the effective ZOI diameter determination methodology is appropriate for use in GSI-191 debris source term analyses.

CABLES Westinghouse completed a cable jet impingement test program at National Technical Systems (NTS, formerly Wyle Laboratories) in Huntsville Alabama to assess the potential of cable being directly impinged upon by a LOCA jet and becoming a GSI-191 debris source. Cable jet impingement testing was performed since a material zone of influence radius (ZOIr) did not exist for cables or their constituent components. Establishment of a cable ZOI enabled Westinghouse to address potential debris from cables with respect to the AP1000 plant licensing basis requirements.

The cable jet impingement test program utilized 5 different types of AP1000 plant cables that bound all cable material design criteria and specifications for cables that will be utilized in the AP1000 plant containment. Jet impingement tests were performed on both fresh and aged cables at LOCA conditions representative of the AP1000 plant. The aged cables were subjected to conditions that bound 60 years of AP1000 plant operation.

The test program was developed and undertaken with the following objectives:

• To assess the different cable types that bound in-containment cable materials under postulated large-break LOCA jet

• To identify the onset of damage when a cable is exposed to postulated large-break LOCA jet conditions 1

. AP1000 is a registered trademark of Westinghouse Electric Company LLC, its affiliates and/or its subsidiaries in the United States of America and may be registered in other countries throughout the world. All rights reserved. Unauthorized use is strictly prohibited.

WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 xiv

• To establish a cable ZOI based on the onset of incipient damage In total, 10 cable tests were performed meeting all requirements for a valid test. The results of the cable tests showed that:

• The threshold for incipient damage from the blowdown jet for all cable types was [****]a,b,c, where D is the equivalent inner diameter of the LOCA pipe as defined by the debris generation break size assessment.

• All cables remained undamaged from the blowdown jet at [**]a,b,c.

The AP1000 plant cable jet impingement test program identified the material-specific performance of the AP1000 plant cables and identified where the onset of damage occurs when the cables are exposed to representative LOCA conditions.

The cable jet impingement tests clearly show the transition between:

• [******************************************************************** ]a,b,c2

• [*****************************************************]a,b,c

• [************************************************************ ]a,b,c Based on the results of the cable jet impingement test program that consisted of 5 different types of AP1000 plant cables that bound all cable material design criteria and specifications, a ZOI of 4D (as defined by the jet impingement testing) is conservatively applied to AP1000 plant in-containment cables that may be directly impinged upon by a LOCA jet. In conclusion, all AP1000 in-containment cables are bounded by those tested and analyzed; cables outside this 4D ZOI will not generate debris under postulated LOCA conditions.

NEUTRON SHIELD BLOCKS To ensure the NMI contained in neutron shield blocks meets the definition of a suitable equivalent insulation, Westinghouse developed a test program that included jet impingement testing at NTS and submergence testing at the Westinghouse Churchill Science and Technology Center (STC). The results of the jet impingement test programs led to changes in the neutron shield block design such that the encapsulation is now [**************]a,c. In addition to [ ************

]a,c from some neutron shield block applications as discussed in Section 2.

Jet impingement tests were performed on pairs of neutron shield blocks with the objective to assess the robustness of the neutron shield block construction to ensure the encapsulated NMI met the suitable equivalency requirements. Fully exposed blocks were tested by eliminating all of the surrounding 2

L is the distance from the nozzle exit to the target. D is the inside diameter of the broken pipe. With respect to jet impingement testing, [*********]a,c.

WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 xv intervening structures; however [

• • • • ]a,b,c.

Each neutron shield block subjected to jet impingement testing consists of three main components:

• [*************************************************** ****]a,c

• [************************************************************* *** ]a,c

• [**

• • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • *]a,c The jet impingement test program tested pairs of blocks in the following three configurations of the neutron shield block:

• [****************************************************** **]a,c

• [*****************************************************

]a,b,c

• [*****************************************************************************

• • • • • • • *]a,b,c In total, seven neutron shield block tests were performed on pairs of blocks at the jet impingement test facility; these tests met all system requirements for valid tests.

Type I neutron shield blocks tested at [**** *]a,c were undamaged by the jet.

Neutron shield blocks tested at [*****]a,c showed that:

• [*

• • • • • • • • *]a,b,c

• [*

• • • • • • • • • • • • • • • • • • • • • • • **]a,b,c

• [*****************************************************************************

• • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • **]a,b,c The results of the Type III neutron shield blocks tested at [*****]a,b,c;3 showed that:

3

[********************************************************** ********]a,b,c WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 xvi

• [

• X x

• X**********************************]a,b,c For the Type III design tested at [

• • • • • • • • • • • • • • • • • • • • • • • • • • • • • *]a,b,c In all cases, there was [******************************************* ***]a,b,c In parallel with neutron shield block jet impingement testing, Westinghouse completed a neutron shield block component submergence test program to assess the chemical implications of neutron shield blocks in the floodup zone of the AP1000 plant. This program involved a 30-day test with a Type I [ *

• • • • ]a,b,c in fluid representative of the expected post-accident conditions. The submergence test program showed that the [

]a,b,c, it has been concluded that there are no physical or chemical debris implications, making the NMI within the [********* *****] a,b,c neutron shield blocks a suitable equivalent to MRI for the locations bounded by testing and analysis.

It should be noted that:

• [

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Westinghouse Non-Proprietary Class 3 xvii

• • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • **]a,b,c The NMI contained within the [***************]a,b,c neutron shield blocks meets the definition of a suitable equivalent insulation provided in the AP1000 plant licensing basis, the AP1000 Design Control Document (DCD) (Reference 3), and does not contribute to the AP1000 plant debris source term. The NMI contained within the [***************]a,b,c neutron shield blocks may therefore be used in the AP1000 plant as a suitable equivalent to MRI for this application.

In conjunction with the cable and neutron shield block test programs, Westinghouse implemented the alternate evaluation methodology described in NEI-04-07 demonstrating acceptable containment sump performance. This methodology and the companion safety evaluation, NEI-04-07 Pressurized Water Reactor Sump Performance Evaluation Methodology; Volume 1 (Reference 4) and NEI 04-07, Pressurized Water Reactor Sump Performance Evaluation Methodology; Volume 2 - Safety Evaluation by the Office of Nuclear Reactor Regulation Related to NRC Generic Letter 2004-02 (Reference 5) have been applied to the AP1000 plant to determine debris generation break sizes for potential debris sources.

REFERENCES

1. GSI-191, Assessment of Debris Accumulation on PWR Sumps Performance, Footnotes 1691 and 1692 to NUREG-0933, 1998, Nuclear Regulatory Commission, May 14, 1997.

2. NRC Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors, September 13, 2004.

(U.S. Nuclear Regulatory Commission ADAMS Accession No. ML042360586).

3. APP-GW-GL-700, Revision 19, AP1000 Design Control Document, June 2011.

4. NEI 04-07, Pressurized Water Reactor Sump Performance Evaluation Methodology; Volume 1, Revision 0, December 6, 2004.

5. NEI 04-07, Pressurized Water Reactor Sump Performance Evaluation Methodology; Volume 2 -

Safety Evaluation by the Office of Nuclear Reactor Regulation Related to NRC Generic Letter 2004-02, Revision 0, December 6, 2004.

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Westinghouse Non-Proprietary Class 3 1-1 1 INTRODUCTION The AP1000 plant safety evaluation addressing GSI-191, Assessment of Debris Accumulation on PWR Sumps Performance (Reference 1-1) and GL 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors (Reference 1-2) accounts for zero LOCA-generated debris from insulation or cables. This is documented in the licensing basis in subsection 6.3.2.2.7.1 of APP-GW-GL-700, AP1000 Design Control Document (Reference 1-3), which states that a LOCA in the AP1000 does not generate fibrous debris due to damage to insulation or other materials included in the AP1000 design. This is based on the use of MRI or a suitable equivalent and the elimination of fibrous insulation and other sources of fiber.

The AP1000 plant design includes encapsulated NMI in the reactor cavity that is designed to be a suitable equivalent 1 to MRI. The encapsulated NMI is contained within the RVIS lower neutron shielding (LNS),

the water inlet doors, and the refueling cavity floor module (CA31) neutron shield blocks. Additionally, the AP1000 plant design includes cabling that may contain fibrous and other materials (jackets, wrappings, and filler materials) that may be directly impinged upon by a LOCA jet.

To address the potential for LOCA-generated debris from non-metallic materials in the reactor cavity and cables that may be directly impinged upon by a LOCA jet, Westinghouse embarked on a jet impingement test program at NTS to qualify the encapsulated NMI as a suitable equivalent to MRI and to define a ZOI for cables.

1.1 PURPOSE Westinghouse developed an extent of condition program to evaluate any potential impacts to the current licensing basis from the exposure of cables to direct jet impingement by a LOCA jet and to qualify the encapsulated NMI in the reactor cavity as a suitable equivalent to MRI. The purpose of the program was to define a cable ZOI and to confirm that the encapsulated NMI meets the requirements of suitable equivalency and may be used in place of MRI. The program included jet impingement testing of in-containment cables and neutron shield blocks containing encapsulated NMI, and submergence testing of neutron shield blocks.

1

. A suitable equivalent insulation is one that is encapsulated in stainless steel that is seam welded, so that LOCA jet impingement does not damage the insulation and generate debris. Another suitable insulation is one that may be damaged by LOCA jet impingement as long as the resulting insulation debris is not transported to the containment recirculation screens, to the in-containment refueling water storage tank (IRWST) screens, or into a DVI or a cold leg LOCA break that becomes submerged during recirculation. In order to qualify as a suitable equivalent insulation, testing must be performed that subjects the insulation to conditions that bound the AP1000 plant conditions and demonstrates that debris would not be generated. If debris is generated, testing and/or analysis must be performed to demonstrate that the debris is not transported to an AP1000 plant screen or into the core through a flooded break. It would also have to be shown that the material used would not generate chemical debris.

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Westinghouse Non-Proprietary Class 3 1-2 The purpose of this topical report is to obtain Nuclear Regulatory Commission (NRC) approval for the following items:

• A ZOI of 4D is applicable to all AP1000 plant in-containment cabling bounded by testing and analysis presented in this topical report.

• The NMI in the [**************]a,b,c RVIS LNS and water inlet doors, and the NMI in the CA31 module neutron shield blocks is a suitable equivalent to MRI for the locations bounded by testing and analysis presented in this topical report.

• The use of NEI-04-07 Pressurized Water Reactor Sump Performance Evaluation Methodology; Volume 1 (Reference 1-4) and NEI 04-07, Pressurized Water Reactor Sump Performance Evaluation Methodology; Volume 2 - Safety Evaluation by the Office of Nuclear Reactor Regulation Related to NRC Generic Letter 2004-02 (Reference 1-5) alternative methodology for defining debris generation break size for postulated accidents in the AP1000 plant.

The information within this document provides the background and justification that supports this request.

1.2 LIMITS OF APPLICABILITY The results and conclusions of the cable and neutron shield block test programs presented in this document are applicable to all AP1000 plants with cable and encapsulated NMI designs as described in Section 2 of this report.

The conclusions of this topical report are not intended to be used for any nuclear plant designs other than the AP1000 plant.

1.3 REFERENCES

1-1. GSI-191, Assessment of Debris Accumulation on PWR Sumps Performance, Footnotes 1691 and 1692 to NUREG-0933, 1998, Nuclear Regulatory Commission, May 14, 1997.

1-2. NRC Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors, September 13, 2004. (U.S. NRC ADAMS Accession No. ML042360586).

1-3. APP-GW-GL-700, Revision 19, AP1000 Design Control Document, June 2011.

1-4. NEI 04-07, Pressurized Water Reactor Sump Performance Evaluation Methodology; Volume 1, Revision 0, December 6, 2004.

1-5. NEI 04-07, Pressurized Water Reactor Sump Performance Evaluation Methodology; Volume 2 -

Safety Evaluation by the Office of Nuclear Reactor Regulation Related to NRC Generic Letter 2004-02, Revision 0, December 6, 2004.

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Westinghouse Non-Proprietary Class 3 2-1 2 POTENTIAL SOURCES OF ADDITIONAL DEBRIS 2.1 CABLES The AP1000 plant safety evaluation addressing GSI-191 (Reference 2-1) and GL 2004-02 (Reference 2-2) accounts for zero LOCA-generated fibrous debris. The AP1000 plant design includes cabling that may be directly impinged upon by a LOCA jet. These cables may contain fibrous and other materials (jackets, wrappings, and filler materials) that were not considered in the initial GSI-191 debris source term evaluation.

Jet impingement testing and component characterization performed on AP1000 plant cables that may be directly impinged upon by a LOCA jet is provided in Section 3.3.3.

Submergence testing of cables was not necessary since submerged cables have been a part of the AP1000 plant design since its inception and have been dispositioned as having a negligible chemical effects, as discussed in the Letter to the Honorable Gregory B Jaczko, Chairman, NRC, from Said Abdel-Khalik, Chairman, ACRS, dated December 20, 2010, Long-Term Core Cooling For The Westinghouse AP1000 Pressurized Water Reactor (Reference 2-3). Additionally, the NRC performed a phenomena identification and ranking table (PIRT) evaluation on post-LOCA chemical effects (Phenomena Identification and Ranking Table Evaluation of Chemical Effects Associated with Generic Safety Issue 191, NUREG-1918, February 2009, Reference 2-4), which concluded that the consequences of chemical effects based on cable debris are small when compared with other sources of chemical effects. The main concern with submerged cables is the formation of chloride ions due to radiolytic breakdown of the cable insulation. For many plants this may be more or less important based on the amount of polyvinylchloride (PVC) cable jacket and insulation used since the chloride in PVC is the dominant source of the chloride ions via radiolysis. The chloride ions bond with hydrogen, creating acids.

In sufficient quantities acids may affect corrosion of certain materials and can have an effect on the sump pH value. The impact of chlorides was simulated in the integrated chemical effects program (Reference 2-5). Cable jacket and insulation used in the AP1000 plant is constructed of cross-linked polyolefin and cross-linked polyethylene, respectively, and does not have a major chloride component.

Additionally, APP-G1-E1-002 (Reference 2-6) specifically states that [******

]a,b,c. Final Report Evaluation of Chemical Effects Phenomena in Post-LOCA Coolant, NUREG/CR-6988 March 2009 (Reference 2-6) concludes that even at maximum anticipated concentrations, chloride containing compounds (chlorate, hypochlorite, and hypochlorous acid sourced from electric cable insulation would have a negligible impact on the pH in the RCS. Based on these conclusions, no further work is required to address chemical effects from submerged cables for the AP1000 plant.

2.1.1 Description of the Cables Westinghouse obtained five types of AP1000 plant cables (Table 2-1, Figure 2-1) for the cable jet impingement test program including low voltage (LV) jacketed insulated single conductor cables, LV jacketed multi-conductor cables, and medium voltage (MV) jacketed power cables. The cables utilized in the test met all wire and cable design criteria (Reference 2-6) and all cable design specifications (Reference 2-8, Reference 2-9, Reference 2-10). The configuration of the tested cable may be different from production cables, e.g., the number of conductors within the cable or the size of the WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 2-2 conductor may be different, but the materials used in construction are identical. The AP1000 plant cable specification includes cables that are comprised of [*******************

]a,c The cables used in the AP1000 plant jet impingement test program were procured from cables produced for in-containment equipment qualification (EQ) testing. The five types of cables tested bound all cable material design criteria and specifications for cables that will be utilized in the AP1000 plant containment. The five types of cables tested and the results reported herein bound all AP1000 plant in-containment cables.

Table 2-1. Cable Jet Impingement and Characterization Test Program Samples (Figure 2-1) a,c a,c Figure 2-1. Cables used in the Cable Jet Impingement Test Program 2.2 REACTOR VESSEL CAVITY NON-METALLIC INSULATION The AP1000 plant safety evaluation addressing GSI-191 (Reference 2-1) and GL 2004-02 (Reference 2-2) accounts for zero LOCA-generated debris from insulation. This is based on the use of MRI or a suitable equivalent.

NMI is used in the AP1000 plant RVIS and the CA31 reactor cavity floor structural module neutron shields. Radiation shielding is located in [

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Westinghouse Non-Proprietary Class 3 2-3

• • • • • • • • • • • • • • • • • • • • • ]a,c.

In addition to radiation shielding, the RVIS has MRI attached to the reactor vessel and buoyant water inlet doors that are designed to allow water free access to the region between the reactor vessel and the MRI to promote in-vessel retention following severe accidents. These doors are made of [

]a,c.

All NMI in the reactor cavity is located below the LOCA floodup level of 110-2 and has the potential to be fully submerged. The neutron shielding located in CA31 is in close proximity to the reactor coolant loop and DVI piping. Potential debris from the reactor cavity NMI in the RVIS was not considered in the licensing basis since only suitable equivalent insulation to MRI is allowed in the AP1000 plant containment. To achieve the goal of suitable equivalency for the RVIS and CA31 NMI, jet impingement testing and component characterization was performed on AP1000 plant NMI to assess the potential of NMI to become a debris source following a design basis LOCA as discussed in Section 3.3.4.

Additionally, submergence testing and component characterization of NMI was performed since the RVIS is also below the maximum containment flood-up level following a design basis LOCA. An evaluation of NMI submergence testing is provided in Section 3.4.

2.2.1 Description of the RVIS Water Inlet Doors The RVIS water inlet doors are located in the lower reactor cavity at the 71-6 plant elevation (Reference 2-11). These doors are required to be buoyant so that they will float open in the event the reactor cavity floods and allow water into the annulus between the reactor vessel and the reactor vessel insulation for in-vessel retention following a core melt sequence (beyond design basis event). The doors are made with [x x ]a,c .

2.2.2 Description of the RVIS Lower Neutron Shielding The RVIS LNS is located in the lower reactor cavity (78'-0" elevation). The LNS consists of [

x x

x ] a,c The LNS is constructed by fabricating a [

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Westinghouse Non-Proprietary Class 3 2-4

]a,c The individual shielding blocks are positioned in a ring formation in the annulus between the reactor cavity walls (CA04 module) and the reactor vessel and are supported by a cantilevered structure. The LNS blocks are cooled on all sides by the containment ventilation system and their location at the lower end of the reactor pressure vessel near the lower plenum ensures that their temperature remains at approximately 150ºF during normal operation.

2.2.3 Description of the CA31 Neutron Shielding The neutron shield blocks that are part of the CA31 module are located at the 105-2 plant elevation, directly below the reactor vessel closure flange (Reference 2-13). These blocks reduce the amount of radiation streaming upwards into containment. The CA31 neutron shield blocks [

]a,c The CA31 neutron shield blocks are constructed using [

x ] a,c The CA31 neutron shielding system is shown in Figure 2-4.

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Westinghouse Non-Proprietary Class 3 2-5 a,c Figure 2-2. CA31 Neutron Shield Block (typical) a,c Figure 2-3. CA31 Supplemental Neutron Shield Block (typical)

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Westinghouse Non-Proprietary Class 3 2-6 a,c Figure 2-4. CA31 Shield Block System (typical)

2.3 REFERENCES

2-1. GSI-191, Assessment of Debris Accumulation on PWR Sumps Performance, Footnotes 1691 and 1692 to NUREG-0933, 1998, Nuclear Regulatory Commission, May 14, 1997.

2-2. NRC Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors, September 13, 2004. (U.S. NRC ADAMS Accession No. ML042360586).

2-3. Letter to the Honorable Gregory B Jaczko, Chairman, NRC, from Said Abdel-Khalik, Chairman, ACRS, dated December 20, 2010, Long-Term Core Cooling For The Westinghouse AP1000 Pressurized Water Reactor, (U.S. NRC ADAMS Accession No ML103140348).

2-4. NUREG-1918, Phenomena Identification and Ranking Table Evaluation of Chemical Effects Associated with Generic Safety Issue 191, February 2009.

2-5. NUREG/CR-6914, Integrated Chemical Effects Test Project: Consolidated Data Report, December 2006.

2-6. APP-G1-E1-002, Wire and Cable Design Criteria.

2-7. NUREG/CR-6988, Final Report - Evaluation of Chemical Effects Phenomena in Post-LOCA Coolant, March 2009.

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Westinghouse Non-Proprietary Class 3 2-7 2-8. APP-EW40-Z0-001, Design Specification for Medium Voltage Power Cables for Various Systems.

2-9. APP-EW50-Z0-001, Class 1E Low Voltage 600V Power Cables.

2-10. APP-EW21-Z0-002, Class 1E Instrumentation and Thermocouple Extension Cables.

2-11. APP-MN20-V2-101, RV Bottom Head Insulation Layout RBH1.

2-12. APP-MN20-V2-148, RV Lower Neutron Shielding.

2-13. APP-CA31-S5-001, Containment Building Areas 1, 2, 3 & 4 Module CA31 EL 107-2 Isometric View.

2-14. APP-CA31-GEF-024, Modifications to CA31 Shield Blocks.

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Westinghouse Non-Proprietary Class 3 3-1 3 GSI-191 TEST PROGRAM

SUMMARY

A series of tests was conducted in order to better understand the effects of jet impingement and submergence on materials that are currently in question for debris production in the AP1000 plant. These tests included the following:

• Cable jet impingement testing

• Reactor vessel insulation jet impingement testing

• Reactor vessel insulation submergence testing 3.1 JET IMPINGEMENT TEST BACKGROUND The jet impingement test program was initiated with two goals: establish a defensible cable ZOI based on the onset of incipient cable damage and to establish the reactor cavity encapsulated NMI as a suitable equivalent to MRI from a debris production standpoint.

In order to show that an insulation type qualifies as a suitable equivalent to MRI, the AP1000 plant DCD (Reference 3-1) states that a suitable equivalent insulation is one that is encapsulated in stainless steel that is seam welded so that LOCA jet impingement does not damage the insulation and generate debris.

Another suitable insulation is one that may be damaged by LOCA jet impingement as long as the resulting insulation debris is not transported to the containment recirculation screens, to the IRWST screens, or into a DVI or a cold leg LOCA break that becomes submerged during recirculation, and that testing must be performed that subjects the insulation to conditions that bound the AP1000 plant conditions and demonstrates that debris would not be generated. The only form of testing that could subject the reactor cavity NMI to conditions that bound the AP1000 plant is jet impingement testing performed at bounding conditions. The jet impingement testing focused only on the RVIS type blocks, as this was the most limiting configuration at the time of testing.

The RVIS LNS and water inlet doors are located much farther from a potential pipe break than CA31. As the LNS and water inlet doors are significantly farther away from any postulated breaks, testing blocks of similar construction bounds these configurations. The test specimen also bounds the CA31 shield blocks because of the increased minimum thickness [********************]a,c external stainless steel when compared to the test specimen [************* *****]a,b,c Since the LNS is constructed with [

))a,c minimum thickness stainless steel that is thicker than the [********]a,c stainless steel test specimen, the results from the test specimen can be used to draw conclusions about the behavior of the LNS and CA31 shield blocks, and the water inlet doors.

The method for resolving GSI-191 debris generation involves identifying specific materials, defining ZOIs for those materials, and determining how much debris may be generated by those materials. A review of the current GSI-191 documentation discovered no data on the ZOI associated with cabling.

Many of the ZOIs that are used by the nuclear industry to resolve GSI-191 were referenced from Boiling Water Reactor Owners Group (BWROG) testing, NEDO-32686-A, Volumes 1 through 4, Utility Resolution Guidance for ECCS Suction Strainer Blockage (Reference 3-2) performed using air jets at a pressure of about 1100 psia and testing funded by the NRC performed by Ontario Power, N-REP-34320-10000, Jet Impact Tests - Preliminary Results and Their Application (Reference 3-3),

using saturated water jets at a pressure of about 1500 psia. Both test programs used jets with lower initial WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 3-2 pressures and different fluid conditions that are not representative of pressurized water reactors (PWRs) that operate with subcooled water at a nominal pressure of 2250 psia. Both the BWROG and Ontario Power tests were cited in Reference 3-4, NEI-04-07 Pressurized Water Reactor Sump Performance Evaluation Methodology; Volume 1 and were used by the NRC as the basis for reducing the damage pressures for all material types characterized with air jet testing by 40 percent for PWR applications. The reduction in damage pressure to account for the potentially enhanced debris generation from a two-phase PWR jet resulted in an increased ZOI for all materials listed in NEI 04-07, Pressurized Water Reactor Sump Performance Evaluation Methodology; Volume 2 (Reference 3-5). The reduction in damage pressure was reported as being based, in large part, on the lack of applicable test data for those materials at PWR conditions.

To address the uncertainties in formulating ZOIs for specific materials based on non-prototypical conditions, a group of licensees within the Pressurized Water Reactors Owners Group (PWROG) chose to pursue jet impingement testing to establish technically defensible ZOI values for materials of interest within their plants. The program had the following objectives:

• Perform jet impingement testing using subcooled water at PWR nominal pressure and temperature

• Perform jet impingement testing to determine the ZOI for the materials and designs of interest to the participating licensees The PWROG tried twice to perform jet impingement testing at typical PWR conditions to decrease the ZOI of in-containment materials only to have both programs rejected because of a facility design flaw that placed the choke point of the system upstream of the exit nozzle instead of at the exit nozzle. With the choke point upstream of the exit nozzle, the program was called into question as to whether the jet was fully formed when it hit the target.

To address the upstream choke issue, the PWROG redesigned the blowdown facility and met the following objectives in its most recent jet impingement test program:

• Designed a new discharge nozzle to ensure that the choke point was at the nozzle exit and that the resulting jet would be fully formed

• Confirmed by inspection that the limiting flow area of the system was the discharge nozzle

• Performed instrumented tests to collect data to support the development of a subcooled jet expansion model to demonstrate the test jet was fully expanded

• Performed jet impingement testing using subcooled water at near PWR nominal pressure and temperature

• Performed jet impingement testing to determine the ZOI for the materials and designs of interest to the participating licensees WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 3-3 The results of the PWROG jet impingement testing reported in FAI/110497, PWROG Model for the Two Dimensional Free Expansion of a Flashing, Two Phase, Critical Flow Jet (Reference 3-6) showed that the NTS facility was capable of producing a subcooled jet that was representative of the range of temperatures and pressures associated with a PWR large-break LOCA.

For the neutron shield block and cable jet impingement test program, the test loop was walked down and inspected to confirm that the limiting flow area was in fact the discharge nozzle, ensuring that the resulting jet would be a fully formed representative LOCA jet. The data collected by the PWROG and documented in Reference 3-6 will be shown to be applicable to the neutron shield block and cable jet impingement program.

The approach to qualifying encapsulated NMI as a suitable equivalent to MRI considers an acceptance criterion that allows for some damage to the target material as specified in Reference 3-1:

• A suitable equivalent insulation is one that is encapsulated in stainless steel that is seam welded, so that LOCA jet impingement does not damage the insulation and generate debris; or, if debris is generated, the resulting debris is not transported to the containment recirculation screens, to the IRWST screens, or into a DVI or a cold leg LOCA break that becomes submerged during recirculation. It would also have to be shown that the material used would not generate chemical debris.

The approach to ZOI development for cable considers acceptance criteria that allows for and accounts for some damage to the cable material as seen in damage curves in Appendix II, Confirmatory Debris Generation Analyses provided in Reference 3-5. Using this approach requires both:

• Determining incipient damage (the amount of damage that resulted in the generation of a negligible amount of debris)

• Using an approach to measure the amount of debris generated or clearly define a no damage ZOI For both the neutron shield block and cable jet impingement test programs, a simple process was developed to ascertain the amount of material lost during each test by [

• • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • ]a,c 3.2 JET IMPINGEMENT TEST FACILITY The High Flow Test Facility at NTS shown schematically in Figure 3-1 and as built in Figure 3-2 is based on the High Flow Test Facility setup developed by the PWROG (Reference 3-6).

The facility used for jet impingement testing is capable of simulating the conditions of a high-energy line break within the AP1000 plant containment. The thermal hydraulic conditions (pressure, temperature, and flow) were selected so that conditions associated with a postulated break in the primary piping were accurately simulated, and the data from the experiment will be directly applicable to and bounding of the AP1000 plant. The high energy line break condition will be representative of a cold leg break. The choice of cold leg conditions is in accordance with the SE on NEI 04-07 that concluded that the cold leg break WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 3-4 condition bounds the hot leg break condition with respect to debris generation. To achieve the greatest damage potential the discharge nozzle was designed to assure that the maximum mass flux and momentum flux would be achieved. The cold leg break condition provides the highest mass flow rate, the highest thrust, and thus the highest damage potential for the surrounding structures in the containment.

The cold leg break condition is conservative in regard to subcooling margin and critical mass flux, which are two additional parameters that maximize the damage potential of jet impingement on targets. The parameters associated with the cold leg break applied in the test facility are limiting with respect to the NMI and cables.

3.2.1 Comparison of PWROG Facility with AP1000 Plant Facility Figure 3-3 shows the PWROG as-built blowdown test facility. The major difference between the PWROG facility and the AP1000 plant facility is [

1

]a,b,c As seen in Table 3-1, the major components of the two facilities are identical and a review of the components ensures that the choke point of the facility is at the nozzle exit. The notable differences between the two facilities are that [

]a,b,c The facility instrumentation used in the PWROG program and the AP1000 plant program was recorded in the same locations with one exception: in the PWROG program, temperature and pressure were recorded in the [

• • • • • • • • • • • • • • • • • • • ]a,b,c 1

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Westinghouse Non-Proprietary Class 3 3-5 Table 3-1. Blowdown Facility Components a,b,c WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 3-6 a,b,c Figure 3-1. Schematic of NTS High Flow Blowdown Facility WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 3-7 a,b,c Figure 3-2. As-Built AP1000 Plant High Flow Test Facility Configuration WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 3-8 a,b,c Figure 3-3. PWROG As-Built Facility WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 3-9 a,b,c Figure 3-4. AP1000 Plant High Flow Test Facility Configuration for Jet Impingement Testing WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 3-10 3.2.2 Comparison of PWROG Facility Data with AP1000 Facility Data Subsection 3.2.1 showed that the AP1000 plant blowdown test facility was constructed from the same components as the PWROG blowdown test facility. The following comparisons ensure that the AP1000 plant blowdown test facility reproduced the same conditions and output as the PWROG blowdown test facility.

The PWROG ran nine instrument tests at distances of [ ]a,b,c with two of the tests repeated due to facility anomalies for a total of seven valid instrumented tests (Reference 3-6).

Figure 3-5 through Figure 3-9 show the data recorded during the PWROG test program. As noted in subsection 3.2.1, the instrumentation was identical for both facilities except for the different instrument location noted in subsection 3.2.1 for the [ ]a,b,c. The PWROG tests are identified by run number in the figure legend. Note that the tests are manually terminated at [

]a,b,c.

a,b,c Figure 3-5. PWROG Jet Impingement Test Program A3 Tank Pressure WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 3-11 a,b,c Figure 3-6. PWROG Jet Impingement Test Program Reducer Pressure a,b,c Figure 3-7. PWROG Jet Impingement Test Program Mass Flow Rate WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 3-12 a,b,c Figure 3-8. PWROG Jet Impingement Test Program Exit Nozzle Pressure a,b,c Figure 3-9. PWROG Jet Impingement Test Program Reducer Temperature WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 3-13 Figure 3-10 through Figure 3-14 show a comparison of the data recorded during the PWROG test program and the AP1000 plant cable jet impingement test program runs. The AP1000 plant cable tests are described in the legends by cable test (CT) number and facility test (FT) number (CT#FT#). Note that due to the large number of data points per second of test (approximately 350 data points), only the cable jet impingement test results are shown. A comparison of the neutron shield block jet impingement test results with the PWROG jet impingement test program can be found in WCAP-17616, Jet Impingement Testing of AP1000 Reactor Vessel Insulation System Neutron Shielding Blocks (Reference 3-9).

a,b,c Figure 3-10. PWROG/AP1000 Plant Jet Impingement Test Program A3 Tank Pressure Figure 3-11 and Figure 3-14 highlight the facility difference in the pressure and temperature measurement location. As noted in subsection 3.2.1, the AP1000 plant measured pressure and temperature at [

x ]a,b,c, ensuring applicability of the PWROG test data to the current test program.

WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 3-14 a,b,c Figure 3-11. PWROG/AP1000 Plant Jet Impingement Test Program Reducer Pressure a,b,c Figure 3-12. PWROG/AP1000 Plant Jet Impingement Test Program Mass Flow Rate WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 3-15 a,b,c Figure 3-13. PWROG/AP1000 Plant Jet Impingement Test Program Exit Nozzle Pressure a,b,c Figure 3-14. PWROG/AP1000 Plant Jet Impingement Test Program Reducer Temperature WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 3-16 The comparison of test data from the PWROG instrumented test runs and the AP1000 plant cable tests shows good agreement in all aspects and confirms that the facility is repeatable and that the data recorded in the PWROG instrumented jet impingement test program are directly applicable to the AP1000 plant jet impingement test program.

The following figures (Figure 3-15 through Figure 3-18) show stagnation pressures recorded at the target locations in the PWROG instrumented tests performed at [

]a,b,c a,b,c Figure 3-15. Stagnation Pressure at [ ]a,b,c WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 3-17 a,b,c Figure 3-16. Stagnation Pressure at [ ]a,b,c a,b,c Figure 3-17. Stagnation Pressure at [ ]a,b,c WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 3-18 a,b,c Figure 3-18. Center Line Stagnation Pressure at the Target Based on the data comparisons between the PWROG jet impingement test program and the AP1000 plant jet impingement test program, the stagnation pressures recorded at the target locations in the PWROG jet impingement test program shown in Figure 3-16 through Figure 3-18 were achieved at the corresponding target locations in the AP1000 plant cable jet impingement test program.

3.2.3 Comparison of AP1000 Plant licensing basis with AP1000 Plant facility data Section 3.2.2 showed that the AP1000 plant blowdown test facility, constructed from the same components as the PWROG blowdown test facility, produced data that is in excellent agreement with the PWROG blowdown test facility and that data recorded during the PWROG test program is directly applicable to the AP1000 plant test program.

Table 3-2 shows the nominal conditions for the AP1000 plant and the two jet impingement test programs.

Due to facility constraints, the initial steady state pressure and subcooling are slightly higher in the AP1000 plant. However, given that the jet impingement test programs used for the purposes of establishing a ZOI occurred outside of the core region, this has an insignificant impact on the results.

Table 3-2. AP1000 Plant Conditions vs. Test Conditions a,b,c WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 3-19 Reference 3-6 verified that the [

]a,b,c. To demonstrate that the experimental jet is conservative compared with a large-break LOCA in the full-scale AP1000 plant RCS, Figure 3-19 through Figure 3-21 show a comparison of the licensing basis AP1000 plant large-break LOCA as documented in LTR-LIS-14-339, AP1000 Plant LBLOCA DECLG Data for GSI-191 Jet Impingement Testing Support (Reference 3-8),

and the data recorded for the AP1000 plant jet impingement test programs. All three plots show that

[x x

x x ]a,b,c.

Thus, in all respects, the high-pressure subcooled jet generated at the experimental facility is representative, compared with two-phase critical flow literature, and conservative, compared with a full-scale large-break LOCA in the AP1000 plant.

a,b,c Figure 3-19. Comparison of AP1000 Plant Licensing Basis and the AP1000 Plant Jet Impingement Test Program Stagnation Pressure WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 3-20 a,b,c Figure 3-20. Comparison of AP1000 Plant Licensing Basis and the AP1000 Plant Jet Impingement Test Program Mass Flux a,b,c Figure 3-21. Comparison of AP1000 Plant Licensing Basis and the AP1000 Plant Jet Impingement Test Program Reservoir Pressure WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 3-21 The comparison of the AP1000 plant licensing basis large-break LOCA and the AP1000 plant jet impingement test programs shows that the AP1000 plant cable and NMI jet impingement tests are conservative with respect to the AP1000 plant.

3.3 JET IMPINGEMENT TESTS 3.3.1 Jet Impingement Test Objectives The objective of the AP1000 plant cable jet impingement test program was to perform jet impingement tests on AP1000 plant cables to determine the performance characteristics of the cables and to define the onset of damage when the cables are exposed to representative LOCA jet load conditions for the purpose of defining a cable ZOI.

The objective of the AP1000 plant neutron shield block jet impingement test program was to perform jet impingement tests on AP1000 plant neutron shield blocks to determine the performance characteristics of the neutron shield block encapsulated NMI and characterize damage when exposed to representative LOCA jet load conditions for the purpose of establishing suitable equivalency to MRI.

Data and observations collected from the tests and test specimens were used as follows:

• Determine an appropriate, technically defensible, realistic ZOI for cables

• Determine suitable equivalency for encapsulated NMI The AP1000 plant cable and neutron shield block jet impingement test program included a facility design that accurately and realistically reproduced the phenomena and processes associated with a postulated LOCA blowdown. Once the NTS facility was shown to meet the requirements of the AP1000 plant jet impingement test program by reproducing the results from the PWROG instrumented jet impingement test program, AP1000 plant cables and neutron shield blocks were exposed to the phenomena and processes of a two-phase jet originating from a subcooled, high-pressure, high-temperature reservoir (WCAP-17617-P, Jet Impingement Testing of AP1000 In-containment Cables {Reference 3-7} and WCAP-17616-P, Jet Impingement Testing of AP1000 Reactor Vessel Insulation System Neutron Shielding Blocks {Reference 3-9}). The conditions of interest to which the targets were exposed included elevated temperature and pressure, high mass flux, and subcooling.

3.3.2 Jet Impingement Test Specimens 3.3.2.1 Cable Specimens Westinghouse tested five types of cables (Table 2-1, Figure 2-1) included LV jacketed insulated single conductor cables and LV jacketed multi-conductor cables (defined as small cables; Figure 3-22), and MV jacketed power cables (defined as large cables; Figure 3-23) that were manufactured for the AP1000 plant EQ test program.

To ensure that the life of the plant was represented in the cable jet impingement test program, each cable type was tested both with and without aging to bound cable conditions over 60 years of operation. The WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 3-22 cable aging specification (Reference 3-10) included

• • **]a,c Each cable test was performed with both fresh and aged (thermally and radiologically) cable specimens in each run.

The placement of cables in the AP1000 plant design is governed by specifications and standards as to the types of cable trays, cable tray supports, cable tray installation, cable clamps and ties, and associated fittings and conduits that can be used in the AP1000 plant containment.

The cable arrangement used in the cable jet impingement tests was primarily defined to place cables in the known jet field established in the PWROG instrumented test program (Reference 3-8) for the purpose of defining the cable material destruction pressure at the point of incipient damage to establish the cable ZOI (Reference 3-11).

There was no preheating of the cables prior to jet impingement. The cables as installed in the plant operate at a nominal temperature of [

• • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • ***************************** *]a,c a,b,c Figure 3-22. Small Cables WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 3-23 a,b,c Figure 3-23. Large Cables 3.3.2.2 Neutron Shield Block Specimens Each neutron shield block design tested consists of three main components (Figure 3 24):

• [**************************************************************** *]a,c

• [*************************************************************************

• x ]a,c

• [*****************************************************

x********** *]a,c Three configurations of the neutron shield block test specimens were included in the neutron shield block jet impingement test program:

• Type I test specimens represented the standard design of the neutron shield blocks at the time of testing. [*******************************************************************

• • • • • • • • • • **]a,c (Figure 3-25)

• Type II test specimens represent a modification of the Type I test specimen [*****************

• • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • ****]a,c (Figure 3-26)

• Type III test specimens represent a modification of the Type I test specimen by [*************

• • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • ]a,c (Figure 3-27)

WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 3-24 a,b,c Figure 3-24. Typical Neutron Shield Block Construction a,b,c Figure 3-25. Neutron Shield Block (Type I)

WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 3-25 a,b,c Figure 3-26. [************]a,c Neutron Shield Block (Type II) a,b,c Figure 3-27. [*****************]a,c Neutron Shield Block (Type III)

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Westinghouse Non-Proprietary Class 3 3-26 3.3.3 Cable Jet Impingement Test Summary The purpose of the cable jet impingement test program is to define a ZOI for AP1000 plant in-containment cables.

3.3.3.1 Cable Jet Impingement Test Matrix The cable tests were run in two sequences, large and small, as shown in the test matrix, Table 3-3.

Table 3-3. Cable Jet Impingement Test Matrix a,b,c 3.3.3.2 Cable Jet Impingement Test Acceptance Criteria The goal of cable jet impingement testing is to define a ZOI outside of which cables exposed to a LOCA jet do not contribute to the AP1000 plant debris source term.

The acceptance criteria for cable jet impingement testing are that all facility requirements are met resulting in a successful blowdown of the facility that bounds the licensing basis AP1000 plant large-break LOCA.

3.3.3.3 Post Test Evaluation of Cables Prior to jet impingement testing, all cable specimens are weighed.

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Westinghouse Non-Proprietary Class 3 3-27 Following a successful test, all cables are evaluated for any damage caused by the blowdown jet (rips, tears, or other indications that the cable jacket is breached and material is lost) that will be used in the determination of the cable ZOI.

[

X WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 3-28 x

]a,b,c a,b,c Figure 3-28. Original Cable Test Fixture Design - Cables Behind Fixture a,b,c Figure 3-29. Original Cable Test Fixture Design (Cable Clamp Placed Behind the Holding Plate)

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Westinghouse Non-Proprietary Class 3 3-29 a,b,c Figure 3-30. Cable Damage from Cable/Fixture Interaction (Equivalent Damage at Both Top and Bottom of Window) a,b,c Figure 3-31. Revised Cable Test Fixture that Eliminates Cable/Fixture Interaction by Placing Cable Clamp in Front of the Holding Plate WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 3-30 3.3.3.4 Cable Arrangement in Test Fixture Prior to testing, the cables were assigned positions in the fixture for ease of identification. The arrangement is shown in Table 3-4.

Table 3-4. Cable Arrangement in Test Fixture a,b,c X

x X

x a,b,c Figure 3-32. Large and Small Cable Arrangements in the Test Fixture WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 3-31 3.3.3.5 Cable Jet Impingement Test Results The cable jet impingement test results are presented in Table 3-5.

Table 3-5. Cable Jet Impingement Test Results a,b,c The ZOI as defined in the NEI 04-07 safety evaluation (SE) in Reference 3-5 is based on the production of small fines at the initiation of incipient damage, meaning that a loss of material indicates the point of failure and the initiation of debris generation. The material ZOI is based on incipient damage and is readily seen in Appendix II of the SE (Reference 3-5) for the various materials included in the section.

Appendix II of the SE examines various materials from different fibrous materials to steel encapsulated MRI. Take for example K-Wool in Figure II-6 of Reference 3-5, there is ~4 percent fines production at a jet pressure of 24 psi, which relates to a 5.4D ZOI in Table 3-2 of Reference 3-5. Similarly, calcium WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 3-32 silicate produced as much as 22 percent small fines at its assigned jet pressure and ZOI (Figure II-12 of Reference 3-5).

These examples and review of the other materials presented in Appendix II of Reference 3-5 indicate that the onset of incipient damage (production of a measurable quantity of small fines) is the point at which the material specific ZOI is defined.

The AP1000 plant cables were tested at [****************** ****]a,b,c in the cable jet impingement program. From the results compiled in Reference 3-7 and shown in Table 3-6 and Table 3-7, it is easy to infer where incipient damage occurs. At [

[***************************************************

• • • • • • • • • ))a,b,c For the large cable tests with cables in front of the jet (Table 3-6), damage to the cables at [ [*****

]a,b,c For the small cable tests (Table 3-7), damage to the cables at [ [****************************

• • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • 1**************************************

• ]a,b,c The results of the cable jet impingement test program clearly show that [***********************

• • • • • • • • • • • • • • • ]a,b,c 1

[

]a,b,c WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 3-33 Table 3-6. Large Cable Damage Summary (percent loss of material) a,b,c WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 3-34 Table 3-7. Small Cable Damage Summary (percent loss of material) a,b,c WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 3-35 a,b,c Figure 3-33. Large Cable [***********************]a,b,c a,b,c Figure 3-34. Small Cable [***********************]a,b,c WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 3-36 a,b,c Figure 3-35. Large Cable [*****************]a,b,c a,b,c Figure 3-36. Small Cable [*****************]a,b,c WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 3-37 a,b,c Figure 3-37. Large Cables at [*****************]a,b,c a,b,c Figure 3-38. Small Cables at [*****************]a,b,c WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 3-38 a,b,c Figure 3-39. [***********************]a,b,c (Cable Test CT9FT12) a,b,c Figure 3-40. [**********************]a,b,c (Cable CT10FT13)

WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 3-39 3.3.3.6 Cable Jet Impingement Test Conclusions The AP1000 plant cable jet impingement test program identified the material-specific performance of the AP1000 plant cables and identified where the onset of damage occurs when the cables are exposed to representative LOCA conditions. Meeting these overall objectives allowed the establishment of a material ZOI for the cables tested that can be applied to AP1000 plant cables.

The cable jet impingement tests clearly show the transition between:

• [**************************************************************************]a,b,c

• [**************************************************** ]a,b,c

• [*************************************************************]a,b,c The transitions noted above were shown to be repeatable in both the large and small cable test series.

Based on the results of the cable jet impingement test program, a ZOI of 4D is conservatively applicable to AP1000 plant in-containment cables that may be directly impinged upon by a LOCA jet. It should be noted that there are several cables that cannot be relocated outside of a 4D ZOI. These cables are protected by a number of different protection schemes. Each instance and protection scheme is described in detail in Reference 3-31.

3.3.4 Neutron Shield Block Jet Impingement Test Summary A summary of the neutron shield block jet impingement is provided in the following sections.

3.3.4.1 Neutron Shield Block Jet Impingement Test Acceptance Criteria The acceptance criteria for neutron shield block jet impingement testing are that all facility requirements are met resulting in a successful blowdown of the facility that bounds the licensing basis AP1000 plant large-break LOCA.

3.3.4.2 Neutron Shield Block Jet Impingement Test Results Summary The goal of the neutron shield block jet impingement testing is to assess the behavior of the different configurations (Types) of neutron shield blocks so that the information from jet impingement testing can be used to determine if the standard Type III current design neutron shield block satisfies the definition of suitable equivalent insulation for the AP1000 plant encapsulated NMI or if one of the modified standard designs must be substituted. Damage caused by interaction with the test fixture is evaluated for consideration in the assessment of suitable equivalency. The neutron shield block test results are presented in Table 3-8.

WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 3-40 Table 3-8. Neutron Shield Block Jet Impingement Test Results Summary a,b,c Three neutron shield block configurations were tested. The results of the tests show that:

• The Type I neutron shield block design [**********************************************

• • • • • • ]a,b,c.

• The Type II neutron shield block design [*********************************************

• • • • • • • ]a,b,c.

• The Type III neutron shield block design [** ******************************************

• • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • ]a,b,c.

• The Type III neutron shield block design [******************************* experienced

• • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • ]a,b,c.

Based on the results presented above and shown in Figure 3-41, Figure 3-42, and Figure 3-43, the [****

WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 3-41

• • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • ]a,b,c, enhancing the robustness and durability of the design installed in the AP1000 plant.

a,b,c Figure 3-41. [***********]a,b,c Type I Neutron Shield Block at [*****]a,b,c a,b,c Figure 3-42. [************]a,b,c Type II Neutron Shield Block at [*****]a,b,c WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 3-42 a,b,c Figure 3-43. [ *]a,b,c Type III Neutron Shield Block at [*****]a,b,c a,b,c Figure 3-44. [* *** ********]a,b,c Type III Neutron Shield Block with [

• • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • ]a,b,c WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 3-43 a,b,c Figure 3-45. [*****************]a,b,c Type III Neutron Shield Block at [*****]a,b,c Each neutron shield block test exposed a pair of blocks to the blowdown jet. The results show that each block in the pair received similar damage, providing a measure of repeatability in each test. Additionally, the facility repeatability justified that subsequent jet tests were similar and comparable. Therefore, the use of two blocks in the neutron shield block tests demonstrates the neutron shield block performance repeatability.

[************************************************************************************

• • • • ]a,b,c WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 3-44 a,b,c Figure 3-46. Neutron Shielding Block Test 6 at [*************************** *]a,b,c

[*

• • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • ]a,b,c WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 3-45 a,b,c Figure 3-47. Post-shot Neutron Shield Block Test 6 - [

• • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • ]a,b,c WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 3-46 a,b,c Figure 3-48. Neutron Shield Block Test 6 [******************************************** *** x xxxxxx]a,b,c Neutron Shielding Blocks a,b,c Figure 3-49. Neutron Shield Block Test 6 [*********************************************

• • • • • • • • • • • • • • • • • • • • • • **]a,b,c WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 3-47 a,b,c Figure 3-50. Neutron Shield Block Test 6 [*******************************************

• • • • • • • • • • • • • • • • • • • • • • • ]a,b,c a,b,c Figure 3-51. Neutron Shield Block Test 6 [**************************************

• • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • ] ]a,b,c WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 3-48

[************************************************************************************

• • ]a,b,c WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 3-49 a,b,c Figure 3-52. Neutron Shielding Block Test at [************************]a,b,c a,b,c Figure 3-53. Neutron Shield Block Test 7 - [ *******************************************

x **]a,b,c WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 3-50 a,b,c Figure 3-54. Neutron Shield Block Test 7 - [

• • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • ]a,b,c

[* *****************************************************************************

• • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • ]a,b,c a,b,c Figure 3-55. A and B Neutron Shield Block Test 7 - [****************]a,b,c WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 3-51

[

• • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • ]a,b,c a,b,c Figure 3-56. Neutron Shield Block Test 7 [ ************************************************ x

• • ]a,b,c

[

]a,b,c WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 3-52 a,b,c Figure 3-57. Neutron Shield Block Test 7 [**********************************************

• • • • • • • • • • • • • • • • • • • • • • • • • • • *]a,b,c a,b,c Figure 3-58. Neutron Shield Block Test 7 [*********************************************

• • • • • • • • • • • • • • • • • • • • • • • • • • • • *****]a,b,c WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 3-53 a,b,c Figure 3-59. Neutron Shield Block Test 7 [8 88888888888888888888888888888888888888888888888 -

88888888888888888888888888 ]a,b,c a,b,c Figure 3-60. Neutron Shield Block Test 7 [************************************************

5555555555555555555555555555555]a,b,c WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 3-54 a,b,c Figure 3-61. Neutron Shield Block Test 7 [5555555555555555555555555555555555555555555555 a,b,c 555555555555555555555555555]a,b,c Figure 3-62. Neutron Shield Block Test 7 [555555555555555555555555555555555555555555555555 -

555555555555555555555555555555555555]a,b,c WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 3-55 a,b,c Figure 3-63. Neutron Shield Block Test 7 [55555555555555555555555555555555555555555555555 555555555555555555555555555 55555]a,b,c 3.3.5 Considerations Resulting from Confined Jet Behavior The NMI is located within the reactor vessel cavity. The reactor vessel cavity is a more confined space as compared to other regions of containment. The jet testing of the neutron shield blocks containing encapsulated NMI performed at the NTS facility prototypically recreated a freely expanding jet. The following discussion will develop a model for confined jet behavior from empirical data contained in the open literature to demonstrate that the testing performed at the NTS facility, based on free jet expansion, was conservative and bounding for the NMI located in the reactor vessel cavity (CA31 and LNS). The purpose of this section is to understand the implications of a jet discharging in a confined space on the accepted free expanding jet model, and to compare the confined jet behavior and characteristics to that of the jet impingement testing at the NTS facility.

The NTS facility tested a freely expanding jet. What this means physically is that the jet was allowed to expand into an infinite volume, which corresponds to the outdoor environment. This means there were no discharge volume constraints on the jet behavior. This section will survey the open literature for confined jet empirical data, relate the data to the AP1000 plant configuration, relate the confined jet considerations to develop a prediction of the NMI jet impingement pressure in the plant, and relate to the NTS testing performed at [ ]a,b,c.

3.3.5.1 Description of Flow Regions Figure 3-64 shows the jet regions resulting from jet impingement on a flat plate. The description of the jet regions was taken from Reference 3-15.

WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 3-56 a,b,c Figure 3-64. Depiction of Impinging Jet Regions on a Flat Plate Region 1 Region 1 is the region defined as the jet core and represents the region of flow establishment. It extends from the nozzle exit to the apex of the potential core. In this region pressure is totally recoverable (stagnation region) and the velocity is equal to the exit velocity of the pipe.

Region 2 Region 2 is the region of established flow in the direction of the jet beyond the apex of the potential core.

It is characterized by a dissipation of the centerline jet velocity and by a spreading of the jet in the transverse (radial) direction.

Region 3 Region 3 is the region in which the jet is deflected from the axial direction associated with impingement on the plate or surface. This region is characterized by high turbulence but lower local pressures due to the formation of a recirculation wake region. Essentially, the velocity term in the jet impingement equation is reduced significantly, resulting from localized turbulence generation as a result of jet impingement on the solid structure.

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Westinghouse Non-Proprietary Class 3 3-57 a,c Figure 3-65. Pressure distribution in as a function of L/D (x/B0) and R/D (y/2b).

Region 4 Region 4 is the wall jet region, where the directed flow increases in thickness as the boundary layer builds up along the solid surface. This is a region characterized by boundary layer growth radially from the jet centerline.

Some key takeaways from Reference 3-15 are:

1. The disturbance created by a solid wall on an impinging jet is propagated upstream at a rate equal to the difference between the sonic velocity and the fluid velocity. The corollary of this is as the fluid velocity approaches a Mach (Ma) number equal to 1, the disturbance vanishes and can no longer propagate upstream and affect the jet behavior. Empirical two-phase critical flow models (Moody, Henry-Fauske) assume Ma=1 flows as the choking condition at the break plane and benchmark well to empirical data.

2. It was confirmed empirically that no change in jet surface characteristics occurred for a nozzle to plate spacing of 2D, where D is the nozzle and the corresponding jet diameter at the nozzle exit.

For slot jets this distance was confirmed to be 1D.

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Westinghouse Non-Proprietary Class 3 3-58

3. The core length derived from Prandtls second hypothesis shows the core length is independent of the Re number. This is further substantiated in Reference 3-16, where core lengths are shown to vary with pressure and subcooling.

4. Figure 3-65 shows the pressure distribution away from the stagnation point (as r/z increases) approaches the ambient pressure. This occurs as a result of the velocity decay resulting from turbulence formation (in the wake and boundary layer due to impingement on the wall shown by Region IV).

Figure 3-66 shows key dimensioning of the NMI as it pertains to the cold leg break location and the reactor vessel. These dimensions will be used going forward to relate the test results contained in the open literature to the AP1000 plant geometry. The dimensions in Figure 3-66 were taken from References 3-12 and 3-13.

The dimensioned values in Figure 3-66 correspond to the [

]a,c a,c Figure 3-66. Key Dimensions for Assessing Confined Jet Behavior Table 3-9 shows the key coordinates resolved from Figure 3-66 that will be used to relate the empirical data contained in Reference 3-17 to the AP1000 plant geometry. The red block in Figure 3-66 represents the NMI in the plant configuration. The value of X is the distance from the reactor vessel outer boundary to the weld at the RV nozzle inlet. H is the vertical distance from the cold leg weld at the reactor vessel WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 3-59 inlet nozzle to the top of the NMI. X represents the horizontal distance from the cold leg weld at the reactor vessel inlet nozzle to the nearest horizontal surface of the NMI. With these key dimensions, a relation to the testing contained in the literature reviewed in the following sections can be derived.

Table 3-9. Key NMI Dimensions for Assessing Confined Jet Behavior a,c 3.3.5.2 Review of Confined Jet Literature Reference 3-17 provides empirical data over a range of Reynolds numbers for jet pressure and velocity fields in a confined space. Figure 3-67 shows the experimental setup used to develop the Reference 3-17 data along with a side-by-side comparison of the test geometry (LEFT) to the AP1000 plant geometry (RIGHT).

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Westinghouse Non-Proprietary Class 3 3-60 a,c Figure 3-67. Confined Jet Facility Apparatus and Comparison to AP1000 Plant Geometrical Configuration The left hand side of Figure 3-67 shows the experimental facility setup for determining confined jet behavior. The right hand side is the AP1000 plant geometry. The test apparatus shown in Figure 3-67 is similar to the AP1000 plant geometry. The confinement plate simulates the reactor cavity wall; the impingement plate represents the reactor vessel. The traversing system allowed for both motion in the axial (z) and radial (r) directions. In this manner the pressure ratios in z and r could be determined based on a range of confinements (impingement plate distance to nozzle diameter) ranging from 0.2 to 6. From Table 3-9, the AP1000 plant reactor vessel spacing to cold leg diameter (X'/D) is [ ]a,c; so, the tested ranges encompasses the geometric confinement of the AP1000 plant configuration.

Key takeaways and observations from the Reference 3-17 testing are as follows:

1. For confinements X/D>4, the core length did not change.

2. For impingement plate spacing (X'/D) less than the core length, there is no discernible increase in turbulence intensity along the jet centerline.

3. The local pressure distribution does not exhibit dependence on Re number. This observation is valid for fully turbulent flows. Based on the Reference 3-17 data, this is valid for Re > 30,000.

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Westinghouse Non-Proprietary Class 3 3-61

4. The experimentally determined pressure coefficients are approximately unity for 1 X/D 2 (note: this encompasses the AP1000 plant value of X'/D of [ ]a,c) at an H/D (which is referred to as r/D in Figure 3-68) of zero.

5. The location of the sub-atmospheric (recirculating wake) region migrates toward the centerline with decreasing values of X'/D.

6. The presence of the impingement plate causes the flow to deflect about a jet diameter above the impingement plate (Reference 3-17).

Figure 3-68 shows the variability in pressure coefficient, turbulence intensity, and Nusselt number associated with confined jet behavior. The quantity H/D denoted by the red line in Figure 3-68 corresponds to R/D as related to the AP1000 plant.

a,c Figure 3-68. Reference 3-17 Data For AP1000 plant X'/D value of [ ]a,c, the pressure coefficient, Cp, is approximately zero; this is consistent with the X'/D value of 6, which corresponds to that of the free jet. This provides indication that the localized pressure field is analogous to that of the free jet. The pressure coefficient is defined per Reference 3-17 as:

0

= 1 (Equation 3-1) 02 2

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Westinghouse Non-Proprietary Class 3 3-62

where, P0 = The local impingement plate static pressure Patm = The local atmospheric pressure

= Fluid density U0 = Nozzle exit velocity Physically, Equation 3-1 represents the ratio of the difference in static to atmospheric pressure differential to that of the dynamic pressure. The results of the Reference 3-17 testing indicate that there is no difference between a confined air jet at a confinement distance of 1 to that of the free jet.

There is a difference, however, in the local Nusselt number. The severely confined jet (X'/D=0.25) appears to have a much lower Nusselt number, which indicates the result of the confinement is a negative effect on heat transfer.

Also, plots of the turbulence intensity for X'/D=1 seem to indicate the confinement causes a boundary layer detachment (indicated by the drop in Nusselt number) and a reattachment (second Nusselt peak) as a result of the confinement. The detached boundary region also referred to as the recirculating wake region has a lower Nusselt number (and also a lower heat transfer coefficient), because the boundary layer is not exchanging energy with the wall, effectively retarding the jet heat transfer.

The results of the Reference 3-17 testing imply that there is not a large impact on pressure distribution resulting from the confined jet behavior, but that heat transfer could be affected. A reduction in heat transfer would not challenge the bounding nature of the free jet testing of the NMI facility. This is because the heat transfer to the target does not impact the impingement pressure.

Figure 3-68 shows that there is a very small difference in the pressure coefficient, Cp, of the X'/D=1 value and that it is very similar to that of the X'/D=6 value (Reference 3-17) which corresponds to that of the free jet. This indicates that at a value of X/D=1, the jet pressure field is nearly converged to that of the free jet pressure field. For the AP1000 plant confinement ratio of [888]a,c, this implies the jet pressure field is analogous to that of the free jet.

Reference 3-18 provides additional insight into confined jet behavior and enforces the conclusions from Reference 3-17 that say:

1. The result of confinement can cause the formation of a recirculating wake region and associated boundary layer detachment that can act to degrade the jet heat transfer.

2. The location of the wake region/boundary layer detachment occurs for X'/D <1 and is independent of Re number, provided the flow is turbulent.

Figure 3-69 shows the schematic for the Reference 3-18 test facility. Similar to Reference 3-17, the Reference 3-18 testing included pressure measurement capability, flow measurement capability, pressure oscillations on the impingement plate, and infrared imaging to develop an indication of thermal radiation heat transfer losses from the facility.

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Westinghouse Non-Proprietary Class 3 3-63 Figure 3-69. Reference 3-18 Test Configuration Figure 3-70 shows the static pressure distribution on the top (confining) and bottom plate for the Reference 3-18 testing. Figure 3-70 also shows the free jet static pressure distributions (solid markers).

The AP1000 plant value for R/D is again denoted by the vertical red line.

As can be seen, the free jet tests show a decreasing coefficient with increasing X'/D. This is to be expected and is well substantiated in the literature. This means the pressure drops with increasing L/D.

Some important takeaways from Figure 3-70 are that the confinement at X'/D of 0.25 has a lower pressure coefficient at the stagnation point than that of the confined jet at 0.25 X'/D. This again appears to be true for the free and confined jet comparison at X'/D of 0.38. The higher pressure coefficients of the free jets at X'/D <1 could be indicative of the suppression of the core region discussed previously.

Also, for a confinement of X'/D >1, the wall pressure distributions are nearly identical, indicating that the effects of confinement are negligible at this location. This is substantiated by the Reference 3-17 testing.

There is a small subatmospheric pressure region at X'/D=1, but it is not pronounced as in smaller confinement ratios. Figure 3-68 confirms that at X'/D=6, the pressure coefficients are nearly identical to that of X'/D=1 indicating the jet behavior is analogous to that of the free jet.

Figure 3-71 shows the radial Nusselt number distributions for the unconfined jets, marked by the solid line -and for the confined jets, marked by the dashed line ---. The AP1000 plant value of R/D is denoted by the red vertical line. The confinement does have an impact on the heat transfer characteristics due to the impact the confinement has on turbulence generation. For severe confinements (X'/D <1), the detachment of the momentum boundary layer is seen to occur. For less severe confinements (X'/D 1),

the Nusselt number is nearly identical for the unconfined jets as that of the confined jets. This is substantiated through comparisons to the multiple literature sources presented.

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Westinghouse Non-Proprietary Class 3 3-64 a,c Figure 3-70. Static Pressure Fields at Impingement and Confining Plate A review of the confined jet literature is beginning to yield repetitive conclusions regarding the pressure distribution as a result of confinement. Namely:

• For small confinements (X'/D) < 1, a slight decrease in the pressure coefficient is seen to occur at the stagnation point, and this is postulated to occur due to the suppression of the potential core length resulting from the confinement.

• For small confinements (X'/D) < 1, boundary layer detachment is seen to occur resulting in sub-atmospheric regions indicative of a recirculating wake zone.

• At H/D (H/D for plant equals r/D for literature testing) >2, the pressure coefficient appears to converge to that of the free jet indicating the results of confined jet impingement have dissipated and the jet behavior is indicative of that of the free jet.

• The Re number can impact the Nusselt number distribution, which is well understood in turbulent heat transfer, but does not impact the pressure distribution. The pressure distribution is solely a function of confinement ratio X'/D for fully turbulent jet flow.

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Westinghouse Non-Proprietary Class 3 3-65 a,c Figure 3-71. Effect of Unconfined and Confined Jet Impingement on Nusselt Number Reference 3-19 is a confined jet study that focused on the static wall pressure distribution to help understand the impact of confinement ratio on pressure distribution. Figure 3-72 shows the experimental setup. It is similar to the other test apparatuses in that an impingement plate and a confinement plate are attached to a traversing system to allow for testing at different X'/D values. A pressure tap is used to determine the variability of H/D (r/D) on pressure distribution.

Similar to the reviews of References 3-15, 3-16, 3-17, and 3-18, Reference 3-19 testing concludes:

• The pressure coefficient is not impacted by Re number.

• At X'/D values > 1, the existence of the boundary layer detachment disappears

• The pressure coefficient decreases at the stagnation point as X'/D increases. This is analogous to impingement pressure decreasing with increasing L/D, which is well substantiated. The pressure coefficient decrease converges to a constant value for 1 X'/D 4.

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Westinghouse Non-Proprietary Class 3 3-66 Figure 3-72. Reference 3-19 Experimental Setup Figure 3-73 (Reference 3-19) shows the results of confined jet impingement testing as compared to the results of free jet testing. The AP1000 plant value for R/D is denoted by the red vertical line. It is shown in Figure 3-73 that the wall static pressure distribution is independent of Re number and at 2 r/D (equivalent to H/D in Table 3-9), the pressure coefficient has converged to that of the free jet regardless of the confinement ratio.

One anomaly observed in the results of the Reference 3-18 and Reference 3-19 testing is that the pressure coefficient for free jets with confinement ratio < 1 do exhibit a higher pressure coefficient than that of the free jet. This is not explained in the literature but for the AP1000 reactor vessel cavity confinement ratio (X'/D) of [888]a,c, this is a moot point, as all the literature agrees that the pressure coefficient is analogous to that of the free jet both at the stagnation point and in the radial direction. In addition to this, the choked flow at the break outlet for the AP1000 plant is characterized by a Mach number of 1, indicating sonic velocities. This means that the disturbance cannot propagate upstream and affect the jet behavior because the disturbance speed is the difference in sonic velocity and that of the fluid; thus, for a Ma 1 flow (sonic velocity), the impingement surface should not impact the jet behavior because mathematically the disturbance speed is zero (Reference 3-15).

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Westinghouse Non-Proprietary Class 3 3-67 a,c Figure 3-73. Results of Reference 3-19 Testing for Confined and Unconfined (Normal) Jet Wall Pressure Distribution WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 3-68 3.3.5.3 Conclusions Based on the literature reviews of confined jet behavior for the AP1000 plant values of confinement ratio

[888]a,c and H/D (r/D as compared to testing) value of [999]a,c, the pressure coefficient is the same as that of the free jet; thus, the AP1000 plant free jet testing at the NTS facility accounts for the effects of confinement for the AP1000 plant-specific geometry.

The literature review confirmed that confined jet behavior is of interest when evaluating jet heat transfer, which is important in the process industry for cooling. Thus, for small confinement ratios (X'/D < 1), a low-pressure recirculating wake region can form resulting in boundary layer detachment and thus a degradation in heat transfer. However, for confinement ratios > 1, this effect dissipates and the boundary layer detachment has been shown via multiple experimental apparatuses to not occur.

3.3.5.4 Shock Waves The previous literature review provides good insight into the pressure dependence on confinement ratio resulting from confined jet impingement for single phase jets. Single phase jet behavior has been substantiated as adequate for addressing GSI-191 considerations based on air jet testing provided in References 3-4 and 3-5. However, the literature review did not address the shock phenomena resulting from the impingement of underexpanded jets.

To assess the impact on the shock formed by two-phase critical jet flows as a result of confinement, the first thing one must understand is the nature of the flow resulting from the break condition. From Reference 3-16, the shock wave for critical jet impingement of two-phase flows is that of a standing shock wave. This is largely because while the choking condition is established as a Ma =1 flow due to the thermodynamics of phase change, the jet flow is accelerated to supersonic conditions downstream of the break exit prior to impinging on the target resulting from a rapid expansion caused by vaporization. For a two-phase jet in confinement, it is possible for these conditions to occur simultaneously. This can occur as a result of the physical characteristics of the core region. The core of a two-phase jet is one in which the boundary layer entrainment at the jet periphery (this is the phenomenon that results in jet spreading) has not penetrated, and the conditions of the core are the same as that of the break exit plane. Since for the AP1000 plant in the reactor vessel cavity the confinement ratio (X'/D=[888]a,c) is less than the potential core length (Lc/D2), the core impinging on a surface will contain a Mach disk and is represented by a Ma=1 flow. Figure 3-74 shows schematically the structure of the underexpanded jet, and demonstrates the co-existence of both Ma=1 and Ma1 flows simultaneously within the jet structure.

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Westinghouse Non-Proprietary Class 3 3-69 M=1 Does not occur for two-phase jets as the core is not supersonic.

Figure 3-74. Structure of Underexpanded Jet A few points of clarification on Figure 3-74 are needed. The first point is that for a two-phase critical flow the exit plane M = 1, and in the core region the velocity and thermodynamic state is preserved as in the break exit plane (Reference 3-16). This is well substantiated for two-phase jet discharge.

The second point is that the reflected shock lines do not occur for a two-phase jet. The reflected shock is based on the core being at M>>1 and the resulting core expansion shock reflecting off the jet boundary expansion shock and yielding resulting compressive shock waves.

The justification for this comes from an understanding of Prandtl theory and the Rankine-Hugoniot relationships (Reference 3-22).

The fundamental premise of the application of Prandtl theory and the Rankine-Hugoniot relationships is that across a shock wave, entropy must increase. This follows the second law of thermodynamics, and the nature of shock formation must be taken into consideration. These relationships prove mathematically that no shock is generated for M=1 flows, because entropy doesnt change from the upstream or downstream condition. Figure 3-75 (Reference 3-16) shows the relationship of upstream and downstream conditions associated with the formation of a shock wave.

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Westinghouse Non-Proprietary Class 3 3-70 Figure 3-75. Reference 3-16 Shock Wave Formation within a Pipe A plot of the Rayleigh and Fanno lines (h-s diagram) demonstrates that the maximum entropy of the flow corresponds to M = 1 (Reference 3-23). So if entropy is maximized at Ma = 1, it is not possible for a shock to occur as the second law cannot be satisfied.

Since there is no shock emitted in the core region, the shock reflection and the associated compression waves do not occur. This is further substantiated from the two-phase jet testing both at Marviken (Reference 3-22) and at the NTS facility. Both facilities indicate a stagnation pressure in the core equal to the reservoir conditions, indicating pressure is completely recoverable. A shock formation in the core would not allow this to be true as the pressure downstream of the shock would have to be less than that upstream of the shock, and thus the pressure would not be totally recoverable.

For the Ma >1 flow outside the jet boundary, Reference 3-16 substantiates the formation of a standing/stationary shockwave and justifies the conditions of the two-phase jet, as well as substantiates the formation of this type of shock wave.

The nature of the standing/stationary shock wave is that it doesnt move. It is a phenomenon that results from the rapid deceleration of the fluid due to impingement. Figure 3-76 from Reference 3-21 is a Schlieren photograph of the impingement of an underexpanded jet.

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Westinghouse Non-Proprietary Class 3 3-71 Figure 3-76. Shlieren Photograph of Impingement of an Underexpanded Jet So, if the jet core does not emit a shock wave as a result of second-law considerations, and the jet periphery shock wave is stationary, then the impact of confinement on shock behavior is that there is no impact on the shock behavior resulting from confinement.

This is further substantiated by considering the propagation time of the sonic jet:

=

where, tp = disturbance propagation time Lr = reference length through which the disturbance propagates = X' = [88888]a,c for RV cavity VP = disturbance propagation speed/acoustic wave speed = speed of sound in air = 1130 ft/s The disturbance propagation time tp is significant in that it communicates the time it takes the jet to communicate pressure changes in the system and is equal to a value of ~0.003s. Physically, what this is saying is that the pressurization of the confined space can only occur as fast as the acoustic wave speed.

Since both the core and the supersonic periphery are moving at least as fast as the acoustic wave speed, it does not seem reasonable that their behavior can be affected by the confinement, as that would mean the confinement pressurization would have to occur faster than the mechanism that is causing the pressurization, which is inconsistent.

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Westinghouse Non-Proprietary Class 3 3-72 Reference 3-32 provides additional discussion regarding the behavior of confined jets and asserts that in Region 1 of the jet region, or what is known as the core region, that the jet behavior is unaffected by the confinement. An excerpt from Reference 3-32 is included below (Figure 3-77):

Figure 3-77. Excerpt from Reference 3-32 Using Figure A.101 and A.125 from Reference 3-16 the core region for the AP1000 plant cold leg is approximately [ ]a,c. For the AP1000 plant confinement ratio of [

]a,c In conclusion, the jet pressure distribution as conveyed in multiple literature sources via the empirical data from multiple test facilities appears to approach that of the free jet for confinement ratios (X'/D)>1.

This conclusion is independent of jet Re number provided the flow is turbulent. The AP1000 plant-specific confinement ratio for a double-ended cold-leg guillotine (DECLG) break was calculated and used to relate the empirical data. This comparison showed for the AP1000 plant-specific confinement ratio that the confined plate and impingement plate pressure distribution were analogous with that of the free jet.

The treatment of the shock waves showed that the shock-generating conditions for two-phase critical jet-flows lead to the formation of stationary/standing shock waves created by the rapid deceleration of the fluid as a result of the impinging surface. These shocks are stationary and as such are not a concern for reflection. Additionally, the sonic condition in the jet core was considered and Prandtl theory, Rankine-Hugoniot relationships, and Fanno-Rayleigh lines were used to demonstrate an M = 1 flow does not generate a shock wave due to second-law considerations. As a result of these considerations, special treatment of shock propagation resulting from two-phase critical jets need not be addressed because there would be no difference in the shock behavior under these conditions.

3.3.5.5 Predicting the Impingement Pressure on the NMI Based on the conclusions of the preceding sections, we will proceed with a prediction of the impingement pressure that the NMI would see in the plant and contrast/compare that with the impingement pressure exhibited at the NTS facility.

The prediction for the plant condition will come from Reference 3-16, Figure A.103, reproduced in Figure 3-79, with the AP1000 plant value of R/D denoted by the red dot. This figure, along with the NMI X/D and H/D values contained in Table 3-9 can be used to determine target pressure resulting from a DECLG. Table 3-10 shows the cold leg conditions related in SI units for application to Figure 3-76.

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Westinghouse Non-Proprietary Class 3 3-73 Reference 3-16 was developed to predict the direct jet impingement on a target resulting from a jet emanating from a jet source. Using Equation C-6 in ANS 58.2 confirms that the jet region which impacts the NMI is Region I outside the jet core. This confirms the applicability of the jet impingement prediction using Reference 3-16. However, it is known that ANS 58.2 conservatively over predicts the asymptotic plane and jet area. [

]a,c What this means physically is that placing the NMI at the jet centerline stagnation point of the test facility resulted in a conservative test condition as compared to what the NMI would actually experience in the plant configuration.

Figure 3-78. Dimensionless velocity profile for Region IV jet as compared to slot jet without impingement plate WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 3-74 Table 3-10. SI units of Cold Leg Temperature, Pressure, and Subcooling for use in developing impingement pressures from Figure 3-76 (Reference 3-6) a,c a,c Figure 3-79. NUREG/CR-2913 Target Pressure Distributions Application of Figure 3-79, along with the values in Table 3-10, yields a target pressure value of 10 bars (145 psia). This will be taken to be a constant pressure distribution along the top surface of the neutron shield blocks containing NMI. This is conservative as it does not assume a radial degradation of the impingement pressure which Figure 3-79 shows will occur as R/D increases. [

]a,c so based on this information, the value determined in Figure 3-79 is accurate and appropriate.

The NTS jet facility rake data will be used to determine the facility-induced pressure distribution on the top surface of the neutron shield blocks containing NMI. From the rake data for [8888 8]a,b,c, Table 3-11 provides the rake data as a function of R/D.

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Westinghouse Non-Proprietary Class 3 3-75 Table 3-11. NTS Stagnation Pressures on Neutron Shield Blocks a,b,c

[

]a,b,c for the NTS facility as described in this program.

Integrating the plant pressure distribution and the fitted test pressure distribution will provide an indication of the pressure distribution and the associated magnitude ratios from that of the test to that of the plant. Assuming that the jet is located at the center of the square, this integral is given by:[

]a,b,c This indicates that the integrated pressure distribution in r from the test is approximately [

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Westinghouse Non-Proprietary Class 3 3-76 x ]a,b,c WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 3-77 3.4 NEUTRON SHIELD BLOCK SUBMERGENCE TEST

SUMMARY

AND OBJECTIVES Submergence testing (Reference 3-27) was conducted in order to better understand how the neutron blocks would affect chemical debris generation. The objective was to determine if debris was generated by exposed RVIS components ([ ]a,c) with both broken and intact encapsulation.

The data from this test were then incorporated into the chemical effects model as appropriate.

3.4.1 Neutron Shield Blocks Submergence Test Summary The AP1000 plant RVIS and CA31 module utilize [

]a,b,c All three of these locations are below the maximum floodup level and therefore have the potential to be fully submerged following a LOCA.

A submergence test program was performed that evaluated the neutron shield blocks constituent components for potential generation of debris (Reference 3-23). Additionally, results were taken from this test and integrated into the AP1000 plant chemical effects model to determine the integrated effects on chemical precipitate formation (Reference 3-26).

The objective of the neutron shield block submergence test was to observe how the specimens behaved in post-LOCA fluid, examine the specimens after the test, and obtain data from the test fluid to draw conclusions about the effects of submergence on the RVIS and CA31 in relation to debris generation.

3.4.2 Submergence Test Specimens The neutron shield blocks submergence test program tested seven specimens. They were chosen to study the different components of the RVIS and CA31. The seven specimens are as follows:

1. [ ]a,b,c

2. [ ]a,b,c

3. [ ]a,b,c

4. [ ]a,b,c

5. [ ]a,b,c

6. [ ]a,b,c

7. [ ]a,b,c

[

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Westinghouse Non-Proprietary Class 3 3-78

]a,b,c 3.4.2.1 Acceptance Criteria The acceptance criteria of this experiment is that the tests were carried out in accordance with the test procedure, including that temperature, pressure, and chemistry conditions were maintained within the specified bounds (Reference 3-27). Because the acceptance criteria were met, the results of the tests are valid and applicable to use in terms of GSI-191 debris generation determination and disposition.

3.4.2.2 Test Conditions The specimens were tested in [

]a,b,c WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 3-79 a,b,c Figure 3-80. Schematic of Autoclave Test Vessel 3.4.2.3 Sampling Procedure Samples were taken at [

]a,b,c.

Between the bounding maximum temperature of fluid that the samples were subjected to during testing, which would encourage dissolution and degradation of material, and the bounding minimum temperature before the sample fluid was filtered, which would encourage precipitation, the test creates a scenario that drives for the highest possible amount of chemical precipitate and other elemental release from the samples.

3.4.2.4 Submergence Test Results

[

]a,b,c WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 3-80

[

]a,b,c Table 3-12. Integrated Elemental Release a,c

[

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Westinghouse Non-Proprietary Class 3 3-81 X ]a,b,c 3.4.2.5 Impact on Chemical Effects From Sample 1, the submergence testing showed significant [

]a,b,c As the RVIS and CA31 components are all [ ]a,c, there is not expected to be communication between the post-LOCA fluid and the neutron shield block internal components;

[

]a,b,c as discussed in Section 5.1.3.

By utilizing a [ ]a,b,c design, the NMI discussed in this report meets the licensing basis requirement that a suitable equivalent to MRI does not generate chemical effects.

3.5 CHARACTERIZATION OF CA31 NEUTRON SHIELDING MATERIAL

[ ]a,b,c has excellent thermal neutron shielding capabilities and fair fast neutron shielding. It is stable at high temperature in dry air [

]a,b,c 3.5.1 Impact of CA31 Neutron Shielding Material on Chemical Effects As the RVIS and CA31 components are all [ ]a,c, there is not expected to be communication between the post-LOCA fluid and the neutron shield blocks containing [

x ] a,c is non-reactive in the expected environment and therefore not a chemical precursor, it has no impact on the current chemical effects analysis.

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Westinghouse Non-Proprietary Class 3 3-82 3.5.2 Impact on Containment Recirculation Screens As described in UFSAR/DCD Section 6.3.2.2.7 (Reference 3-1), the passive core cooling system (PXS) has two different sets of screens used following a LOCA - IRWST screens and containment recirculation screens. The screens prevent large debris from entering the reactor and blocking core cooling passages during a LOCA and are designed to pass the maximum injection flow with one half of their area blocked.

The screen design also provides a trash rack function to prevent a single object from blocking a large portion of the screen. The two screens provided for containment recirculation have a two-foot high debris curb to prevent debris that is swept along the floor from entering or blocking the screens.

Reference 3-24 evaluated [

]a,b,c

3.6 REFERENCES

3-1. APP-GW-GL-700, Rev. 19, AP1000 Design Control Document, June 2011.

3-2. NEDO-32686-A, Volumes 1 through and including 4, Utility Resolution Guidance for ECCS Suction Strainer Blockage, Boiling Water Reactor Owners Group, October 1998.

3-3. N-REP-34320-10000, Rev. 0, Jet Impact Tests - Preliminary Results and Their Application, Ontario Power Generation, April 2001.

3-4. NEI 04-07, Pressurized Water Reactor Sump Performance Evaluation Methodology; Volume 1, Rev. 0, December 6, 2004.

3-5. NEI 04-07, Pressurized Water Reactor Sump Performance Evaluation Methodology, , Volume 2

- Safety Evaluation by the Office of Nuclear Reactor Regulation Related to NRC Generic Letter 2004-02, Revision 0, December 6, 2004.

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Westinghouse Non-Proprietary Class 3 3-83 3-6. Report No.:FAI/110497, Rev. 1, PWROG Model for the Two Dimensional Free Expansion of a Flashing, Two Phase, Critical Flow Jet, February 2012.

3-7. WCAP-17617, Volume I, Jet Impingement Testing of AP1000 In-containment Cables, November 2014.

3-8. LTR-LIS-14-339, AP1000 Plant LBLOCA DECLG Data for GSI-191 Jet Impingement Testing Support, Westinghouse Electric Company LLC, July 18, 2014.

3-9. WCAP-17616, Jet Impingement Testing of AP1000 Reactor Vessel Insulation System Neutron Shielding Blocks, November 2014.

3-10. APP-GW-T1-660, Rev. 0, AP1000 Cable Aging for Jet Impingement Testing, Westinghouse Electric Company LLC, January 2014.

3-11. WCAP-17617, Volume II, Jet Impingement Testing of AP1000 In-containment Cables, November 2014.

3-12. APP-MN20-V1-101, Rev. 1, RV Insulation Key Layout Elevation March 2013.

3-13. APP-MN20-V1-102, Rev. 1, RV Insulation Key Plan April 2011.

3-14. APP-MN20-V2-147, Rev. 1, RV Upper Neutron Shielding Support Steel MKUNS Details, April 2011.

3-15. NASA Technical Note D-5652, Survey of Literature on Flow Characteristics of a Single Turbulent Jet Impinging on a Flat Plate, J.W. Gauntner, J.N.B. Livingwood, P. Hrycak Lewis Research Center, Cleveland Ohio February 1970.

3-16. NUREG/CR-2913 Two-Phase Jet Loads Sandia Labs. January 1983.

3-17. Baydar, E., Ozmen Y., An experimental and numerical investigation on a confined impinging air jet at high Reynolds numbers Applied Thermal Engineering 25 (2005) pp. 409-421.

3-18. Gao, N.; Ewing, D. Investigation of the effect of confinement on the heat transfer to round impinging jets exiting a long pipe International Journal of Heat and Fluid Flow 27 (2006) 33-41 3-19. Deogonda, P., Vijaykumar N.C., Murugesh, M.C., Dixit, V., Wall static Pressure Distribution Due to Confined Impinging Circular Jet 3-20. Crist, S., Sherman, P.M., Glass D.R., Study of the Highly Underexpanded Sonic Jet AIAA Journal Vol. 4, No. 1 pp 68-71.

3-21. Lamont, P.J., Hunt, B.L., The impingement of underexpanded, axisymmetric jets on perpendicular and inclined flat plates J. Fluid Mech. (1980), vol. 100, part 3, pp. 471-511.

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Westinghouse Non-Proprietary Class 3 3-84 3-22. Compressible Fluid Flow, Saad, Michel 1985 Prentice-Hall, Inc. Englewood Cliffs, New Jersey 3-23. EPRI-NP-4362, Two-Phase Jet Modeling and Data Comparison December 1985 3-24. APP-GW-GER-201, Test Report: [ ]a,c Settling Analysis, April 2017.

3-25. TR-CCOE-14-02, Rev. 0, AP1000 Insulation Submergence Testing, November 2014.

3-26. APP-PXS-M3C-221, Aluminum Inventory for AP1000 Containment.

3-27. APP-GW-T5-006, Submergence Test Plan for Encapsulated [ ]a,c Insulation and Neutron Shielding in the Reactor Cavity {Safety Related}, December 2013.

3-28. APP-PXS-M3C-052, AP1000 GSI-191 Chemistry Effects Evaluation.

3-29. APP-NS27-Z0-102, Procurement Specification for Non-Safety AP1000 [Xxx Xxxxxxx]a,c, Material used As Neutron Shielding.

3-30. APP-GW-T1R-001, AP1000 Cable Deconstruction Test Report (U.S. Licensing Basis),

July 2014.

3-31. APP-PXS-M3C-080, AP1000 Non-Coating Debris Contributions Towards GSI-191 Debris Limits.

3-32. ML050830344, The ANSI/ANS Standard 58.2-1988: TWO-PHASE JET MODEL Graham Wallis September 15, 2004.

3-33. NUREG/CR-6808, Knowledge Base for the Effect of Debris on Pressurized Water Reactor Emergency Core Cooling Sump Performance, February 2003.

3-34. WCAP-16914-P, Revision 6, Evaluation of Debris Loading Head Loss Tests for AP1000 Recirculation Screens and In Containment Refueling Water Storage Tank Screens, April 2015.

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Westinghouse Non-Proprietary Class 3 4-1 4 DEBRIS GENERATION BREAK SIZE DETERMINATION The documents providing guidance for resolution of GSI-191 (NEI-04-07 Volume I [Reference 4-1]) and Volume II (Reference 4-2) provide an alternative method to the baseline calculations that may be used for demonstrating acceptable containment sump performance relative to the requirements identified in the baseline methods.

This alternative methodology allows for both an alternate design basis break size in conjunction with the baseline methodology for the RCS and attached piping, and the use of realistic analysis assumptions, credit for non-safety systems and structures, and operator actions when evaluating up to a full double-ended guillotine break (DEGB) of the RCS main loop piping.

The alternative methodology consists of three main components: the alternate debris generation break size, the Region I analysis, and the Region II analysis. Each of the components and their application to the AP1000 plant will be discussed in the following sections.

4.1 DEBRIS GENERATION BREAK SIZE The debris generation break sizes for the evaluation of sump performance are defined as:

• For all American Society of Mechanical Engineers (ASME) Code Class 1 auxiliary piping (attached to the AP1000 plant RCS main loop piping) up to and including a DEGB of any of these lines, the GSI-191 design-basis rules apply.

• For breaks in the AP1000 plant RCS main loop piping (hot leg and cold leg piping) up to a size of equivalent break diameter to that of a 14-inch Schedule 160 pipe (approximately 11.188 inches),

the GSI-191 design-basis rules apply (applicable to Region I) .

• For breaks in the AP1000 plant RCS main loop piping (hot leg and cold leg piping) with equivalent diameter greater than that of a 14-inch Schedule 160 pipe (approximately 11.188 inches) and up to the DEGB, mitigative capability must be demonstrated, but GSI-191 design-basis rules may not necessarily apply (applicable to Region II).

Per the SE on NEI-04-07 (Reference 4-2), the basis for acceptance of the alternate break sizes stems from the information developed by the NRCs Office of Nuclear Regulatory Research (RES) regarding the frequency of RCS ruptures of various sizes. The RES study determined the frequency of primary pressure boundary failures under normal operational loading and transients and concluded that the probability of a PWR primary-piping system rupture is generally very low, and that the break frequency decreases with increasing piping diameter.

Reference 4-1 reports that the NRC staff considered the fact that there is a substantial difference from a deterministic, margins to failure or flaw tolerance perspective between large diameter main coolant loop piping and the next largest ASME Code Class 1 attached auxiliary piping: certain ASME Code Class 1 auxiliary piping systems may be more susceptible to failure as a result of environmental conditions that are conducive to known degradation mechanisms and/or loading conditions that routinely apply significant stresses to the piping system. The Staff provided an example of both of these WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 4-2 considerations, citing a typical PWR pressurizer surge line in which Alloy 82/182 dissimilar metal welds are subjected to a high-temperature operating environment known to abet primary water stress-corrosion cracking (PWSCC) and are subjected to significant bending loads during startup/shutdown conditions because of the large temperature gradient between the pressurizer and the hot leg of the main coolant loop.

Per the AP1000 plant DCD (Reference 4-3), the AP1000 plant RCS piping is fabricated of forged seamless Series 300 austenitic stainless steel without longitudinal or electroslag welds or cast fittings. The piping complies with the requirements of the ASME Code,Section II (Parts A and C),Section III, and Section IX, and adheres to the requirements of Regulatory Guide 1.44 for the use of Series 300 stainless steel materials. These fabrication features, including no cast fittings or Alloy 600 weld metals for attached RCS loop piping, minimize the risk of PWSCC failure throughout the RCS piping system. Although these features contribute to the overall safety of the plant and reduce the risk of high energy line breaks, the alternative method for determining the debris generation break size does not credit these features.

The NEI 04-07 discussion in the SE (Reference 4-2) concluded that the division of the pipe break spectrum noted above for evaluating debris generation was acceptable and that licensees could use the defined debris generation break size for distinguishing between Region I and Region II analyses when using the alternative methodology. These conclusions are applicable to the AP1000 plant RCS loop piping and the associated RCS main loop branch piping.

4.2 REGION I ANALYSIS The Region I analysis is applicable to the AP1000 plant and includes evaluation of the RCS main loop piping and every branch line attached to the RCS main loop piping. For the Region I analysis, all lines are evaluated for debris generation and transport using the baseline methods as defined in NEI 04-07 (Reference 4-1) and modified by the SE on NEI 04-07 (Reference 4-2).

In the Region I analysis, a DEGB with full separation and an inner break diameter equivalent to that of a 14-inch Schedule 160 pipe is assumed for determining debris generation from the AP1000 plant RCS main loop piping due to the low probability of a DEGB of the main loop piping. The RCS main loop piping is identified as the hot leg and cold leg piping only. For RCS main loop branch piping, a DEGB with full separation and an inner break diameter equivalent to that of the branch line inner diameter (ID) must be assumed. This remains applicable to RCS main loop branch lines with IDs larger than that of a 14-inch Schedule 160 pipe, including but not limited to the following:

• [ ]a,c

• [ ]a,c

• [ ]a,c For Region I analyses, the spherical model defined in NEI 04-07 (Reference 4-2) shall be used for the determination of the ZOI from the RCS main loop piping and the RCS main loop branch piping. In accordance with this model, the ZOI is defined as a sphere with a radius equivalent to the ID of the WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 4-3 assumed DEGB size multiplied by a scaling factor, N, determined for the specific material to be evaluated.

=

In the Region I analysis, the radius of the spherical ZOI for the RCS main loop piping and a material with a 4D ZOI is determined as follows based on the ID of a 14-inch Schedule 160 pipe.

= 4 x 11.188 = 44.752 The radius of the spherical ZOI for an RCS main loop branch line is determined in a similar manner, except that the ID of the line in question is used, even when larger than the 14-inch Schedule 160 ID. For example, the ZOI radius for the [ ]a,c, as applied to a material with a 4D ZOI, is calculated below.

= [ ]a,c The spherical ZOI is modeled with the defined radius applied to the axial centerline of the pipe.

Following this methodology, the ZOI and subsequent debris source term calculations can be determined for all Region I break evaluations.

While the SE in NEI 04-07 (Reference 4-2) allows for the use of a volume-equivalent ZOI radius for the calculation of the ZOI volume, this method is not employed for the AP1000 plant analyses.

The volume-equivalent ZOI radius is determined by using the methodology presented in American National Standards Institute / American Nuclear Society (ANSI/ANS) standard 58.2-1988 (Reference 4-4) to model the effects of the jet originating from a postulated pipe break. In typical applications of this methodology, the pipe break dimensions and applicable plant conditions during the break are utilized to model the DEGB jet using ANSI/ANS 58.2-1988 (Reference 4-4). From this model, jet isobar volumes may be calculated. For a given material, the isobar volume may be determined corresponding to the material destruction pressure or a bounding pressure value at which testing has shown that the material will not be destroyed. This jet isobar volume is doubled to account for both ends of the DEGB and is then equated to the volume of a sphere; the radius of this sphere is the volume-equivalent ZOI radius.

The ZOI radii proposed in Section 3 are based on jet impingement testing with the target material placed on the jet centerline. The ZOI radius for the material, as determined from testing, is equal to the centerline distance from the break location to the target at which the material did not fail. This ZOI radius can then be scaled as a function of pipe break diameter. While the volume-equivalent spherical ZOI model has been determined to be conservative, additional conservatism exists in the use of a ZOI radius determination based on jet impingement testing of a centerline target.

To qualitatively observe this conservatism, a comparison of the two methods was performed. For centerline distances of 3D and 4D at which jet impingement tests were performed, the volume of the theoretical isobar at that centerline distance was determined based on the ANSI/ANS 58.2-1988 methodology. The volume and radius of a sphere with twice the volume of this isobar was used to approximate the dimensions of the volume-equivalent ZOI. For distances of 3D and 4D, these values WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 4-4 were compared to the volume and radius of a sphere where the radius was equal to the centerline target distance.

Table 4-1 provides a comparison of the ZOI method proposed herein based on jet impingement testing with the ANSI/ANS 58.2-1988 volume-equivalent ZOI. To support this comparison, the spreadsheet documented in Reference 4-8 for the calculation of stagnation pressure isobars using the ANSI/ANS 58.2-1988 methodology was used to model an 11.188-inch ID cold leg DEGB (14-inch Schedule 160) and a 22-inch ID cold leg DEGB (RCS cold leg ID). The approximate jet centerline pressure was determined for target distances of 3D and 4D (column 2), with D representing the actual break diameter. The corresponding isobar volume determined for the centerline pressure at each target distance was doubled to account for both ends of the DEGB and reported for each break size as the volume-equivalent spherical ZOI volume (columns 5 and 9). From this value, the volume-equivalent spherical ZOI radius can be determined (columns 4 and 8). For comparison, the proposed ZOI radius at each target distance (columns 3 and 7), as well as the proposed ZOI volume based on jet impingement testing, are provided (columns 6 and 10). The ANSI/ANS 58.2 jet model from Reference 4-8 used to generate this data was benchmarked against the results provided in Table I-3 of Reference 4-2. A comparison of the benchmark results to the Table 1-3 results is provided in Table 4-2.

Table 4-1. Comparison of Proposed ZOI to Volume-Equivalent ZOI a,c X x xx x

x x

x x

x X

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Westinghouse Non-Proprietary Class 3 4-5 Table 4-2. Comparison of Jet Model Benchmark to Table I-3 of Reference 4-2 a,b,c The results provided in Table 4-1 illustrate that the volume-equivalent ZOI radii based on target distance, as calculated using the ANSI/ANS 58.2-1988 jet model, are bounded by the material destruction distances determined from jet impingement testing along the jet centerline at distances of 3D and 4D. This is due to the fact that the jet pressure in the ANSI/ANS 58.2-1988 jet model is greatest along the jet centerline, as is also true in reality. Therefore, the theoretical pressure at a given centerline target distance results in a volume-equivalent ZOI radius that is less than that centerline target distance. Since the proposed ZOI is based on centerline target distance rather than the isobar volume-equivalent model, the approach taken for calculating the ZOI dimensions for the AP1000 plant is conservative.

4.3 REGION II ANALYSIS The Region II analysis includes evaluations of break sizes in the RCS main loop piping (hot and cold) greater than the debris generation break size defined in the Region I analysis (approximately 11.188 inches in equivalent diameter), and up to a DEGB of the largest pipe in the RCS. The Region II analysis considers only RCS main loop piping because all primary-side attached auxiliary piping is fully addressed as part of the Region I analysis.

The Region II analysis is performed in the same manner and with the same methods used in the baseline analyses with respect to ZOI models and assumptions; however, the Region II analysis allows for more realistic analytical methods and assumptions: credit can be taken for limited pipe displacement; operation of non-safety systems, structures, and components; and operator actions where applicable. Crediting a limited pipe displacement requires the application of results from pipe break analyses and/or pipe stiffness analyses that may be utilized to limit the maximum break size to be evaluated. If no piping analyses have been performed for the RCS main loop piping application, then a DEGB assuming the full hot leg or cold leg pipe ID must be evaluated. If structural analyses have been performed for the RCS main loop piping WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 4-6 showing that limited pipe displacement, also referred to as a limited separation break, will occur, then an equivalent break diameter as determined for the limited separation break may be used to determine the ZOI for the Region II analysis.

No credit is taken for better estimate assumptions or operator actions in the Region II assessment. For the purpose of this topical report it is assumed that Region II analysis employs Region I assumptions. Credit for best estimate assumptions or operator actions are not used and are not part of the scope of the review of this topical report.

4.3.1 Full Separation Break If a pipe displacement or structural analysis is not performed, or if the analysis results in a break with lateral pipe displacement greater than 1 diameter and axial displacement greater than 0.5 diameter, then a full separation DEGB of the RCS main loop piping must be assumed. The radius of the spherical ZOI used in the Region II analysis is defined using the ID of the RCS main loop piping multiplied by a scaling factor, N, determined for the specific material to be evaluated. Therefore, the ZOI radius for a material with a 4D ZOI is defined below for the RCS hot leg and cold leg piping.

Hot Leg:

= [ ]a,c Cold Leg:

= [ ]a,c The spherical ZOI is modeled with the defined radius applied to the axial centerline of the pipe.

4.3.2 Limited Separation Breaks As stated in NEI 04-07 (Reference 4-1) and modified by the SE on NEI 04-07 (Reference 4-2), Region II analyses are limited to a DEGB of the RCS main loop piping unless physically limited by piping restraints and supports, other plant structural members, or piping stiffness as may be demonstrated by analysis.

If a structural or pipe displacement analysis is performed that results in a lateral pipe displacement, d, of less than 1 diameter and an axial displacement, Wf, of less than or equal to 0.5 diameter, then a limited separation break may be utilized in the Region II analysis in determining the alternate break size. The geometry of a limited separation break is discussed in ANSI/ANS 58.2-1988 (Reference 4-4) and is shown in Figures 4-1 and 4-2. ANSI/ANS 58.2-1988 defines a limited separation break as having an axial displacement, Wf, of less than or equal to 0.5 diameter and a lateral displacement, d, of less than or equal to the pipe wall thickness, t; however, this definition is extended for the purposes of AP1000 plant Region II analyses to include lateral displacements greater than the pipe wall thickness and up to 1 diameter.

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Westinghouse Non-Proprietary Class 3 4-7 Figure 4-1. Circumferential Break with Limited Separation Figure 4-2. Geometry of Circumferential Break with Limited Separation One of the key parameters of the circumferential break with limited separation is the value of the axial displacement, Wf. Similar to the ID of the pipe, De, for a full separation break, the axial displacement acts as the characteristic length for the simplified expanding jet model proposed in Appendix C of ANSI/ANS 58.2-1988 (Reference 4-4). The length of the jet core region for a full separation break is provided below as a function of the pipe ID, De.

= 0.26 + 0.5 WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 4-8 Similarly, the length of the jet core region for a limited separation break is defined using the same expression where the axial displacement, Wf, acts as the characteristic length.

= 0.26 + 0.5 From the two-phase jet loading model documented in NUREG/CR-2913 (Reference 4-5), the centerline jet pressure for a subcooled two-phase flow jet is proportional to the ratio of the characteristic length of the break, D, to the distance from the break plane to the target, L, as shown in the expression below.

4.3.2.1 Limited Separation Break; 0 d t A limited separation break with lateral displacement, d, less than or equal to the pipe wall thickness, t, shall be defined by the characteristic length, Wf, where:

[ ]a,c In this case, the radius of the spherical ZOI used in the Region II analysis is defined [

]a,c.

= [ ]a,c Therefore, the ZOI radius for a material with a [

]a,c.

= [ ]a,c The spherical ZOI is modeled with the [ ]a,c.

4.3.2.2 Limited Separation Break; t < d < De A limited separation break with lateral displacement, d, greater than the pipe wall thickness, t, and less than 1 diameter shall be defined, [

]a,c. The expression for the equivalent diameter, Dequivalent, of such a limited separation break is provided in the equation below.

[ ]a,c WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 4-9 The characteristic length component contributed by the exposed cross-sectional break area, Dlateral is determined by [

]a,c using the expression below.

[ ]a,c In this case, [

]a,c.

= [ ]a,c The spherical ZOI is modeled with [

]a,c 4.4 RCS MAIN LOOP PIPING DISPLACEMENT ANALYSIS 4.4.1 Analysis Overview In support of the Region II analyses, a structural evaluation of the RCS main loop piping was performed using ANSYS' LS-DYNA' to evaluate pipe whip and displacement in the RCS main loop piping. This analysis, documented in APP-PL01-P0C-003 (Reference 4-7), used finite element models run on LS-DYNA to demonstrate that the hot legs and cold legs do not fully separate as a result of a LOCA. The LOCA pipe reaction forces assumed for each DEGB scenario in Reference 4-7 were calculated in APP-PL01-P0C-002 (Reference 4-6). The Reference 4-7 structural analysis postulated DEGBs in multiple locations on both the RCS hot legs and cold legs. The following five break scenarios were analyzed:

1. [ ]a,c

2. [ ] a,c

3. [ ] a,c

4. [ ] a,c

5. [ ]a,c The locations of the five postulated break scenarios are shown in Figure 4-3.

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Westinghouse Non-Proprietary Class 3 4-10 a,c Figure 4-3. Primary Coolant Loop Break Locations (Reference 4-7)

The five postulated break scenarios and locations were strategically chosen to bound breaks at all locations along the hot leg and the cold leg piping. Due to the curved geometry of the hot leg and the cold legs, the thrust force generated from a break can act on the piping to create a moment about the pipe anchor points located at the primary equipment connections. The subsequent displacement of the piping is a function of the length of the moment arm or moment arms generated about one or more anchor points, the structural integrity of the piping, and the presence of any structures that may act to limit the displacement of the piping. In the case of the AP1000 plant [

]a,c WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 4-11

[

]a,c a,c Figure 4-4. RCS Cold Leg Steam Generator Compartment Break Locations

[

]a,c WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 4-12 a,c Figure 4-5. RCS Cold Leg Nozzle Gallery Break Locations

[

]a,c WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 4-13 a,c Figure 4-6. RCS Cold Leg CA01 Module Penetration Break Location

[

]a,c WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 4-14 a,c Figure 4-7. RCS Hot Leg Break Locations 4.4.2 Pipe Displacement Results The results of the hot leg and cold leg pipe displacement analyses performed in APP-PL01-P0C-003 (Reference 4-7) are provided in Table 4-3 for the five postulated break locations discussed in subsection 4.4.1.

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Westinghouse Non-Proprietary Class 3 4-15 Table 4-3. Results of Pipe Displacement Analysis - APP-PL01-PL0C-003 (Reference 4-7) a,c 4.4.3 Application of Results to Region II Analyses The Region II analyses, as discussed in Section 4.3 and in NEI 04-07 (Reference 4-1), need only be performed for RCS main loop piping under the alternative methodology and may credit the results of a piping displacement or structural analysis in the determination of the break size. Therefore, the hot leg and cold leg displacement results from the ANSYS LS-DYNA analysis presented in Table 4-3 may be used to support the AP1000 plant Region II debris source term calculations Following the methodology described in Section 4.3, the equivalent break diameter was determined for each of the five DEGB scenarios, and was compared to the 14-inch Schedule 160 break size already used as the basis for the Region I analyses. To ensure the validity and conservatism of the use of a spherical ZOI model based on a break in a 14-inch Schedule 160 break diameter, the representative geometry of each limited separation break was also determined using approximate operating and boundary conditions anticipated during the AP1000 plant LOCA transient and compared to the geometry of a fully separated DEGB. For this comparison, the SE on NEI 04-07 (Reference 4-2) was used in conjunction with the expanding jet models provided in Appendix C of ANSI/ANS 58.2-1988 (Reference 4-4).

As a qualitative method of comparing the circumferential break jet expansion geometries with limited separation to the full separation DEGB jet expansion geometry, a simplified volumetric comparison was proposed. Using the methodology outlined in Appendix C of Reference 4-4, the volume of the expanding jet for each of the five DEGB scenarios with limited separation would be determined up to the asymptotic plane. This jet volume accounts for the entirety of jet Regions 1 and 2, rather than specific isobars. The Region 1 and 2 volumes for each limited separation DEGB scenario would be compared to the volume of Regions 1 and 2 of the expanding jet, up to the asymptotic plane, for both ends of a DEGB with full separation for a 14-inch Schedule 160 pipe. The Region 1 and 2 volumes were selected for comparison because the pressure at the asymptotic plane, which serves as the end plane of jet Region 2 in both models, is assumed to reach absolute ambient pressure; therefore, the pressure is the same at the end of each jet model. This method was chosen as a simplified alternative to the calculation and comparison of jet isobars using the circumferential limited separation jet model and the full separation jet model. The described comparison is intended to provide reasonable assurance that, for instances where Dequivalent is WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 4-16 less than the 14-inch Schedule 160 break diameter, the axial component of the limited separation break is bounded by the spherical ZOI for a 14-inch Schedule 160 break. The jet geometries described above are shown in Figure 4-8.

Figure 4-8. Limited Separation (A) and Full Separation (B) Break Geometries While the industry has guidance for limited separation breaks with no axial displacement and limited separation breaks with very limited lateral displacement, no guidance exists for break scenarios featuring both break types, up to a full DEGB. The method proposed in Section 4.3.2 to fill this gap by scaling the spherical ZOI based on a sum of the characteristic diameters of both break types is conservative for the purpose of generating a sphere that will envelop the centerline destruction pressure for a given material based on the discussion of the spherical model in Section 4.2. Note that the application of the circumferential break model beyond its recommended bounds of application, while understood to not be a true representation of the postulated jet geometry, tends to predict a larger volume and longer centerline jet reach than the DEGB model when considering the proposed characteristic length methodology. For example, a circumferential break in a 22 inch pipe with Wf = 11 inches has half the characteristic length of a DEGB in the same pipe; however, the reach of the asymptotic plane is greater than half the distance (7.3D as compared to 10.3D from the example given above). Similarly, up to the asymptotic plane the volume of this jet is approximately two thirds of that of the DEGB jet volume up to the asymptotic plane.

Thus, the circumferential break geometry is favored in the comparison of the spherical ZOI where possible.

The initial conditions for the determination of the jet expansion geometries were assumed to be similar to those of [

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Westinghouse Non-Proprietary Class 3 4-17 X1 x

x x ]a,c 4.4.3.1 Hot Leg Break Evaluations 4.4.3.1.1 Hot Leg DEGB at the Reactor Vessel Nozzle From results presented in Table 4-3, the hot leg break at the reactor vessel nozzle resulted in an axial displacement, Wf, [

]a,c Given that the ZOI for a fully separated break of a [

]a,c is bounded by the Region I analysis.

4.4.3.1.2 Hot Leg DEGB at the Steam Generator From results presented in Table 4-3, the hot leg break at the steam generator resulted in an axial displacement, [

1 The existence of the asymptotic plane has been debated throughout the industry. The intent of the evaluation contained herein and in the following sections does not endorse the existence of the asymptotic plane; it only uses the jet region volume calculations to demonstrate the conservative treatment of a spherical volume ZOI as opposed to that of a cylindrical geometry exhibited by a limited separation break.

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Westinghouse Non-Proprietary Class 3 4-18 X x a,c x ]

Given that the ZOI for a fully separated break of a [

]a,b,c is bounded by the Region I analysis.

4.4.3.1.3 Hot Leg Break Region II Results The SE on NEI 04-07 (Reference 4-2) provides an accepted three step approach to performing alternate evaluations incorporating realistic and risk-informed elements to the PWR sump analysis.

1. Define a debris generation LOCA break size to distinguish between customary and more realistic design-basis PWR sump analyses.

2. Perform customary design-basis analyses for break sizes up through the debris generation break size identified above (i.e., Region I analyses).

3. Perform analyses demonstrating long-term cooling and mitigative capability for break sizes larger than the debris generation break size up through the double-ended rupture of the largest RCS piping (i.e., Region II analyses).

For the Region I hot leg break analysis, a debris generation break size was defined as that of a 14-inch Schedule 160 pipe to be used for the performance of customary design-basis analyses. The performance of analyses demonstrating long-term cooling and mitigative capability for break sizes larger than the debris generation break size up through the double-ended rupture of the larger RCS piping for the RCS hot leg is bounded by the Region I analysis, even for the largest break shown to be mechanically possible.

This approach is to satisfy the third point of the proposed methodology because, as stated in NEI 04-07 (Reference 4-1) and modified by the SE on NEI 04-07 (Reference 4-2), Region II analyses are limited to a DEGB of the RCS main loop piping unless physically limited by piping restraints and supports, other plant structural members, or piping stiffness as may be demonstrated by analysis. The Region II analysis is performed to demonstrate long-term cooling and mitigative capability using best estimate assumptions, of which piping displacement and/or structural analyses are permitted by the industry guidance. As shown by the results of such analyses presented in subsections 4.4.3.1.1 and 4.4.3.1.2, the worst possible mechanical break displacement results in a ZOI that is smaller than that assumed in the Region 1 analysis.

Therefore, when this ZOI is considered at all points along the RCS hot leg to determine the location that will produce the maximum head loss across the sump screen as directed by the industry guidance, the resultant debris generation, and therefore resultant head loss, with no other best estimate assumptions considered, cannot be greater than the debris generation and resultant head loss from the Region 1 analysis.

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Westinghouse Non-Proprietary Class 3 4-19 4.4.3.2 Cold Leg Break Evaluations 4.4.3.2.1 Cold Leg DEGB at the Reactor Vessel Nozzle From results presented in Table 4-3, the cold leg break at the reactor vessel nozzle resulted in an axial displacement, [

Xx ]a,c 4.4.3.2.2 Cold Leg DEGB at the Reactor Coolant Pump From results presented in Table 4-3, the cold leg break at the RCP resulted in an axial displacement, [

]a,c would also conservatively bound this limited separation break.

The ZOI for a fully separated break of a 14-inch Schedule 160 pipe would bound the parameters of the limited separation jet; however, the calculated equivalent break diameter, Dequivalent, was found to be greater than the ID of the 14-inch Schedule 160 pipe (11.188 inches). As a result, a Region II analysis must be assessed for the cold legs from the RCP to the CA01 module penetration and must use a

[ ]a,c.

4.4.3.2.3 Cold Leg DEGB at the CA01 Module Penetration From results presented in Table 4-3, the cold leg break at the module penetration resulted in an axial displacement, [

WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 4-20

]a,c Based on the break results for the cold leg at the reactor vessel nozzle and the RCP presented in subsections 4.4.3.2.1 and 4.4.3.2.2, respectively, Region II analyses should be performed for the cold leg with a spherical ZOI based on a diameter of [ ]a,c from the reactor vessel nozzle to the CA01 module penetration, and a spherical ZOI based on an equivalent diameter of [ ]a,c from the CA01 module penetration to the RCP.

4.4.3.2.4 Cold Leg Break Region II Results The SE on NEI 04-07 (Reference 4-2) provides an accepted three step approach to performing alternate evaluations incorporating realistic and risk-informed elements to the PWR sump analysis.

1. Define a debris generation LOCA break size to distinguish between customary and more realistic design-basis PWR sump analyses.

2. Perform customary design-basis analyses for break sizes up through the debris generation break size identified above (i.e., Region I analyses).

3. Perform analyses demonstrating long-term cooling and mitigative capability for break sizes larger than the debris generation break size up through the double-ended rupture of the largest RCS piping (i.e., Region II analyses).

For the Region I cold leg break analysis, a debris generation break size was defined as that of a 14-inch Schedule 160 pipe to be used for the performance of customary design-basis analyses. In addressing the third step of the approved methodology, the AP1000 plant RCS cold leg is divided into two primary segments: the cold leg segment located inside the steam generator compartment, extending from the RCP through the CA01 module penetration, and the cold leg segment located inside the nozzle gallery, extending from the CA01 module penetration to the reactor vessel nozzle. The performance of analyses demonstrating long-term cooling and mitigative capability for break sizes larger than the debris generation break size up through the double-ended rupture of the larger RCS piping for the RCS cold leg is bounded by the Region I analysis for the largest break shown to be mechanically possible in the steam generator compartment and through the CA01 module penetration. For the largest break shown to be mechanically possible in the nozzle gallery, this analysis is not bounded by the Region I analysis, and Region II analyses shall be performed, bounding break sizes greater than the debris generation break size and up to the largest break size determined to be possible through the piping displacement and structure analysis.

This approach is to satisfy the third point of the proposed methodology because, as stated in NEI 04-07 (Reference 4-1) and modified by the SE on NEI 04-07 (Reference 4-2), Region II analyses are limited to a DEGB of the RCS main loop piping unless physically limited by piping restraints and supports, other plant structural members, or piping stiffness as may be demonstrated by analysis. The Region II analysis WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 4-21 is performed to demonstrate long-term cooling and mitigative capability using best estimate assumptions, of which piping displacement and/or structural analyses are permitted by the industry guidance.

As shown by the results of such analyses presented in subsections 4.4.3.2.2 and 4.4.3.2.3, the worst possible mechanical break displacement for all possible break locations inside the steam generator compartment and through the CA01 module penetration results in a ZOI that is smaller than that assumed in the Region 1 analysis. Therefore, when this ZOI is considered at all points along the RCS cold leg from the RCP through the CA01 module penetration to determine the location that will produce the maximum head loss across the sump screen as directed by the industry guidance, the resultant debris generation, and therefore resultant head loss, with no other best estimate assumptions considered, cannot be greater than the debris generation and resultant head loss determined from the Region 1 analysis.

As shown by the result presented in subsections 4.4.3.2.1 and 4.4.3.2.3, the worst possible mechanical break displacement for all possible break locations inside the reactor nozzle gallery from the reactor nozzle to the CA01 module wall was determined to occur at the cold leg reactor nozzle, and is greater than the debris generation break size used in the Region I analysis. Therefore, this maximum break size must be applied to the location along the RCS cold leg from the reactor nozzle to the CA01 module wall that is determined to produce the maximum debris generation and therefore the maximum head loss across the sump screen as part of the Region II analysis.

While taken in a segmented approach, the Region II cold leg break analysis follows the methods suggested by the industry guidance in Reference 4-1. Limiting break locations were identified for each segment of the AP1000 plant RCS cold leg; however, the resultant limiting break size was not to be limited to only the identified break locations. The limiting break size for each segment shall be applied to any location on the corresponding cold leg segment that results in the maximum generation of debris as suggested by the Reference 4-1 industry guidance.

4.5 REFERENCES

4-1. NEI 04-07, Pressurized Water Reactor Sump Performance Evaluation Methodology; Volume 1, Rev. 0, December 6, 2004.

4-2. NEI 04-07, Pressurized Water Reactor Sump Performance Evaluation Methodology; Volume 2 -

Safety Evaluation by the Office of Nuclear Reactor Regulation Related to NRC Generic Letter 2004-02, Rev. 0, December 6, 2004.

4-3. APP-GW-GL-700, Rev. 19, AP1000 Design Control Document, June 2011.

4-4. ANSI/ANS 58.2-1988, Design Basis for Protection of Light Water Nuclear Power Plants Against the Effects of Postulated Pipe Rupture, American Nuclear Society, October 6, 1988.

4-5. NUREG/CR-2913, Two-Phase Jet Loads, Sandia Labs, Albuquerque, NM January 1983.

4-6. APP-PL01-P0C-002, Pipe Reaction Force Determination for Large Break LOCA.

4-7. APP-PL01-P0C-003, RCL Pipe Movement in Large Break LOCAs.

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Westinghouse Non-Proprietary Class 3 4-22 4-8. CN-CRA-04-12, Revision 0, Jet Expansion Calculations Using ANSI/ANS 58.2-1988, Performed in Support of GSI-191 Methodology Development.

WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 5-1 5 NON-METALLIC INSULATION SUITABLE EQUIVALENCY The [

]a,c, is a suitable equivalent to MRI for the locations bounded by testing and analysis.

5.1.1 Debris Generation There is no debris production from the NMI contained within the neutron shield blocks or the water inlet doors. Their [ ]a,c construction maintains integrity under direct jet impingement outside a[ ]a,c The neutron shield blocks containing NMI and the water inlet doors are all located outside a a,c

[ ] for any potential pipe breaks in the reactor nozzle gallery; therefore, no debris will be generated from LOCA jet impingement or by floodup. The NMI in the shield blocks and the water inlet doors will not create chemical debris when submerged because of the [ ]a,c design.

There is no debris generation or chemical precipitate production from the NMI in the neutron shield blocks or in the water inlet doors.

5.1.1.1 Jet Impingement Debris The results of the jet impingement testing were used to determine the ZOI for the RVIS LNS, the CA31 neutron shielding, and the water inlet doors. With the ZOI defined, it is possible to verify that debris would not be generated from a break of the hot leg, cold leg, or DVI piping.

5.1.1.1.1 Establishing the Zone of Influence The ZOI is defined as the spherical volume about a break in which the fluid escaping from the break has sufficient energy to generate debris from insulation, coating, and other materials within the zone. The center of the spherical ZOI is located at the center of the break site.

The jet impingement testing was conducted on a [

]a,c.

a,c WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 5-2 The CA31 neutron shielding constructed of [

] a,c CA31 design provides additional margin against potential debris generation.

The water inlet doors constructed of [

]a,c 5.1.1.1.2 NEI-04-07 Approach to Determining the Zone of Influence The secondary approach to determining the ZOI requires experimentally determined impingement pressures within the jet for the location in which the neutron shielding test samples were positioned. The experimentally determined impingement pressures can be matched to the guidance report recommended ZOI values in Table 5-1 (Reference 5-1).

Table 5-1. NEI 04-07 Comparison of Computed Spherical ZOI Radii from Independent Evaluations of the ANSI Jet Model

• ZOI radius/break diameter ratios within the red box are used.

The jet impingement test facility used for the neutron shielding testing was previously used by the PWROG. Their testing utilized an array of pressure sensors to determine the impingement pressures at various locations within the jet. The centerline pressure data recorded during the PWROG testing (Reference 5-2) at a distance of [ ]a,c from the nozzle can be seen in Figure 5-1. The peak recorded pressure during the jet testing [ ]a,c psig.

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Westinghouse Non-Proprietary Class 3 5-3 The above approach for determining a ZOI is not endorsed by this submittal; it is provided to demonstrate conservatism in the empirically determined ZOI based on prototypic two-phase jet impingement testing.

a,b,c Figure 5-1. Centerline Stagnation Pressure at [ ]a,c from the Jet Nozzle The correlation between impingement pressure and the ZOI from NEI 04-07, shown in Table 5-1, provides information for several pressures. To find the ZOI associated with the experimentally determined pressure, the guidance report recommended ZOI radius/break diameter information be plotted (Figure 5-2) and trend lines be added. The resulting trend line equations provide a method of calculating the ZOI.

0.719 ZOI Radius 58.915 x P 0 < P < 333 psig

=

Break Diameter 1 P 333 psig P = Impingement Pressure (psig)

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Westinghouse Non-Proprietary Class 3 5-4 25 Pressures Greater than or Equal to 333 psig Pressures Less than 333 psig Linear (Pressures Greater than 20 or Equal to 333 psig)

Power (Pressures Less than 333 psig)

ZOI Radius/Break Diameter 15 10 y = 58.915x-0.719 5

y=1 0

0 200 400Impingement600 Pressure (psig)800 1000 1200 Figure 5-2. Plot of NEI 04-07 Spherical ZOI Radii As previously stated, the jet impingement testing of the [

]a,b,c 5.1.1.1.3 Debris Generation Assessment A LOCA debris generation assessment was performed for each potential pipe break in the reactor vessel cavity for the RVIS and the CA31 module neutron shielding. Figure 5-3 through Figure 5-6 provide the layout of the nozzle gallery area and the major equipment and structures located in the reactor cavity.

[

]a,c The RVIS LNS and the water inlet doors are far enough from the potential pipe breaks that they are not within a [ ]a,b,c. The LNS and the water inlet doors do not generate debris from a LOCA jet.

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Westinghouse Non-Proprietary Class 3 5-5 a,c Figure 5-3. Nozzle Gallery Plan View a,c Figure 5-4. Nozzle Gallery Elevation View at the DVI Piping Centerline WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 5-6 a,c Figure 5-5. Nozzle Gallery Elevation View at the Hot Leg Piping Centerline a,c Figure 5-6. Nozzle Gallery Elevation View at the Cold Leg Piping Centerline WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 5-7 5.1.1.1.3.1 Direct Vessel Injection Pipe Break The debris generation assessment of the DVI piping was performed for a DEGB. The ZOI was calculated using the full ID of the DVI pipe [ ]a,c. The assessment determined that the [

a,c

] CA31 neutron shielding would not generate fibrous or particulate debris from a break at the DVI nozzle. The ZOI would be small enough that the CA31 neutron shielding is outside the ZOI (Figure 5-7).

[

]a,c a,c Figure 5-7. Elevation View of DVI Break Zone of Influence [ ]a,c WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 5-8 a,b,c Figure 5-8. Plan View of DVI Break Zone of Influence [ ]a,c 5.1.1.1.4 Hot Leg Pipe Break As discussed in Section 4, NEI 04-07 (Reference 5-1) allows the use of an alternative methodology to evaluate the debris generation potential of breaks in the RCS main loop piping. The alternative methodology requires that a Region I and Region II analysis be performed in the debris generation assessment of the hot leg piping.

5.1.1.1.4.1 Region I Analysis A Region I analysis of the hot leg piping is evaluated for a maximum debris generation break size equivalent to the DEGB of a 14-inch Schedule 160 pipe. The spherical ZOI, used to assess the debris generation potential from a hot leg break was generated using the ID of a 14-inch Schedule 160 pipe (11.188 inches). The resulting ZOI sphere is small enough that the CA31 neutron shielding is outside the ZOI (Figure 5-9). The Region I analysis determined that no fibrous or particulate debris would be generated from encapsulated NMI at a break at the hot leg nozzles.

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Westinghouse Non-Proprietary Class 3 5-9 a,b,c Figure 5-9. Elevation View of Hot Leg Break Zone of Influence [ ]a,c a,b,c Figure 5-10. Plan View of Hot Leg Break Zone of Influence [ ]a,c WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 5-10 5.1.1.1.4.2 Region II Analysis Region II analyses allow for more realistic analysis methods and assumptions, such as limited pipe displacement, operation of non-safety systems, intervening structures, and operator actions. A pipe break analysis was performed for a break at the reactor vessel hot leg nozzle to determine the amount of movement and the resulting break area. The analysis showed that the resulting pipe break area was less than the break area of a DEGB of a 14-inch Schedule 160 pipe. Because the break size used in the Region I analysis is greater than the break size used in the Region II analysis, the Region I analysis findings are more conservative.

5.1.1.1.5 Cold Leg Pipe Break Similar to the hot leg piping, the debris generation assessment for the cold leg piping utilized the alternative methodology as discussed in Section 4. A Region I and Region II analysis was performed for the cold leg piping.

5.1.1.1.5.1 Region I Analysis The spherical ZOI used to assess the debris generation potential from a cold leg break was generated using the ID of a 14-inch Schedule 160 pipe (11.188 inches). The resulting ZOI sphere is small enough that the CA31 neutron shielding is outside the ZOI (Figure 5-11). The Region I analysis determined that no debris would be generated from a break at the cold leg nozzles.

a,b,c Figure 5-11. Elevation View of Region I Analysis Cold Leg Break Zone of Influence [ ]a,c WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 5-11 a,b,c Figure 5-12. Plan View of Region I Analysis Cold Leg Break Zone of Influence [ ]a,c 5.1.1.1.5.2 Region II Analysis As with the hot leg piping, a pipe break analysis was performed for a break at the reactor vessel cold leg nozzle to determine the amount of movement of the cold leg piping and the resulting break area. The analysis indicated that [

]a,c The Region II analysis allows for more realistic assumptions to be used in the debris generation assessment, such as taking credit for intervening robust structures. Reference 5-1 states that when the spherical ZOI extends beyond robust barriers, such as walls, or encompasses large components, the extended volume can be truncated.

The CA31 neutron shielding is located [

]a,c WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 5-12 a,c Figure 5-13. CA31 Module and Neutron Blocks Top View; Top Liner Plate Removed The CA31 neutron shielding is outside the ZOI (Figure 5-14). Both the Region I and Region II analyses have shown that no particulate or fibrous debris would be generated from a break at the cold leg nozzle.

a,b,c Figure 5-14. Elevation View of Cold Leg Break Zone of Influence [ ]a,c WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 5-13 a,b,c Figure 5-15. Plan View of Cold Leg Break Zone of Influence [ ]a,c 5.1.1.2 Submergence Debris The design of the neutron shield blocks and water inlet doors is such that little to no communication between the components and the post-LOCA fluid would occur since:

• [

]a,c

• [

]a,c

• The current design [ ]a,c neutron shield blocks do not produce debris.

• The water inlet doors do not produce fiber, particulate, or chemical debris. The [

]a,c water inlet doors are outside of any high energy line break ZOI and are not subject to jet impingement damage. The [ ]a,c design also prevents interaction between the post LOCA fluid and the encapsulated NMI resulting in no post LOCA chemical implications.

Therefore, no additional chemical effects have to be considered with respect to the RVIS.

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Westinghouse Non-Proprietary Class 3 5-14 5.1.2 Aging Effects The test samples used in the submergence and jet impingement testing were not thermally or radiation aged. The effects of thermal and radiation aging would not impact the conclusion that the neutron shielding and water inlet doors would not generate debris as a result of a LOCA jet or submergence. The materials used in CA31 neutron shielding, LNS, and RVIS water inlet doors are a combination of

[

]a,c 5.1.2.1 Aging of CA31 Shielding Material The aging mechanisms acting on [xxxxx xxxxxx]a,c contained within CA31 include chemical, nuclear, and thermal reactions. The impact of these aging mechanisms on [xxxxx xxxxxx]a,c has been studied and found to be acceptable from a sixty year plant life perspective. The following paragraphs summarize the results of calculations that investigated these effects.

[

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Westinghouse Non-Proprietary Class 3 5-15

]a,c 5.1.3 Thermal Expansion of LNS Neutron Shielding Material

[

]a,b,c WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 5-16 5.1.4 Additional Conservatisms There are several conservatisms that can be taken under consideration when determining the suitable equivalency of the NMI in the neutron shield blocks and water inlet doors. Some of the conservatisms are qualitative, while others are more quantitative. All are discussed herein.

In the process of updating the chemical effects calculation, the allowable amount of [

]a,b,c The AP1000 plant chemical effects model was updated to reflect the encapsulated NMI as a suitable equivalent to MRI but assumed that the [

]a,b,c can be found in Table 5-2 (Reference 5-8).

Table 5-2. Chemical Debris Generation for AP1000 Plant Licensing Basis Case a,b,c X

s X

sdx The limiting factor in the chemical effects model is the amount of aluminum dissolved in the sump fluid.

[

]a,c, the debris load remains below the licensing basis limit of 57 pounds.

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Westinghouse Non-Proprietary Class 3 5-17 Additionally, the jet impingement testing on the [ ]a,c neutron shield blocks was performed at [ ]a,c on both fully exposed blocks and blocks with a replicate reactor cavity cover plate. [

]a,c The submergence test also had a few inherent conservatisms in the test procedure. The blocks were [

]a,c The sampling system for the neutron shield block submergence testing [

]a,c

5.2 REFERENCES

5-1. NEI 04-07, Pressurized Water Reactor Sump Performance Evaluation Methodology, ,

Volume 2 - Safety Evaluation by the Office of Nuclear Reactor Regulation Related to NRC Generic Letter 2004-02, Revision 0, December 6, 2004.

5-2. Report No.: FAI/110497, Revision 1, PWROG Model for the Two Dimensional Free Expansion of a Flashing, Two Phase, Critical Flow Jet, February 2012.

5-3. APP-CA31-S5-404, Containment Building Areas 1, 2, 3, & 4 Module CA31 El 107'-2" Installation Sequence III.

5-4. Haley, T. (2012), Boraflex, RACKLIFE, and BADGER.

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Westinghouse Non-Proprietary Class 3 5-18 5-5. APP-CA31-S5B-002, Containment Building Areas 1, 2, 3, & 4 Module CA31 El 107-2 Bill of Materials II.

5-6. EPRI Report Materials Reliability Program A Review of Radiation Embrittlement of Stainless Steels for PWRs (MRP-79) - Revision 1 1008204.

5-7. APP-PXS-M3C-221, Aluminum Inventory for AP1000 Containment.

5-8. APP-PXS-M3C-052, AP1000 GSI-191 Chemistry Effects Evaluation.

5-9. DCP-DCP-008462, MN20 Neutron Shield Block Thermal Testing.

5-10. DCP_DCP-008711, Boron-Silicone Neutron Shield Block Thermal Testing.

5-11. APP-VCS-M3C-011, Thermal Analysis for MN20 Lower Neutron Shield Blocks.

5-12. APP-PXS-M3C-057, Loss of Coolant Accident Deposition Model (LOCADM) Analysis for AP1000 Plant Design.

5-13. APP-NS27-GEC-001, Analyses and Interpretations of Test Results for [Xxxxx Xxxxxxx]a,c Material.

5-14. APP-CA31-N4C-001, An Evaluation and Analysis of Gasses Generated within AP1000 CA31 Neutron Shielding Materials.

5-15. APP-CA31-GEC-003, Expansion Calculations for [Xxxxx Xxxxxxx]a,c in the CA31 Module.

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Westinghouse Non-Proprietary Class 3 6-1 6 REGULATORY IMPACTS Revisions to the AP1000 plant licensing basis are proposed in support of the integrated debris evaluation approach. Appendix A of this report provides representative examples of proposed Final Safety Analysis Report (FSAR) markups based on the NRC reviewed and approved AP1000 plant DCD Revision 19 (Reference 6-1). The final changes to the FSAR will be made by the Licensee, with an evaluation performed in accordance with Title 10 of the Code of Federal Regulations, Part 52, Appendix D, Volume VIII.

6.1 LICENSING BASIS CHANGES It should be noted that this report does not propose any changes to AP1000 plant DCD / Updated Final Safety Analysis Report (UFSAR) Tier 2*, DCD Tier 1 / Combined Operating License (COL) Appendix C or Technical Specifications (COL Appendix A).

The changes to DCD/UFSAR Tier 2 are as follows:

• In DCD/UFSAR subsection 6.3.2.2.7.1, item 3, this WCAP is added as a reference that demonstrates suitable equivalency for the non-metallic insulation in the AP1000 plant reactor vessel neutron shield blocks. The shielding locations that were considered in this report are detailed in Section 2.2.

• In DCD/UFSAR subsection 6.3.2.2.7.1, item 12, it is noted that D can be determined using double ended guillotine break of primary system piping or an alternate debris generation size from the alternate evaluation approach of NEI 04-07 (Reference 6-3). The usage of the alternate evaluation approach used is detailed in Section 4 of this report.

• In DCD/UFSAR subsection 6.3.2.2.7.1, item 12, this WCAP is added as a reference that supports a 4D ZOI radius for AP1000 plant in-containment cables.

• In DCD/UFSAR subsection 6.3.9, this WCAP and NEI 04-07 are included in the Section 6.3 references subsection.

• In DCD/UFSAR subsection 9.5.1.2.1.1 under Control of Combustible Material a statement is added related to off gassing from neutron shield blocks as presented in RAI-ICC&NMI-036 (Reference 6-4 and Reference 6-5) .

6.2 REFERENCE 6-1. APP-GW-GL-700, Rev. 19, AP1000 Design Control Document, June 2011.

6-2. NEI 04-07, Pressurized Water Reactor Sump Performance Evaluation Methodology; Volume 1, Revision 0, December 6, 2004.

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Westinghouse Non-Proprietary Class 3 6-2 6-3. NEI 04-07, Pressurized Water Reactor Sump Performance Evaluation Methodology; Volume 2 -

Safety Evaluation by the Office of Nuclear Reactor Regulation Related to NRC Generic Letter 2004-02, Revision 0, December 6, 2004.

6-4. APP-GW-GLY-124, Responses to NRC Request for Addition Information (RAI) Letter No. 02 for WCAP-17938 (Proprietary), November 2014.

6-5. APP-GW-GLY-125, Responses to NRC Request for Addition Information (RAI) Letter No. 02 for WCAP-17938 (Non-Proprietary), November 2014.

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Westinghouse Non-Proprietary Class 3 7-1 7 CONCLUSIONS Cable jet impingement testing was performed on AP1000 plant in-containment cables to determine the appropriate ZOI radius for use in cable debris generation assessments for the AP1000 plant. The results of the testing demonstrated that the onset of incipient damage, as defined by NEI-04-07 Volume 2, from a bounding DECLGB, which is limiting as related by maximum subcooling, was found to be [

]a,c where D is taken to be the pipe diameter as defined in the applicable Region I and II analyses. For conservatism, the ZOI radius is substantiated to be [

]a,c. As a result, all cables inside the AP1000 plant containment are to be located outside of a 4D ZOI radius from all high-energy lines or they must be protected from jet impingement. AP1000 plant in-containment cables are analyzed with respect to this ZOI in Reference 7-1. For all cable locations identified in Reference 7-1 as within the 4D ZOI of a high-energy line, a disposition scheme has been identified including jet impingement protection or movement of the affected cables.

Encapsulated NMI representative of that located in CA31 and the reactor vessel cavity of the AP1000 plant was subjected to jet impingement and submergence testing to qualify the insulation system as a suitable equivalent per the AP1000 plant licensing basis. Results of the jet impingement and submergence testing resulted in a change in the design [

]a,c The geometry of the reactor vessel cavity and the influence of confined jet behavior were addressed in subsection 3.5.3.2 and confirmed the jet impingement testing bounded the plant conditions. The jet impingement testing at

[ ]a,b,c subjected the Type III block to integrated pressures [ ]a,c than the maximum pressures that the blocks would be subjected to in the plant. The [ ]a,c intrinsic to the Type III design ensures no communication with sump fluid in the post-LOCA containment. This ensures no chemical debris is generated by the Type III design. Therefore, the encapsulated NMI in the Type III block design is a suitable equivalent insulation to MRI.

A methodology was presented to direct performance of Region I and Region II analyses for debris generation. The methodology included the assumption in the Region I analyses of a 14-inch Schedule 160 pipe diameter analyzed under DBA assumptions for debris generation of RCS main loop piping. The Region I analysis assumed a DEGB for any RCS loop piping offtakes. These assumptions are commensurate with those contained in References 7-2 and 7-3.

The Region II analyses contain a pipe stiffness calculation to determine the break configuration for RCS loop piping. The results of the pipe stiffness calculation are used to develop the equivalent pipe diameter for determining the appropriate ZOI radius for use in Region II debris generation calculations.

The Region II assessment includes the methodology for determining whether a limited separation break configuration jet volume bounds the assumed spherical ZOI geometry based on the quantified results of the pipe stiffness analysis. The largest ZOI volume is chosen to be limiting for performing debris generation calculations.

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Westinghouse Non-Proprietary Class 3 7-2 This methodology is applicable to:

• ZOI radius determination for AP1000 plant in-containment cables

• Demonstration of AP1000 plant reactor vessel cavity NMI suitable equivalency for the locations that were bounded by testing and analysis

• Performance of debris generation calculations consistent with the alternate method for Region I and Region II analyses In conclusion, this document communicates and substantiates the results of the cable jet impingement test program, the NMI suitable equivalency program, and the application of a conservative and bounding evaluation methodology to determine the ZOI radius for Region I and Region II analyses for an integrated debris assessment.

7.1 REFERENCES

7-1. APP-PXS-M3C-080, AP1000 Non-Coating Debris Contributions Towards GSI-191 Debris Limits.

7-2. NEI 04-07, Pressurized Water Reactor Sump Performance Evaluation Methodology; Volume 1, Revision 0, December 6, 2004.

7-3. NEI 04-07, Pressurized Water Reactor Sump Performance Evaluation Methodology; Volume 2 -

Safety Evaluation by the Office of Nuclear Reactor Regulation Related to NRC Generic Letter 2004-02, Revision 0, December 6, 2004.

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Westinghouse Non-Proprietary Class 3 A-1 APPENDIX A AP1000 DCD REVISION 19 MARKUPS 6.3.2.2.7.1 General Screen Design Criteria

1. Screens are designed to Regulatory Guide 1.82, including:

• Separate, large screens are provided for each function.

• Screens are located well below containment floodup level. Each screen provides the function of a trash rack and a fine screen. A debris curb is provided to prevent high density debris from being swept along the floor to the screen face.

• Floors slope away from screens (not required for AP1000).

• Drains do not impinge on screens.

• Screens can withstand accident loads and credible missiles.

• Screens have conservative flow areas to account for plugging. Operation of the non-safety-related normal residual heat removal pumps with suction from the IRWST and the containment recirculation lines is considered in sizing screens.

• System and screen performance are evaluated.

• Screens have solid top cover. Containment recirculation screens have protective plates that are located no more than 1 foot above the top of the screens and extend at least 10 feet in front and 7 feet to the side of the screens. The plate dimensions are relative to the portion of the screens where water flow enters the screen openings. Coating debris, from coatings located outside of the ZOI, is not transported to the containment recirculation screens, to the IRWST screens, or into a direct vessel injection or a cold leg LOCA break that becomes submerged during recirculation considering the use of high density coatings discussed in subsection 6.1.2.1.5.

• Screens are seismically qualified.

• Screen openings are sized to prevent blockage of core cooling.

• Screens are designed for adequate pump performance. AP1000 has no safety-related pumps.

• Corrosion resistant materials are used for screens.

• Access openings in screens are provided for screen inspection.

• Screens are inspected each refueling.

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Westinghouse Non-Proprietary Class 3 A-2

2. Low screen approach velocities limit the transport of heavy debris even with operation of normal residual heat removal pumps.

3. [Metal reflective insulation is used on ASME class 1 lines because they are subject to loss-of-coolant accidents. Metal reflective insulation is also used on the reactor vessel, the reactor coolant pumps, the steam generators, and on the pressurizer because they have relatively large insulation surface areas and they are located close to large ASME class 1 lines. As a result, they are subject to jet impingement during loss-of-coolant accidents.]* A suitable equivalent insulation to metal reflective may be used. A suitable equivalent insulation is one that is encapsulated in stainless steel that is seam welded so that LOCA jet impingement does not damage the insulation and generate debris. Another suitable insulation is one that may be damaged by LOCA jet impingement as long as the resulting insulation debris is not transported to the containment recirculation screens, to the IRWST screens, or into a direct vessel injection or a cold leg LOCA break that becomes submerged during recirculation. In order to qualify as a suitable equivalent insulation, testing must be performed that subjects the insulation to conditions that bound the AP1000 conditions and demonstrates that debris would not be generated. If debris is generated, testing and/or analysis must be performed to demonstrate that the debris is not transported to an AP1000 screen or into the core through a flooded break. It would also have to be shown that the material used would not generate chemical debris. In addition, the testing and/or analysis must be approved by the NRC.

[In order to provide additional margin, metal reflective insulation is used inside containment where it would be subject to jet impingement during loss-of-coolant accidents that are not otherwise shielded from the blowdown jet.]* As a result, fibrous debris is not generated by loss-of-coolant accidents. Insulation located within the zone of influence (ZOI), which is a spherical region within a distance equal to 29 inside diameters (for Min-K, Koolphen-K, or rigid cellular glass insulation) or 20 inside diameters (for other types of insulation) of the LOCA pipe break is assumed to be affected by the LOCA when there are intervening components, supports, structures, or other objects.

[The ZOI in the absence of intervening components, supports, structures, or other objects includes insulation in a cylindrical area extending out a distance equal to 45 inside diameters from the break along an axis that is a continuation of the pipe axis and up to 5 inside diameters in the radial direction from the axis.]* A suitable equivalent insulation to metal reflective may be used as discussed in the previous paragraph.

[Insulation used inside the containment, outside the ZOI, but below the maximum post-DBA LOCA floodup water level (plant elevation 110.2 feet), is metal reflective insulation, jacketed fiberglass, or a suitable equivalent.]* A suitable equivalent insulation is one that would be restrained so that it would not be transported by the flow velocities present during recirculation and would not add to the chemical precipitates. In order to qualify as a suitable equivalent insulation, testing must be performed that subjects the insulation to conditions that bound the AP1000 conditions and demonstrates that debris would not be generated. If debris is generated, testing and/or analysis must be performed to demonstrate that the debris is not transported to an AP1000 screen or into the core through a flooded break. It would also have to be shown that the WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 A-3 material used would not generate chemical debris. In addition, the testing and/or analysis must be approved by the NRC.

[Insulation used inside the containment, outside the ZOI, but above the maximum post-design basis accident (DBA) LOCA floodup water level, is jacketed fiberglass, rigid cellular glass, or a suitable equivalent.]* A suitable equivalent insulation is one that when subjected to dripping of water from the containment dome would not add to the chemical precipitates; suitable equivalents include metal reflective insulation.

The non-metallic insulation used in the AP1000 reactor vessel neutron shield blocks and the water inlet doors has been determined to be a suitable equivalent for the locations bounded by testing and analysis (Reference 5).

4. Coatings are not used on surfaces located close to the containment recirculation screens. The surfaces considered close to the screens are defined in subsection 6.3.2.2.7.3. Refer to subsection 6.1.2.1.6. These surfaces are constructed of materials that do not require coatings.

5. The IRWST is enclosed which limits debris egress to the IRWST screens.

6. Containment recirculation screens are located above lowest levels of containment.

7. Long settling times are provided before initiation of containment recirculation.

8. Air ingestion by safety-related pumps is not an issue in the AP1000 because there are no safety-related pumps. The normal residual heat removal system pumps are evaluated to show that they can operate with minimum water levels in the IRWST and in the containment.

9. A commitment for cleanliness program to limit debris in containment is provided in subsection 6.3.8.1.

10. [Other potential sources of fibrous material, such as ventilation filters or fiber-producing fire barriers, are not located in jet impingement damage zones or below the maximum post-DBA LOCA floodup water level.]*

11. Other potential sources of transportable material, such as caulking, signs, and equipment tags installed inside the containment are located:

• Below the maximum flood level, or

• Above the maximum flood level and not inside a cabinet or enclosure.

Tags and signs in these locations are made of stainless steel or another metal that has a density 100 lbm/ft3. Caulking in these locations is a high density ( 100 lbm/ft3).

The use of high-density metal prevents the production of debris that could be transported to the containment recirculation screens, to the IRWST screens, or into a direct vessel injection or a cold leg LOCA break location that is submerged during recirculation. If a high-density material is not WCAP-17938-NP April 2018 APP-GW-GSR-013 Revision 3

Westinghouse Non-Proprietary Class 3 A-4 used for these components, then the components must be located inside a cabinet or other enclosure, or otherwise shown not to transport; the enclosures do not have to be watertight, but need to prevent water dripping on them from creating a flow path that would transport the debris outside the enclosure. For light-weight (< 100 lbm/ft3) caulking, signs or tags that are located outside enclosures, testing must be performed that subjects the caulking, signs, or tags to conditions that bound the AP1000 conditions and demonstrates that debris would not be transported to an AP1000 screen or into the core through a flooded break. Note that in determining if there is sufficient water flow to transport these materials, consideration needs to be given as to whether they are within the ZOI (for the material used) because that determines whether they are in their original geometry or have been reduced to smaller pieces. It would also have to be shown that the material used would not generate chemical debris. In addition, the testing must be approved by the NRC.

12. An evaluation consistent with Regulatory Guide 1.82, Revision 3, and subsequently approved NRC guidance, has been performed (Reference 3) to demonstrate that adequate long-term core cooling is available considering debris resulting from a LOCA together with debris that exists before a LOCA. As discussed in subsection 6.3.2.2.7.1, a LOCA in the AP1000 does not generate fibrous debris due to damage to insulation or other materials included in the AP1000 design. The evaluation considered resident fibers and particles that could be present considering the plant design, location, and containment cleanliness program. The determination of the characteristics of such resident debris was based on sample measurements from operating plants. The evaluation also considered the potential for the generation of chemical debris (precipitants). The potential to generate such debris was determined considering the materials used inside the AP1000 containment, the post-accident water chemistry of the AP1000, and the applicable research/testing.

The evaluation considered the following conservative considerations:

• [The COL cleanliness program will limit the total amount of resident debris inside the containment to 130 pounds and the amount of the total that might be fiber to 6.6 pounds.]*

• In addition to the resident debris, the LOCA blowdown jet may impinge on coatings and generate coating debris fines, which because of their small size, might not settle. The amount of coating debris fines that can be generated in the AP1000 by a LOCA jet will be limited to less than 70 pounds for double-ended cold leg and double-ended direct vessel injection LOCAs. In evaluating this limit, a ZOI of 4 IDs for epoxy and 10 IDs for inorganic zinc will be used. A DEHL LOCA could generate more coating debris; however, with the small amount of fiber available in the AP1000 following a LOCA, the additional coating debris fines that may be generated in a DEHL LOCA are not limiting.

• For ZOIs used in the debris evaluation, the diameter can be determined from the double ended guillotine break of primary system piping or the alternate debris generation size (Reference 5) developed in accordance with the alternate evaluation approach from NEI 04-07 (Reference 6).

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Westinghouse Non-Proprietary Class 3 A-5

• The total resident and ZOI coating debris available for transport following a LOCA is 193.4 pounds of particulate and 6.6 pounds of fiber. The percentage of this debris that could be transported to the screens or to the core is as follows:

- Containment recirculation screens is 100 percent fiber and particles

- IRWST screens is 50 percent fiber and 100 percent particles

- Core (via a direct vessel injection or a cold leg LOCA break that becomes submerged) is 90 percent fiber and 100 percent particles

• Fibrous insulation debris is not generated and transported to the screens or into the core as discussed in Item 3.

• Metal reflective insulation, including accident generated debris, is not transported to the screens or into the core.

• Coating debris is not transported to the screens or into the core as discussed in Item 1.

• Debris from other sources, including caulking, signs, and tags, is not generated and transported to the screens or into the core as discussed in Item 11.

• A ZOI radius of 4 IDs is applicable to the AP1000 in-containment cables bounded by testing and analysis, consistent with Reference 5.

• The total amount of chemical precipitates that could form in 30 days is 57 pounds.

• The percentage of the chemical precipitates that could be transported to the:

- Containment recirculation screens is 100 percent.

- IRWST screens is 100 percen.

- Core is 100 percent.

• The range of flow rates during post-LOCA injection and recirculation is as follows:

- CR screens: 2320 to 539 gpm

- IRWST screens: 2320 to 464 gpm

- Core: 2012 to 484 gpm These flows bound operation of the PXS and the RNS. Note that if the RNS operates during post-LOCA injection or recirculation, the RNS flow is limited to 2320 gpm. This limit ensures that the operation of the plant is consistent with screen head loss testing. In addition, the screens will be designed structurally to withstand much higher flow rates and pressure losses to provide appropriate margin during PXS and RNS operation.

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Westinghouse Non-Proprietary Class 3 A-6 No chemical precipitates are expected to enter the IRWST because the primary water input to the IRWST is steam condensed on the containment vessel. However, during a direct vessel injection LOCA, recirculation can transport chemical debris through the containment recirculation screens and to the IRWST screens. As a result, 100 percent of the chemical debris is conservatively assumed to be transported to the IRWST screens.

The AP1000 containment recirculation screens and IRWST screens have been shown to have acceptable head losses. The head losses for these screens were determined in testing performed using the above conservative considerations. It has been shown that a head loss of 0.25 psi at the maximum screen flows is acceptable based on long-term core cooling sensitivity analysis.

Considering downstream effects as well as potential bypass through a cold leg LOCA, the core was shown to have acceptable head losses. The head losses for the core were determined in testing performed using the above conservative considerations. It has been shown that a head loss of 4.1 psi at these flows is acceptable based on long-term core cooling sensitivity analysis.

6.3.8.2 Verification of Water Sources for Long-Term Recirculation Cooling Following a LOCA The Combined License information requested in this subsection has been fully addressed in APP-GW-GLR-079 (Reference 3), and the applicable changes are incorporated into the DCD. The design of the recirculation screens is complete. Testing to assess the screen performance and downstream effects is complete. A study of the effects of screen design and performance on long-term cooling is complete. No additional work is required by the Combined License applicant to address the aspects of the Combined License information requested in this subsection.

The following words represent the original Combined License Information Item commitment, which has been addressed as discussed above:

The Combined License applicants referencing the AP1000 will perform an evaluation consistent with Regulatory Guide 1.82, Rev.n 3, and subsequently approved NRC guidance, to demonstrate that adequate long-term core cooling is available considering debris resulting from a LOCA together with debris that exists before a LOCA. As discussed in DCD subsection 6.3.2.2.7.1, a LOCA in the AP1000 does not generate fibrous debris due to damage to insulation or other materials included in the AP1000 design. The evaluation will consider resident fibers and particles that could be present considering the plant design, location, and containment cleanliness program. The determination of the characteristics of such resident debris will be based on sample measurements from operating plants. The evaluation will also consider the potential for the generation of chemical debris (precipitants). The potential to generate such debris will be determined considering the materials used inside the AP1000 containment, the post-accident water chemistry of the AP1000, and the applicable research/testing.

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Westinghouse Non-Proprietary Class 3 A-7 6.3.9 References

1. WCAP-8966, Evaluation of Mispositioned ECCS Valves, September 1977.

2. WCAP-13594 (P), WCAP-13662 (NP), FMEA of Advanced Passive Plant Protection System, Revision 1, June 1998.

3. APP-GW-GLR-079, AP1000 Verification of Water Sources for Long-Term Recirculation Cooling Following a LOCA, Westinghouse Electric Company LLC.

4. APP-GW-GLN-147, AP1000 Containment Recirculation and IRWST Screen Design, Westinghouse Electric Company LLC.

5. WCAP-17938 (P-A/NP-A), AP1000 In-Containment Cables and Non-Metallic Insulation Debris Assessment, Revision 3, May 2018.

6. NEI 04-07, Pressurized Water Reactor Sump Performance Evaluation Methodology Revision 0.

9.5.1.2.1.1 Plant Fire Prevention and Control Features Control of Combustible Material The plant is constructed of noncombustible materials to the extent practicable. The selection of construction materials and the control of combustible materials are in accordance with BTP CMEB 9.5-1 and Section 3.3 of NFPA 804 (Reference 2) as specified in WCAP-15871 (Reference 20).

The storage and use of hydrogen are according to NFPA 50A and NFPA 50B (Reference 2). Hydrogen lines in safety-related areas are designed to seismic Category I requirements.

Ventilation systems are designed to maintain the hydrogen concentration in the battery rooms well below 2 percent by volume, as described in subsections 9.4.1 and 9.4.2.

The containment recirculation cooling system (VCS, subsection 9.4.6) provides sufficient air flow through the reactor vessel cavity to mix any off-gassing from the neutron shield blocks with the containment volume. The containment air filtration system (VFS, subsection 9.4.7) provides periodic flow of outdoor air to purge and filter the containment atmosphere. The mixing flow rate and containment purging maintains the hydrogen concentration in containment well below 2 percent by volume.

The turbine lubrication oil system, located in the turbine building, is separated from areas containing safety-related equipment by 3-hour rated fire barriers.

Outdoor oil-filled transformers are separated from plant buildings according to NFPA 804 (Reference 2).

The diesel fuel oil storage tanks and the diesel fuel oil transfer pump enclosure are located in the yard area more than 50 feet from any safety-related structure. Potential oil spills from the storage tanks are confined by a diked enclosure. A diesel generator fuel day tank is located within each diesel generator room and is enclosed in a 3-hour fire rated barrier.

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Westinghouse Non-Proprietary Class 3 A-8 The diesel fuel supply for the ancillary diesel generators is in the same room as the diesel generators. The ancillary diesel generator room is separated from the rest of the annex building by a 3-hour rated fire barrier.

The diesel fuel supply for the diesel-driven fire pump is in the diesel-driven fire pump enclosure. The diesel pump enclosure is located in the yard more than 50 feet from safety-related structures. The enclosure includes a fire detector which produces an audible alarm locally with both visual and audible alarms in the main control room and security central alarm station. The fire is extinguished by operation of an automatic sprinkler system or manually, using hose streams or portable extinguishers.

Quantities and locations of other combustible materials are identified in the fire protection analysis (see Appendix 9A).

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