ML110060436

MUAP-09026(R5), US-APWR DCD Revision 2 RAI Tracking Report

ML110060436
Person / Time
Site:	05200021
Issue date:	12/31/2010
From:	Mitsubishi Heavy Industries, Ltd
To:	Office of New Reactors
References
UAP-HF-10329 MUAP-09026(R5)
Download: ML110060436 (192)

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US-APWR DCD Revision 2 RAI Tracking Report MUAP-09026(R5)

US-APWR DCD Revision 2 RAI Tracking Report December 2010 C 2010 Mitsubishi Heavy Industries, Ltd.

C All Rights Reserved Mitsubishi Heavy Industries, LTD.

US-APWR DCD Revision 2 RAI Tracking Report MUAP-09026(R5)

Revision History Revision Page Description Original issued 0 All Including RAI responses that were submitted through October 31, 2009 Including RAI responses that were submitted through 1 All December 31, 2009 Including RAI responses that were submitted through February 28, 2010 2 All Including editorial changes to clarify the English language and to correct typographical errors Including RAI responses that were submitted through April 30, 2010 3 All Including editorial changes to clarify the English language and to correct typographical errors Including RAI responses that were submitted through August 31, 2010 4 All Including editorial changes to clarify the English language and to correct typographical errors Including RAI responses that were submitted through October 30, 2010 5 All Including editorial changes to clarify the English language and to correct typographical errors Mitsubishi Heavy Industries, LTD.

US-APWR DCD Revision 2 RAI Tracking Report MUAP-09026(R5)

© 2010 MITSUBISHI HEAVY INDUSTRIES, LTD.

All Rights Reserved This document has been prepared by Mitsubishi Heavy Industries, Ltd. (MHI) in connection with the U.S. Nuclear Regulatory Commissions (NRC) licensing review of MHIs US-APWR nuclear power plant design. No right to disclose, use or copy any of the information in this document, other than by the NRC and its contractors in support of the licensing review of the US-APWR, is authorized without the express written permission of MHI.

This document contains technology information and intellectual property relating to the US-APWR and it is delivered to the NRC on the express condition that it not be disclosed, copied or reproduced in whole or in part, or used for the benefit of anyone other than MHI without the express written permission of MHI, except as set forth in the previous paragraph.

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

Mitsubishi Heavy Industries, Ltd.

16-5, Konan 2-chome, Minato-ku Tokyo 108-8215 Japan Mitsubishi Heavy Industries, LTD.

US-APWR DCD Revision 2 RAI Tracking Report MUAP-09026(R5)

General Description This report includes a table that identifies the impact of each response to the Request for Additional Information (RAI) relative to the Design Control Document (DCD) Revision 2 of US-APWR. This table shows the RAI responses which have been submitted since October 2009 and also should be incorporated into Tracking Report and DCD in future revision.

The report also includes the DCD Markups and Revision List for the RAI responses that impacted the DCD. Furthermore, the editorial changes to clarify the English language and to correct typographical errors are shown in the DCD Markups and Revision List.

Contents For ease of using this Tracking Report, each chapter is organized in a stand alone fashion that includes a cover sheet and the following relevant information:

• DCD Revision List - a list of the revision resulting from RAI responses and others changes

• DCD Markups - a copy of the DCD pages that have changes resulting from RAI responses or others change.

Mitsubishi Heavy Industries, LTD.

Chapter:6 SRP Section DCD RAI Response DCD Change ID Number for Tracking DCD Impact Impact Impact Other Drivers DCD forthcoming RAI Question Response Response Report Revision No. Title on on on Revision No. No. Date Status Revision DCD COLA PRA 6.1.1 Engineered Safety Features 2009/7/10 Y N N - DCD_06.01.01-9 4 2 379 06.01.01-9 Materials 2010/3/30 Y N N - DCD_06.01.01-9 3 3 379 06.01.01-10 2009/7/10 Y N N - DCD_06.01.01-10 4 2 487 06.01.01-11 2009/12/3 Y N N - DCD_06.01.01-11 1 3 487 06.01.01-12 2009/12/3 N N N - - N/A N/A 544 06.01.01-13 2010/4/21 Y N N - DCD_06.01.01-13 3 3 544 06.01.01-14 2010/4/21 Y N N - DCD_06.01.01-14 3 3 544 06.01.01-15 2010/4/21 N N N - - N/A N/A 544 06.01.01-16 2010/4/21 N N N - - N/A N/A 544 06.01.01-17 2010/4/21 Y N N - DCD_06.01.01-17 3 3 544 06.01.01-18 2010/4/21 N N N - - N/A N/A 544 06.01.01-19 2010/4/21 Y N N - DCD_06.01.01-19 3 3 612 06.01.01-20 2010/8/25 Y N N - DCD_06.01.01-20 5 3 612 06.01.01-21 2010/8/25 Y N N - DCD_06.01.01-21 TBD 2010/8/25 Y N N - 5 3 612 06.01.01-22 2010/10/7 Y N N - DCD_06.01.01-22 5 3 612 06.01.01-23 2010/8/25 Y N N - DCD_06.01.01-23 5 3 Protective Coating Systems 6.1.2 (Paints)

Organic Materials 6.2.1 Containment Functional Design 587 06.02.01.01.A-1 2010/6/7 N N N - - N/A N/A Organic Materials 623 06.02.01-18 2010/9/29 N N N - - N/A N/A 623 06.02.01-19 2010/9/29 Y N N - - TBD 623 06.02.01-20 2010/9/29 N N N - - N/A N/A 6.2.1.2 Subcompartment Analysis 6.2.1.3 Mass and Energy Release Analysis for Postulated Loss-of-Coolant Accidents 6.2.1.4 Mass and Energy Release Analysis for Postulated Secondary System Pipe Ruptures (LOCAs) 6.2.1.5 Min. Containment Pressure Analysis for for Emergency Core Cooling Sys.

Performance Capability Studies 6.2.2 Containment 85 06.02.02-10 2009/11/12 Y N N fin. - DCD_06.02.02-10 1 2 Heat Removal Systems 85 06.02.02-11 2009/11/12 N N N fin. - - N/A N/A 263 06.02.02-12 2009/3/31 Y N N - DCD_06.02.02-12 3 2 263 06.02.02-13 2009/3/31 N N N - - N/A N/A 263 06.02.02-14 2009/3/31 N N N - - N/A N/A 263 06.02.02-15 2009/3/31 N N N - - N/A N/A 278 06.02.02-16 2009/4/10 Y N N - DCD_06.02.02-16 3 2 330 06.02.02-17 2009/5/18 N N N - - N/A N/A 349 06.02.02-18 2009/5/12 N N N - - N/A N/A 354 06.02.02-19 2009/7/7 N N N - - N/A N/A 354 06.02.02-20 2009/7/7 N N N - - N/A N/A 354 06.02.02-21 2009/7/7 N N N - - N/A N/A 354 06.02.02-22 2009/7/7 N N N - - N/A N/A 354 06.02.02-23 2009/7/7 N N N - - N/A N/A 2009/7/7 N N N - - N/A N/A 354 06.02.02-24 10/16/2009 Y N N - DCD_06.02.02-24 TBD 354 06.02.02-25 2009/7/7 Y N N - DCD_06.02.02-25 4 2 354 06.02.02-26 2009/7/7 N N N - - N/A N/A 354 06.02.02-27 2009/7/7 Y N N - DCD_06.02.02-27 4 2 354 06.02.02-28 2009/7/7 N N N - - N/A N/A 354 06.02.02-29 2009/7/7 N N N - - N/A N/A 354 06.02.02-30 2009/7/7 N N N - - N/A N/A 1 of 4

Chapter:6 SRP Section DCD RAI Response DCD Change ID Number for Tracking DCD Impact Impact Impact Other Drivers DCD forthcoming RAI Question Response Response Report Revision No. Title on on on Revision No. No. Date Status Revision DCD COLA PRA 2009/7/7 Y Y N - DCD_06.02.02-31 4 2 354 06.02.02-31 10/06/2009 Y Y N - DCD_06.02.02-31 - 2 2009/7/7 Y Y N - DCD_06.02.02-32 - 2 354 06.02.02-32 10/06/2009 Y N N - DCD_06.02.02-32 - 2 2009/7/7 Y Y N - DCD_06.02.02-33 - 2 354 06.02.02-33 10/06/2009 Y N N - DCD_06.02.02-33 - 2 2009/7/7 Y Y N - DCD_06.02.02-34 - 2 354 06.02.02-34 10/06/2009 Y N N - DCD_06.02.02-34 - 2 2009/7/7 Y Y N - DCD_06.02.02-35 - 2 354 06.02.02-35 10/06/2009 Y N N - DCD_06.02.02-35 - 2 2009/7/7 Y Y N - DCD_06.02.02-36 - 2 354 06.02.02-36 10/06/2009 Y N N - DCD_06.02.02-36 - 2 354 06.02.02-37 2009/7/7 N N N - - N/A N/A 354 06.02.02-38 2009/7/7 Y N N - DCD_06.02.02-38 - 2 354 06.02.02-39 2009/7/7 N N N - - N/A N/A 354 06.02.02-40 2009/7/7 Y N N - DCD_06.02.02-40 4 2 354 06.02.02-41 2009/7/7 Y N N - DCD_06.02.02-41 - 2 354 06.02.02-42 2009/7/7 Y N N - DCD_06.02.02-42 4 2 354 06.02.02-43 2009/7/7 Y N N - DCD_06.02.02-43 4 2 354 06.02.02-44 2009/7/17 Y N N - DCD_06.02.02-44 TBD 366 06.02.02-45 2009/6/11 N N N - - N/A N/A 366 06.02.02-46 2009/6/11 Y N N - DCD_06.02.02-46 3 2 366 06.02.02-47 2009/6/11 N N N - - N/A N/A 366 06.02.02-48 2009/6/11 N N N - - N/A N/A 366 06.02.02-49 2009/6/11 N N N - - N/A N/A 366 06.02.02-50 2009/6/11 N N N - - N/A N/A 366 06.02.02-51 2009/6/11 N N N - - N/A N/A 422 06.02.02-52 2010/1/21 N N N - - N/A N/A 466 06.02.02-53 2009/11/24 N N N - - N/A N/A 466 06.02.02-54 2009/11/24 N N N - - N/A N/A 466 06.02.02-55 2009/11/24 Y N N - DCD_06.02.02-55 TBD 631 06.02.02-56 2010/10/21 N N N - N/A N/A 631 06.02.02-57 2010/10/21 N N N - N/A N/A 637 06.02.02-58 2010/10/21 N N N - - N/A N/A 637 06.02.02-59 2010/10/21 N N N - - N/A N/A 637 06.02.02-60 2010/10/21 Y N N TBD 645 06.02.02-61 2010/11/10 N N N - - N/A N/A

- - - - - - - COL 6.2(9) deleted MAP-06-007 4 2 6.2.4 Containment Isolation System 553 06.02.04-53 2010/4/19 Y N N - DCD_06.02.04-53 3 3 553 06.02.04-54 2010/4/19 Y N N - DCD_06.02.04-54 3 3 6.2.5 Combustible Gas Control 471 6.2.5-35 11/6/2009 Y N N - DCD_6.2.5-35 0 3 in Containment 471 6.2.5-36 2010/5/28 N N N - - N/A N/A 551 6.2.5-37 2010/4/20 Y N N - DCD_6.2.5-37 TBD 551 6.2.5-38 2010/4/20 N N N - - N/A N/A

- - - - - - - COL 6.2(7) deleted MAP-06-009 1 2 635 6.2.5-39 2010/10/20 Y N N - DCD_6.2.5-39 TBD 635 6.2.5-40 2010/10/20 Y N N - DCD_6.2.5-40 TBD 6.2.6 Containment Leakage Testing 472 06.02.06-23 2009/11/13 Y N N - DCD_06.02.06-23 1 3 472 06.02.06-24 2009/11/13 Y N N - DCD_06.02.06-24 1 3 472 06.02.06-25 2009/11/13 N N N - - N/A N/A 472 06.02.06-26 2009/11/27 Y N N - DCD_06.02.06-26 1 3 472 06.02.06-27 2009/11/27 Y N N - DCD_06.02.06-27 1 3 552 06.02.06-28 2010/4/16 N N N - - N/A N/A 552 06.02.06-29 2010/4/16 N N N - - N/A N/A 552 06.02.06-30 2010/4/16 Y N N - DCD_06.02.06-30 3 3

- - - - - - - COL 6.2(8) revised MAP-06-010 - 2 648 06.02.06-31 2010/11/11 Y N N - DCD_06.02.06-31 TBD 648 06.02.06-32 2010/11/11 Y N N - DCD_06.02.06-32 TBD 648 06.02.06-33 2010/11/11 Y N N - DCD_06.02.06-33 TBD 6.2.7 Fracture Prevention of 2 of 4

Chapter:6 SRP Section DCD RAI Response DCD Change ID Number for Tracking DCD Impact Impact Impact Other Drivers DCD forthcoming RAI Question Response Response Report Revision No. Title on on on Revision No. No. Date Status Revision DCD COLA PRA Containment Pressure Boundary 6.3 Emergency Core Cooling System 597 06.03-84 2010/7/8 N N N - - N/A N/A 597 06.03-85 2010/7/8 N N N - - N/A N/A 597 06.03-86 2010/7/8 N N N - - N/A N/A 626 06.03-87 2010/10/14 N N N - - N/A N/A 6.4 Control Room Habitability System 49 2008/9/16 Y N N fin. - DCD_06.04-9 (1) 2 06.04-9 473 11/13/2009 Y N N - DCD_06.04-9 0 3 501 06.04-10 2010/1/21 N N N - - N/A N/A 559 06.04-11 2010/5/20 Y N N - DCD_06.04-11 4 3 559 06.04-12 2010/5/20 Y N N - DCD_06.04-12 4 3 559 06.04-13 2010/5/20 Y N N - DCD_06.04-13 4 3 6.5.1 ESF Atmosphere Cleanup Systems 558 06.05.01-9 2010/5/27 Y N N - DCD_06.05.01-9 4 3 558 06.05.01-10 2010/4/22 N N N - - N/A N/A 558 06.05.01-11 2010/4/22 Y N N - DCD_06.05.01-11 3 3 558 06.05.01-12 2010/4/22 Y N N - DCD_06.05.01-12 3 3 558 06.05.01-13 2010/4/22 Y N N - DCD_06.05.01-13 3 3 558 06.05.01-14 2010/4/22 Y N N - DCD_06.05.01-14 3 3 558 06.05.01-15 2010/4/22 Y N N - DCD_06.05.01-15 3 3 558 06.05.01-16 2010/5/27 Y N N - DCD_06.05.01-16 4 3 558 06.05.01-17 2010/5/27 Y N N - DCD_06.05.01-17 4 3 558 06.05.01-18 2010/4/22 Y N N - DCD_06.05.01-18 3 3 615 06.05.01-19 2010/9/29 Y N N - DCD_06.05.01-19 5 3 615 06.05.01-20 2010/9/29 Y N N - DCD_06.05.01-20 5 3 6.5.2 Containment Spray 460 06.05.02-7 11/13/2009 N N N - - N/A N/A as a Fission Product 517 06.05.02-8 2010/2/25 N N N - - N/A N/A Cleanup System 6.5.3 Fission Product Control Systems and Structures 6.5.5 Pressure Suppression Pool as a Fission Product Cleanup System 6.6 Inservice Inspection and Testing of Class 2 and 3 Components 6.6.2 Inservice Inspection and Testing of Class 2 and 3 Components 6.6.3 Inservice Inspection and Testing of Class 2 and 3 Components 6.6.4 Inservice Inspection and Testing of Class 2 and 3 Components 3 of 4

Chapter:6 SRP Section DCD RAI Response DCD Change ID Number for Tracking DCD Impact Impact Impact Other Drivers DCD forthcoming RAI Question Response Response Report Revision No. Title on on on Revision No. No. Date Status Revision DCD COLA PRA 4 of 4

Chapter:9 SRP Section DCD RAI Response DCD Change ID Number for Tracking DCD Impact Impact Impact Other Drivers DCD forthcoming RAI Question Response Response Report Revision No. Title on on on Revision No. No. Date Status Revision DCD COLA PRA 9.1.1 Criticality Safety of Fresh and 155 09.01.01.-1 2009/2/10 N N N - - N/A N/A Spent Fuel Storage and Handling 155 09.01.01-2 2009/2/10 N N N - - N/A N/A 155 09.01.01-3 2009/2/10 N N N - - N/A N/A 155 09.01.01-4 2009/2/10 N N N - - N/A N/A 155 09.01.01-5 2009/2/10 N N N - - N/A N/A 155 09.01.01-6 2009/2/10 N N N - - N/A N/A 155 09.01.01-7 2009/2/10 N N N - - N/A N/A 155 09.01.01-8 2009/2/10 N N N - - N/A N/A 247 09.01.01-9 2009/3/30 Y N N - DCD_09.01.01-9 - 2 247 09.01.01-10 2009/3/30 Y Y N - DCD_09.01.01-10 - 2 382 09.01.01-11 2009/7/7 N N N - - N/A N/A 382 09.01.01-12 2009/7/7 N N N - - N/A N/A 382 09.01.01-13 2009/7/7 N N N - - N/A N/A 382 09.01.01-14 2009/7/7 N N N - - N/A N/A 382 09.01.01-15 2009/7/7 N N N - - N/A N/A 382 09.01.01-16 2009/7/7 N N N - - N/A N/A 382 09.01.01-17 2009/7/7 N N N - - N/A N/A 382 09.01.01-18 2009/7/7 N N N - - N/A N/A 382 09.01.01-19 2009/7/7 N N N - - N/A N/A 382 09.01.01-20 2009/7/7 N N N - - N/A N/A 382 09.01.01-21 2009/7/7 N N N - - N/A N/A 647 09.01.01-22 2010/11/11 N N N - - N/A N/A 647 09.01.01-23 2010/11/11 N N N - - N/A N/A 9.1.1 Criticality Safety of Fresh and Spent Fuel Storage and Handling 9.1.2 New and Spent Fuel Storage 9.1.3 Spent Fuel Pool Cooling and Cleanup System 9.1.4 Light Load Handling System 507 09.01.04-16 2010/2/15 Y N N - DCD_09.01.04-16 2 3 (Related to Refueling) 2010/6/4 Y N N -

555 09.01.04-17 2010/6/16 Y N N - DCD_09.01.04-17 4 3 2010/6/4 Y N N -

555 09.01.04-18 2010/6/16 Y N N - DCD_09.01.04-18 4 3 2010/6/4 Y N N -

555 09.01.04-19 2010/6/16 Y N N - DCD_09.01.04-19 4 3 2010/6/4 Y N N -

555 09.01.04-20 2010/6/16 Y N N - DCD_09.01.04-20 4 3 633 09.01.04-21 2010/10/21 Y N N - DCD_09.01.04-21 5 3 9.1.5 Overhead Heavy Load 563 09.01.05-14 2010/6/15 Y N N - DCD_09.01.05-14 4 3 Handling Systems 563 09.01.05-15 2010/6/15 Y N N - DCD_09.01.05-15 4 3 563 09.01.05-16 2010/6/15 Y N N - DCD_09.01.05-16 4 3 563 09.01.05-17 2010/6/15 Y N N - DCD_09.01.05-17 4 3 616 09.01.05-18 2010/9/22 Y N N - DCD_09.01.05-18 5 3 616 09.01.05-19 2010/9/22 N N N - - N/A N/A 9.2.1 Station Service Water System 585 09.02.01-32 2010/9/24 Y Y N - DCD_09.02.01-32 TBD 585 09.02.01-33 2010/9/24 Y Y N - DCD_09.02.01-33 TBD 585 09.02.01-34 2010/9/24 Y Y N - DCD_09.02.01-34 TBD 585 09.02.01-35 2010/9/24 Y Y N - DCD_09.02.01-35 TBD 585 09.02.01-36 2010/9/24 Y Y N - DCD_09.02.01-36 TBD 585 09.02.01-37 2010/9/24 Y Y N - DCD_09.02.01-37 TBD 585 09.02.01-38 2010/9/24 Y Y N - DCD_09.02.01-38 TBD 585 09.02.01-39 2010/9/24 Y Y N - DCD_09.02.01-39 TBD 585 09.02.01-40 2010/9/24 Y N N - DCD_09.02.01-40 TBD 585 09.02.01-41 2010/9/24 Y Y N - DCD_09.02.01-41 TBD 585 09.02.01-42 2010/9/24 Y N N - DCD_09.02.01-42 TBD 585 09.02.01-43 2010/9/24 Y Y N - DCD_09.02.01-43 TBD 585 09.02.01-44 2010/9/24 Y N N - DCD_09.02.01-44 TBD 585 09.02.01-45 2010/9/24 Y Y N - DCD_09.02.01-45 TBD 1 of 5

Chapter:9 SRP Section DCD RAI Response DCD Change ID Number for Tracking DCD Impact Impact Impact Other Drivers DCD forthcoming RAI Question Response Response Report Revision No. Title on on on Revision No. No. Date Status Revision DCD COLA PRA 585 09.02.01-46 2010/9/24 Y Y N - DCD_09.02.01-46 TBD 585 09.02.01-47 2010/9/24 N N N - - N/A N/A 585 09.02.01-48 2010/9/24 N N N - - N/A N/A 585 09.02.01-49 2010/9/24 Y Y N - DCD_09.02.01-49 TBD 585 09.02.01-50 2010/9/24 N N N - - N/A 585 09.02.01-51 2010/9/24 Y Y N - DCD_09.02.01-51 TBD 585 09.02.01-52 2010/9/24 Y Y N - DCD_09.02.01-52 TBD 585 09.02.01-53 2010/9/24 N N N - - N/A N/A 585 09.02.01-54 2010/9/24 N N N - - N/A N/A 585 09.02.01-55 2010/9/24 Y Y N - DCD_09.02.01-55 TBD 585 09.02.01-56 2010/9/24 N N N - - N/A N/A 585 09.02.01-57 2010/9/24 Y Y N - DCD_09.02.01-57 TBD 585 09.02.01-58 2010/9/24 Y Y N - DCD_09.02.01-58 TBD 585 09.02.01-59 2010/9/24 Y Y N - DCD_09.02.01-59 TBD 9.2.2 Reactor Auxiliary 567 09.02.02-46 2010/5/7 Y N N - DCD_09.02.02-46 3 3 Cooling Water Systems 571 09.02.02-47 2010/6/8 N N N - - N/A N/A 571 09.02.02-48 2010/6/8 Y N N - DCD_09.02.02-48 TBD 571 09.02.02-49 2010/6/8 N N N - - N/A N/A 571 09.02.02-50 2010/6/8 Y N N - DCD_09.02.02-50 TBD 571 09.02.02-51 2010/6/8 Y N N - DCD_09.02.02-51 TBD 571 09.02.02-52 2010/6/8 Y N N - DCD_09.02.02-52 TBD 571 09.02.02-53 2010/6/8 Y N N - DCD_09.02.02-53 TBD 571 09.02.02-54 2010/6/8 N N N - - N/A N/A 571 09.02.02-55 2010/6/8 Y N N - DCD_09.02.02-55 TBD 571 09.02.02-56 2010/6/8 Y N N - DCD_09.02.02-56 TBD 571 09.02.02-57 2010/6/8 Y N N - DCD_09.02.02-57 TBD 571 09.02.02-58 2010/6/8 Y N N - DCD_09.02.02-58 TBD 571 09.02.02-59 2010/6/8 Y N N - DCD_09.02.02-59 TBD 571 09.02.02-60 2010/6/8 N N N - - N/A N/A 571 09.02.02-61 2010/6/8 N N N - - N/A N/A 571 09.02.02-62 2010/6/8 Y N N - DCD_09.02.02-62 TBD 571 09.02.02-63 2010/6/8 Y N N - DCD_09.02.02-63 TBD 571 09.02.02-64 2010/6/8 Y N N - DCD_09.02.02-64 TBD 571 09.02.02-65 2010/6/8 Y N N - DCD_09.02.02-65 TBD 571 09.02.02-66 2010/6/8 Y N N - DCD_09.02.02-66 TBD 571 09.02.02-67 2010/6/8 Y N N - DCD_09.02.02-67 TBD 571 09.02.02-68 2010/6/8 Y N N - DCD_09.02.02-68 TBD 576 09.02.02-69 2010/6/8 N N N - - N/A N/A 584 09.02.02-70 2010/6/10 Y N N - DCD_09.02.02-70 TBD 584 09.02.02-71 2010/6/10 Y N N - DCD_09.02.02-71 TBD 584 09.02.02-72 2010/6/10 Y N N - DCD_09.02.02-72 TBD 584 09.02.02-73 2010/6/10 N N N - - N/A N/A 584 09.02.02-74 2010/6/10 N N N - - N/A N/A 584 09.02.02-75 2010/6/10 N N N - - N/A N/A 584 09.02.02-76 2010/6/10 Y N N - DCD_09.02.02-76 TBD 584 09.02.02-77 2010/6/10 Y N N - DCD_09.02.02-77 TBD 584 09.02.02-78 2010/6/10 Y N N - DCD_09.02.02-78 TBD 584 09.02.02-79 2010/6/10 N N N - - N/A N/A 2 of 5

Chapter:9 SRP Section DCD RAI Response DCD Change ID Number for Tracking DCD Impact Impact Impact Other Drivers DCD forthcoming RAI Question Response Response Report Revision No. Title on on on Revision No. No. Date Status Revision DCD COLA PRA 9.2.4 Potable and Sanitary Water Systems 9.2.5 Ultimate Heat Sink 2009/5/12 N N N - - N/A N/A 286 09.02.05-1 2010/7/7 Y N N - DCD_09.02.05-1 4 3 2009/5/12 N N N - - N/A N/A 286 09.02.05-2 2010/7/7 Y N N - DCD_09.02.05-2 4 3 2009/5/12 N N N - - N/A N/A 286 09.02.05-3 2010/7/7 Y N N - DCD_09.02.05-3 4 3 2009/5/12 N N N - - N/A N/A 286 09.02.05-4 2010/7/7 Y N N - DCD_09.02.05-4 4 3 2009/5/12 N N N - - N/A N/A 286 09.02.05-5 2010/7/7 N N N - - N/A N/A 2009/5/12 N N N - - N/A N/A 286 09.02.05-6 2010/7/7 Y N N - DCD_09.02.05-6 4 3 2009/5/12 N N N - - N/A N/A 286 09.02.05-7 2010/7/7 Y N N - DCD_09.02.05-7 4 3 2009/5/12 N N N - - N/A N/A 286 09.02.05-8 2010/7/7 Y N N - DCD_09.02.05-8 4 3 2009/5/12 N N N - - N/A N/A 286 09.02.05-9 2010/7/7 Y N N - DCD_09.02.05-9 4 3 9.2.6 Condensate Storage Facilities 9.3.1 Compressed Air System 9.3.2 Process and Post-accident 461 09.03.02-12 2009/11/17 Y N N - DCD_09.03.02-12 1 3 Sampling Systems 526 09.03.02-13 2010/4/7 Y Y N - DCD_09.03.02-13 3 3 526 09.03.02-14 2010/4/7 N N N - - N/A N/A 526 09.03.02-15 2010/4/7 Y N N - DCD_09.03.02-15 3 3 526 09.03.02-16 2010/4/7 N N N - - N/A N/A 9.3.3 Equipment and Floor 426 09.03.03-15 2009/9/14 Y N N - DCD_09.03.03-15 - 2 Drainage System 426 09.03.03-16 2009/9/14 Y N N - DCD_09.03.03-16 0 3 426 09.03.03-17 2009/9/14 Y N N - DCD_09.03.03-17 - 2 591 09.03.03-18 2010/7/7 N N N - - N/A N/A 591 09.03.03-19 2010/7/7 Y N N - DCD_09.03.03-19 4 3 9.3.4 Chemical and Volume Control Syst 2009/7/17 N N N - - N/A N/A 384 09.03.04-10 (PWR) 2010/4/7 Y Y N - DCD_09.03.04-10 3 3 (Including Boron Recovery System) 9.4.1 Control Room Area 2008/10/3 N N N fin. -

63 09.04.01-14 Ventilation System 2010/6/29 N N N - - N/A N/A 2009/6/19 N N N -

327 09.04.01-5 2010/6/29 N N N - - N/A N/A 327 09.04.01-9 2010/1/29 Y N N - DCD_09.04.01-9 2 3 475 09.04.01-12A 2009/11/20 Y Y N - DCD_09.04.01-12A 1 3 475 09.04.01-13A 2009/11/20 Y N N - DCD_09.04.01-13A 1 3 475 09.04.01-14A 2009/11/20 N N N - - N/A N/A 484 09.04.01-15A 2009/12/9 N N N - - N/A N/A 582 09.04.01-16 2010/7/16 Y N N - DCD_09.04.01-16 4 3 582 09.04.01-17 2010/7/16 Y N N - DCD_09.04.01-17 4 3 582 09.04.01-18 2010/7/16 N N N - - N/A N/A 582 09.04.01-19 2010/7/16 N N N - - N/A N/A 582 09.04.01-20 2010/7/16 Y N N - DCD_09.04.01-20 4 3 582 09.04.01-21 2010/7/16 Y N N - DCD_09.04.01-21 4 3 582 09.04.01-22 2010/7/16 Y N N - DCD_09.04.01-22 4 3 582 09.04.01-23 2010/7/16 N N N - - N/A N/A 642 09.04.01-24 2010/11/5 Y N N - DCD_09.04.01-24 5 3 9.4.2 Spent Fuel Pool Area 539 09.04.02-4 2010/4/1 Y N N - DCD_09.04.02-4 3 3 3 of 5

Chapter:9 SRP Section DCD RAI Response DCD Change ID Number for Tracking DCD Impact Impact Impact Other Drivers DCD forthcoming RAI Question Response Response Report Revision No. Title on on on Revision No. No. Date Status Revision DCD COLA PRA Ventilation System 539 09.04.02-5 2010/4/1 Y N N - DCD_09.04.02-5 3 3 592 09.04.02-6 2010/7/7 Y N N - DCD_09.04.02-6 4 3 9.4.3 Auxiliary and Radwaste Area 483 09.04.03-08 2010/2/5 Y N N - DCD_09.04.03-08 2 3 Ventilation System 483 09.04.03-09 2010/2/5 Y N N - DCD_09.04.03-09 2 3 483 09.04.03-10 2010/2/5 N N N - - N/A N/A 634 09.04.03-11 2010/10/15 Y N N - DCD_09.04.03-11 5 3 634 09.04.03-12 2010/10/15 Y N N - DCD_09.04.03-12 5 3 634 09.04.03-13 2010/10/15 Y N N - DCD_09.04.03-13 5 3 9.4.4 Turbine Area Ventilation System 541 09.04.03-4 2010/3/30 N N N - - N/A N/A 541 09.04.04-5 2010/3/30 Y N N - DCD_09.04.04-5 3 3 586 09.04.04-6 2010/6/10 N N N - - N/A N/A 9.4.5 Engineered Safety Feature 2008/10/6 N N N fin. -

64 09.04.05-1/9.4.5-3 Ventilation System 2010/6/29 N N N - - N/A N/A 2008/10/6 N N N fin. -

64 09.04.05-1/9.4.5-4 2010/6/29 N N N - - N/A N/A 2008/10/6 N N N fin. -

64 09.04.05-1/9.4.5-22 2010/6/29 N N N - - N/A N/A 2009/7/17 N N N -

356 09.04.05-3 2010/6/29 N N N - - N/A N/A 2009/7/17 N N N -

356 09.04.05-4 2010/6/29 N N N - - N/A N/A 2009/7/17 N N N -

356 09.04.05-9 2010/6/29 N N N - - N/A N/A 474 09.04.05-10 11/13/2009 Y N N - DCD_09.04.05-10 0 3 583 09.04.05-11 2010/6/22 Y N N - DCD_09.04.05-11 4 3 583 09.04.05-12 2010/6/22 Y N N - DCD_09.04.05-12 4 3 9.5.1 Fire Protection Program 537 09.05.01-18 04/13/2010 Y N N - DCD_09.05.01-18 3 3 537 09.05.01-19 04/13/2010 Y N N - DCD_09.05.01-19 3 3 9.5.2 Communications Systems 9.5.3 Lighting Systems 9.5.4 Emergency Diesel Engine Fuel 467 09.05.04-43 11/10/2009 Y Y N - DCD_09.05.04-43 1 3 Oil Storage and Transfer System 468 09.05.04-44 2009/12/10 Y Y N - DCD_09.05.04-44 1 3 468 09.05.04-45 2009/12/10 Y N N - DCD_09.05.04-45 1 3 468 09.05.04-46 2009/12/10 Y N N - DCD_09.05.04-46 1 3 468 09.05.04-47 2009/12/10 Y N N - DCD_09.05.04-47 1 3 468 09.05.04-48 2009/12/10 Y N N - DCD_09.05.04-48 1 3 468 09.05.04-49 2009/12/10 N N N - - N/A N/A 565 09.05.04-50 2010/6/15 Y N N - DCD_09.05.04-50 4 3 565 09.05.04-51 2010/6/15 Y N N - DCD_09.05.04-51 4 3 9.5.5 Emergency Diesel Engine Cooling Water System 9.5.6 Emergency Diesel Engine Starting System 504 09.05.06-24 12/23/09 Y N N - DCD_09.05.06-24 1 3 504 09.05.06-25 12/23/09 Y N N - DCD_09.05.06-25 1 3 9.5.7 Emergency Diesel Engine 469 09.05.07-18 11/6/2009 N N N - - N/A N/A Lubrication System 469 09.05.07-19 11/6/2009 N N N - - N/A N/A 506 09.05.07-20 2010/1/29 Y N N - DCD_09.05.07-20 2 3 506 09.05.07-21 2010/1/29 N N N - - N/A N/A 506 09.05.07-22 2010/1/29 Y N N - DCD_09.05.07-22 2 3 4 of 5

Chapter:9 SRP Section DCD RAI Response DCD Change ID Number for Tracking DCD Impact Impact Impact Other Drivers DCD forthcoming RAI Question Response Response Report Revision No. Title on on on Revision No. No. Date Status Revision DCD COLA PRA 506 09.05.07-23 2010/1/29 Y N N - DCD_09.05.07-23 2 3 556 09.05.07-24 2010/4/27 Y N N - DCD_09.05.07-24 3 3 9.5.8 Emergency Diesel Engine 470 09.05.08-18 2009/12/2 Y N N - DCD_09.05.08-18 1 3 Combustion Air Intake and 470 09.05.08-19 2009/12/2 N N N - - N/A N/A Exhaust System 470 09.05.08-20 2009/12/2 Y N N - DCD_09.05.08-20 1 3 470 09.05.08-21 2009/12/2 Y N N - DCD_09.05.08-21 1 3 470 09.05.08-22 2009/12/2 Y N N - DCD_09.05.08-22 1 3 505 09.05.08-23 2010/2/1 N N N - - N/A N/A 505 09.05.08-24 2010/2/1 N N N - - N/A N/A 505 09.05.08-25 2010/2/1 Y N N - DCD_09.05.08-25 2 3 557 09.05.08-26 2010/6/14 Y N N - DCD_09.05.08-26 5 3 618 09.05.08-27 2010/11/4 Y N N - DCD_09.05.08-27 5 3 5 of 5

Chapter:10 SRP Section DCD RAI Response DCD Change ID Number for Tracking DCD Impact Impact Impact Other Drivers DCD forthcoming RAI Question Response Response Report Revision No. Title on on on Revision No. No. Date Status Revision DCD COLA PRA 10.2 Turbine Generator 598 10.02-3 2010/7/20 N N N - - N/A N/A 598 10.02-4 2010/7/20 Y N N - DCD_10.2-4 4 3 10.2.3 Turbine Rotor Integrity 574 10.02.03-8 2010/6/10 N N N - - N/A N/A 574 10.02.03-9 2010/6/10 N N N - - N/A N/A 574 10.02.03-10 2010/6/10 Y N N - DCD_10.02.03-10 4 3 574 10.02.03-11 2010/6/10 N N N - - N/A N/A 10.3 Main Steam Supply System 10.3.6 Steam and 10.03.06-1 Feedwater System Materials 10.03.06-2 10.03.06-3 10.03.06-4 10.03.06-5 10.03.06-6 10.03.06-7 10.03.06-8 10.03.06-9 500 10.03.06-10 12/24/2009 Y N N - DCD_10.03.06-10 1 3 500 10.03.06-11 12/24/2009 N N N - - N/A N/A 500 10.03.06-12 12/24/2009 Y N N - DCD_10.03.06-12 1 3 10.4.1 Main Condensers 10.4.2 Main Condenser Evacuation System 10.4.3 Turbine Gland Sealing System 10.4.4 Turbine Bypass System 10.4.5 Circulating Water System 10.4.6 Condensate Cleanup System 441 10.04.06-8 2009/9/16 Y N N - DCD_10.04.06-8 0 3 441 10.04.06-9 2009/9/16 N N N - - N/A N/A 441 10.04.06-10 2009/9/16 N N N - - N/A N/A 543 10.04.06-11/OI 10.04.06-1 2010/4/26 N N N - - N/A N/A 543 10.04.06-12/OI 10.04.06-2 2010/4/26 N N N - - N/A N/A 543 10.04.06-13/OI 10.04.06-3 2010/4/26 N N N - - N/A N/A 543 10.04.06-14/OI 10.04.06-4 2010/4/26 N N N - - N/A N/A 543 10.04.06-15/OI 10.04.06-5 2010/4/26 N N N - - N/A N/A 630 10.04.06-16 2010/10/6 Y N N - DCD_10.04.06-16 5 3 10.4.7 Condensate and Feedwater System 10.4.8 Steam Generator Blowdown System (PWR) 10.4.9 Auxiliary Feedwater System (PWR) 1 of 1

Chapter:11 SRP Section DCD RAI Response DCD Change ID Number for Tracking DCD Impact Impact Impact Other Drivers DCD forthcoming RAI Question Response Response Report Revision No. Title on on on Revision No. No. Date Status Revision DCD COLA PRA 11.1 Source Terms 11.2 Liquid Waste Management System 458 11.02-21 2009/10/26 N N N - - N/A N/A 462 11.02-22 2009/11/17 Y N N - DCD_11.02-22 1 3 462 11.02-23 2009/11/17 N N N - - N/A N/A 462 11.02-24 2009/11/17 N N N - - N/A N/A 462 11.02-25 2009/11/17 Y N N - DCD_11.02-25 0 3 462 11.02-26 2009/11/17 N N N - - N/A N/A 462 11.02-27 2009/11/17 Y N N - DCD_11.02-27 0 3 523 11.02-28 2010/3/15 Y N N - DCD_11.02-28 3 3 523 11.02-29 2010/3/15 Y Y N - DCD_11.02-29 3 3 523 11.02-30 2010/3/15 Y N N - DCD_11.02-30 3 3 523 11.02-31 2010/3/15 Y N N - DCD_11.02-31 3 3 523 11.02-32 2010/3/15 Y N N - DCD_11.02-32 3 3 624 11.02-33 2010/9/24 Y Y N - DCD_11.02-33 5 3 11.3 Gaseous Waste 533 11.03-15 2010/4/20 Y N N - DCD_11.03-15 3 3 Management System 535 11.03-16 2010/4/20 Y N N - DCD_11.03-16 3 3 535 11.03-17 2010/4/20 Y N N - DCD_11.03-17 3 3 629 11.03-18 2010/9/24 Y N N - DCD_11.03-18 5 3 11.4 Solid Waste Management System 534 11.04-19 2010/4/20 Y Y N - DCD_11.04-19 3 3 536 11.04-20 2010/4/20 Y N N - DCD_11.04-20 3 3 536 11.04-21 2010/4/20 Y N N - DCD_11.04-21 3 3 11.5 Process and Effluent 522 11.05-18 2010/3/8 Y N N - DCD_11.05-18 3 3 Radiological Monitoring Instrumentation and Sampling Systems 1 of 1

Chapter:12 SRP Section DCD RAI Response DCD Change ID Number for DCD Tracking Impact Impact Impact Respon Other Drivers DCD forthcoming Revisio RAI Question Response Report No. Title on on on se Revision n No. No. Date Revision DCD COLA PRA Status 12.1 Assuring that Occupational Radiation Exposures Are As Low As Is Reasonably Achievable 12.2 Radiation Sources 2009/9/28 Y N N - DCD_12.02-19 0 3 427 12.02-19 2010/9/14 Y N N - DCD_12.02-19 0 3 427 12.02-21 2009/9/28 Y N N - DCD_12.02-21 0 3 427 12.02-22 2009/9/28 Y N N - DCD_12.02-22 0 3 532 12.02-23 2010/4/9 Y N N - DCD_12.02-23 3 3 532 12.02-24 2010/4/9 N N N - - N/A N/A 532 12.02-25 2010/4/9 Y N N - DCD_12.02-25 3 3 532 12.02-26 2010/4/9 Y N N - DCD_12.02-26 3 3 2010/4/9 Y N N - 3 3 532 12.02-27 2010/9/14 Y N N - DCD_12.02-27 5 3 532 12.02-28 2010/4/9 Y N N - DCD_12.02-28 TBD 3 2010/4/9 Y Y N - 3 3 532 12.02-29 2010/4/9 Y Y N - DCD_12.02-29 5 3 532 12.02-30 2010/4/9 Y Y N - DCD_12.02-30 3 3 561 12.02-31 2010/4/9 N N N - - N/A N/A 12.3- Radiation Protection 425 12.03-12.04-21 2009/9/4 Y N N - DCD_12.03-12.04-21 0 3 12.4 Design Features 428 12.03-12.04-22 2009/9/28 N N N - - N/A N/A 428 12.03-12.04-23 2009/9/28 N N N - - N/A N/A 428 12.03-12.04-24 2009/9/28 N N N - - N/A N/A 429 12.03-12.04-25 2009/9/28 Y N N - DCD_12.03-12.04-25 0 3 429 12.03-12.04-26 2009/9/28 Y Y N - DCD_12.03-12.04-26 0 3 429 12.03-12.04-27 2009/9/28 Y N N - DCD_12.03-12.04-27 0 3 429 12.03-12.04-28 2009/9/28 N N N - - N/A N/A 429 12.03-12.04-29 2009/9/28 N N N - - N/A N/A 429 12.03-12.04-30 2009/9/28 Y N N - DCD_12.03-12.04-30 0 3 429 12.03-12.04-31 2009/9/28 Y N N - DCD_12.03-12.04-31 0 3 453 12.03-12.04-32 2009/9/16 N N N - - N/A N/A 2010/3/12 N N N -

524 12.03-12.04-33 2010/9/14 Y Y N - DCD_12.03-12.04-33 5 3 2010/3/12 N N N - - N/A N/A 524 12.03-12.04-34 2010/10/8 Y Y N - DCD_12.03-12.04-34 5 3 2010/3/12 N N N -

524 12.03-12.04-35 2010/9/14 Y N N - DCD_12.03-12.04-35 5 3 2010/3/12 Y N N -

524 12.03-12.04-36 2010/9/14 Y N N - DCD_12.03-12.04-36 5 3 2010/7/30 Y Y N -

578 12.03-12.04-37 2010/8/9 Y Y N - DCD_12.03-12.04-37 4 3 2010/7/30 Y N N -

578 12.03-12.04-38 2010/8/9 Y N N - DCD_12.03-12.04-38 4 3 2010/7/30 Y N N -

578 12.03-12.04-39 2010/8/9 Y N N - DCD_12.03-12.04-39 4 3 12.5 Operational Radiation Protection Program 1 of 1

Chapter:19 SRP Section DCD RAI Response DCD Change ID Number for Tracking DCD Impact Impact Impact Other Drivers DCD forthcoming RAI Question Response Response Report Revision No. Title on on on Revision No. No. Date Status Revision DCD COLA PRA 19 Probabilistic Risk Assessment 443 19-391 2009/10/1 N N N - - N/A N/A and Severe Accident Evaluation 443 19-392 2009/10/1 N N N - - N/A N/A for New Reactors 443 19-393 2009/10/1 Y N N - DCD_19-393 - 2 443 19-394 2009/10/1 N N N - - N/A N/A 443 19-395 2009/10/1 N N N - - N/A N/A 443 19-396 2009/10/1 Y N N - DCD_19-396 0 3 443 19-397 2009/10/1 Y N N - DCD_19-397 0 3 454 19-398 2009/10/9 N N Y - - N/A N/A 454 19-399 2009/10/9 N N Y - - N/A N/A 454 19-400 2009/10/9 N N Y - - N/A N/A 454 19-401 2009/10/9 Y N Y - DCD_19-401 - 2 479 19-402 2009/11/25 Y N N - DCD_19-402 1 3 479 19-403 2009/11/25 Y N N - DCD_19-403 1 3 479 19-404 2009/11/25 Y N N - DCD_19-404 1 3 479 19-405 2009/11/25 N N N - - N/A N/A 479 19-406 2009/11/25 N N N - - N/A N/A 480 19-*** (1) 2009/11/26 N N N - - N/A N/A 480 19-*** (2) 2009/11/26 N N N - - N/A N/A 480 19-*** (3) 2009/11/26 N N N - - N/A N/A 480 19-*** (4) 2009/11/26 N N N - - N/A N/A 480 19-*** (5) 2009/11/26 N N N - - N/A N/A 480 19-*** (6) 2009/11/26 N N N - - N/A N/A 528 19-407 2010/3/3 Y N N - DCD_19-407 2 3 528 19-408 2010/3/3 Y N N - DCD_19-408 2 3 528 19-409 2010/3/3 Y N N - DCD_19-409 2 3 528 19-410 2010/3/3 Y N N - DCD_19-410 2 3 528 19-411 2010/3/3 N N N - - N/A N/A 528 19-412 2010/3/3 Y N N - DCD_19-412 3 3 528 19-413 2010/3/3 Y N N - DCD_19-413 3 3 528 19-414 2010/3/3 Y N N - DCD_19-414 2 3 528 19-415 2010/3/3 Y N N - DCD_19-415 2 3 528 19-416 2010/3/3 Y N N - DCD_19-416 2 3 528 19-417 2010/3/3 Y N N - DCD_19-417 2 3 528 19-418 2010/3/3 Y N N - DCD_19-418 2 3 528 19-419 2010/3/3 Y N N - DCD_19-419 2 3 528 19-420 2010/3/3 Y N N - DCD_19-420 2 3 528 19-421 2010/3/3 Y N N - DCD_19-421 3 3 528 19-422 2010/3/3 Y N N - DCD_19-422 2 3 564 19-423 2010/4/28 Y N N - DCD_19-423 3 3 564 19-424 2010/4/28 Y N N - DCD_19-424 3 3 564 19-425 2010/4/28 Y N N - DCD_19-425 3 3 564 19-426 2010/4/28 Y Y N - DCD_19-426 3 3 564 19-427 2010/4/28 Y N N - DCD_19-427 3 3 564 19-428 2010/4/28 Y N N - DCD_19-428 3 3 566 19-429 2010/4/28 Y N N - DCD_19-429 3 3 19-430 19-431 19-432 19-433 19-434 19-435 601 19-436 2010/7/26 Y N N - DCD_19-436 4 3 601 19-437 2010/7/26 Y N N - DCD_19-437 4 3 607 19-438 2010/9/3 Y N Y - DCD_19-438 5 3 608 19-439 2010/9/3 Y N N - DCD_19-439 TBD 608 19-440 2010/9/3 Y N N - DCD_19-440 5 3 609 19-441 2010/9/3 Y N N - DCD_19-441 TBD 610 19-442 2010/9/3 N N N - - N/A N/A 619 19-443 2010/9/10 N N N - - N/A N/A 622 19-444 2010/9/29 N N Y - - N/A N/A 622 19-445 2010/9/29 N N N - - N/A N/A 622 19-446 2010/9/29 N N Y - - N/A N/A 627 19-447 2010/11/1 Y N N - DCD_19-447 TBD 627 19-448 2010/11/1 N N N - - N/A N/A 627 19-449 2010/11/1 Y N Y - DCD_19-449 TBD 627 19-450 2010/11/1 Y N Y - DCD_19-450 TBD 627 19-451 2010/11/1 N N N - - N/A N/A 627 19-452 2010/11/1 N N N - - N/A N/A 627 19-453 2010/11/1 N N N - - N/A N/A 627 19-454 11/1/2010 Y N N - DCD_19-454 5 3 639 19-455 10/29/2010 Y N N - DCD_19-455 5 3 639 19-456 10/29/2010 Y N N - DCD_19-456 5 3 1 of 3

Chapter:19 SRP Section DCD RAI Response DCD Change ID Number for Tracking DCD Impact Impact Impact Other Drivers DCD forthcoming RAI Question Response Response Report Revision No. Title on on on Revision No. No. Date Status Revision DCD COLA PRA 639 19-457 10/29/2010 Y N N - DCD_19-457 5 3 639 19-458 10/29/2010 Y N Y - DCD_19-458 5 3 639 19-459 10/29/2010 N N N - - N/A N/A 639 19-460 10/29/2010 N N N - - N/A N/A 639 19-461 10/29/2010 N N Y - - N/A N/A 639 19-462 10/29/2010 N N N - - N/A N/A 639 19-463 10/29/2010 N N N - - N/A N/A 639 19-464 10/29/2010 Y N Y - DCD_19-464 5 3 639 19-465 10/29/2010 N N N - - N/A N/A 640 19-466 10/29/2010 N N N - - N/A N/A 640 19-467 10/29/2010 Y N N - DCD_19-467 5 3 640 19-468 10/29/2010 N N N - - N/A N/A 640 19-469 10/29/2010 Y N N - DCD_19-469 5 3 640 19-470 10/29/2010 Y N N - DCD_19-470 5 3 640 19-471 10/29/2010 N N Y - - N/A N/A 640 19-472 10/29/2010 Y N Y - DCD_19-472 TBD 641 19-473 10/29/2010 N N N - - N/A N/A 641 19-474 10/29/2010 N N N - - N/A N/A 641 19-475 10/29/2010 N N N - - N/A N/A 641 19-476 10/29/2010 Y N Y - DCD_19-476 TBD 641 19-477 10/29/2010 N N Y - - N/A N/A 641 19-478 10/29/2010 Y N N - DCD_19-478 5 3 641 19-479 10/29/2010 Y N Y - DCD_19-479 TBD 641 19-480 10/29/2010 N N N - - N/A N/A 641 19-481 10/29/2010 N N N - - N/A N/A 641 19-482 10/29/2010 N N N - - N/A N/A 641 19-483 10/29/2010 Y N Y - DCD_19-483 TBD 649 19-484 2010/11/12 N N N - - N/A N/A 649 19-485 2010/11/12 Y N Y - DCD_19-485 TBD 649 19-486 2010/11/12 N N Y - - N/A N/A 649 19-487 2010/11/12 N N Y - - N/A N/A 649 19-488 2010/11/12 Y N N - DCD_19-488 TBD 649 19-489 2010/11/12 Y N Y - DCD_19-489 TBD 649 19-490 2010/11/12 Y N N - DCD_19-490 TBD 649 19-491 2010/11/12 Y N Y - DCD_19-491 TBD Determining the Technical 19.1 577 19.01-1 2010/5/26 Y N N - DCD_19.01-1 4 3 Adequacy of Probabilistic Risk Assessment 621 19.01-2 2010/9/29 Y N N - DCD_19.01-2 5 3 Results for Risk-Informed Activities 621 19.01-3 2010/9/29 Y N N - DCD_19.01-3 5 3 621 19.01-4 2010/9/29 Y N N - DCD_19.01-4 5 3 621 19.01-5 2010/9/29 Y N N - DCD_19.01-5 5 3 621 19.01-6 2010/9/29 Y N Y - DCD_19.01-6 5 3 621 19.01-7 2010/9/29 Y N Y - DCD_19.01-7 5 3

- - - - - - - COL 19.3(5) deleted MAP-19-001 - 2 628 19.01-8 2010/10/14 Y N N - DCD_19.01-8 5 3 19.2 Review of Risk Information Used to Support Permanent Plant -

Specific Changes to the Licensing Basis:

General Guidance 2 of 3

Chapter:19 SRP Section DCD RAI Response DCD Change ID Number for Tracking DCD Impact Impact Impact Other Drivers DCD forthcoming RAI Question Response Response Report Revision No. Title on on on Revision No. No. Date Status Revision DCD COLA PRA 3 of 3

US-APWR DCD Revision 2 RAI Tracking Report MUAP-09026(R5)

Chapter 2 Mitsubishi Heavy Industries, LTD.

US-APWR DCD Chapter 2 Rev. 2, Tracking Report Rev. 5 Change List Location (e.g., subsection Page with paragraph/ Description of Change sentence/ item, table with row/column, or figure) 2.4-1 Section 2.4 Add as second sentence: For the purposes of this DCD, st plant grade is synonymous with ground level, and 1 Paragraph these terms are used interchangeably.

1st Sentence Reason: Provided clarification. ACRS Questions for Chapter 2 and 16.

2.4-1 Subsection 2.4.1 Change: Major external hydrologic considerations for st safety operation of the plant include both surface and 1 Paragraph subsurface sources. to Major hydrologic considerations 1st Sentence for safe operation of the plant include both surface and subsurface sources.

Reason: Editorial correction 2.4-1 Subsection 2.4.1 Add as last sentence: The COL Applicant also needs to st provide adequate description of the interface of the plant 1 Paragraph with the Hydrosphere.

Last Sentence Reason: Provided clarification 2.4-4 Subsection 2.4.13 Add as new first sentence: Analyses performed by the COL Applicant to determine the impacts of accidental releases of liquid radioactive effluents are site-specific.

Reason: Provided clarification Page 1 of 1

2. SITE CHARACTERISTICS US-APWR Design Control Document 2.4 Hydrologic Engineering The US-APWR is designed for a maximum ground water elevation of 1 ft. below plant grade as well as a maximum level for flood or tsunami of 1 ft. below plant grade. For the purposes of this DCD, plant grade is synonymous with ground level, and these terms are used interchangeably. The US-APWR is designed for a maximum local intense precipitation of 19.4 in./hr. The local intense precipitation is a measure of the extreme amount of water falling in the immediate vicinity of the site, taken as the one-square-mile probable maximum precipitation (PMP). Table 2.0-1 contains standard plant design input for hydrology.

The COL Applicant is to provide sufficient information to verify that hydrologic-related events will not affect the safety-basis for the US-APWR.

Non safety-related structures and certain safety-related structures whose flooding would not prevent safe operation of the plant need not be designed for the effects of high water or ice. Examples of safety-related structures that may not be adversely affected by flooding or icing include water intake structures and ultimate heat sink basins.

2.4.1 Hydrologic Description Major external hydrologic considerations for safe operation of the plant include both surface and subsurface sources. The hydrologic description includes the location, size, shape, and other hydrologic characteristics of streams, lakes, shore regions, and ground water environments influencing plant siting, and includes a description of existing and proposed water control structures, both upstream and downstream, that may influence conditions at the site. The COL Applicant also needs to provide adequate description of the interface of the plant with the Hydrosphere.

2.4.2 Floods The site-specific design of flood protection for safety-related components and structures of the plant is based on the highest calculated flood water level elevations and flood wave effects (site-characteristic flood) resulting from analyses of several different hypothetical causes. The site-specific design for local probable maximum precipitation demonstrates the capability of site drainage facilities to prevent flooding of safety-related facilities.

2.4.3 Probable Maximum Flood The site-specific probable maximum flood (PMF) is based on the nearby streams and rivers contribution to the design basis flooding. Any reservoir and channel routing assumptions are addressed for site-specific impact, including coefficients and their bases with appropriate discussion of initial conditions, outlet works (controlled and uncontrolled), and spillways (controlled and uncontrolled).

A site-specific flood analysis also includes the translation of the estimated peak PMP discharge to elevation using applicable site profile and precipitation data.

Tier 2 2.4-1 Revision 23

2. SITE CHARACTERISTICS US-APWR Design Control Document dewatering during construction is critical to the integrity of safety-related structures, the bases for subsurface hydrostatic loadings assumed during construction and the dewatering methods to be employed in achieving these loadings are to be discussed.

2.4.13 Accident Releases of Radioactive Liquid Effluent in Ground and Surface Waters Analyses performed by the COL Applicant to determine the impacts of accidental releases of liquid radioactive effluents are site-specific. The site is evaluated for the ability of the ground and surface water environment to delay, disperse, dilute, or concentrate liquid effluents, as related to existing or potential future water users. The bases used to determine dilution factors, dispersion coefficients, flow velocities, travel times, adsorption, and pathways of liquid contaminants is to be discussed, including references to the locations and users of surface waters, and the site-specific release points.

2.4.14 Technical Specification and Emergency Operation Requirements Any emergency protective measures designed to minimize the impact of adverse hydrology-related events on safety-related facilities are described on a site-specific basis.

Applicable site-specific information includes the manner in which to incorporate these requirements into appropriate TSs and emergency procedures, and the need for any TSs for plant shutdown to minimize the consequences of an accident resulting from hydrologic phenomena such as floods or the degradation of the UHS. The potential effects of seismic and non-seismic information on the postulated technical specifications and emergency operations is also to be evaluated for the proposed plant site. If emergency procedures are used to meet safety requirements associated with hydrologic events, the event is to be identified, and appropriate water levels and lead times available are to be provided.

2.4.15 Combined License Information COL 2.4(1) The COL Applicant is to provide sufficient information to verify that hydrologic-related events will not affect the safety-basis for the US-APWR.

2.4-16 References 2.4-1 Minimization of Contamination and Radioactive Waste Generation: Life-Cycle Planning. Regulatory Guide 4.21, Rev. 0, U.S. Nuclear Regulatory Commission, Washington, DC, June 2008.

Tier 2 2.4-4 Revision 23

US-APWR DCD Revision 2 RAI Tracking Report MUAP-09026(R5)

Chapter 6 Mitsubishi Heavy Industries, LTD.

US-APWR DCD Chapter 6 Rev. 2, Tracking Report Rev. 5 Change List Location (e.g., subsection with Page paragraph/sentence/ Description of Change Item, table with column/row, or figure) 6.1-2 6.1.1.1 RAI No.612, Question No. 06.01.01-22 Added Austenitic stainless steel base metal used for the pressure retaining materials has a limited carbon content not exceeding 0.05% (heat analysis) and 0.06% (product analysis) when the standard grade stainless steel is used. During the detailed design, MHI will determine if there are local areas where flow stagnation may be present resulting in dissolved oxygen content greater than 0.10 ppm in piping and components that have a normal operating temperature above 200°F. For piping and components where the above conditions exist, stainless steel with a carbon content less than or equal to 0.03% will be used. in the last of 4th paragraph.

6.1-2 6.1.1.1 RAI No.612, Question No. 06.01.01-23 Added The minimum preheat temperatures used for welding carbon and low alloy steels in ESF systems will meet the guidelines listed in ASME Code Section III, Appendix D, Article D-1000. in the 8th paragraph.

6.1-3 6.1.1.1 RAI No.612, Question No. 06.01.01-23 Added The recommendations of RG 1.50, Control of Preheat Temperature for Welding of Low Alloy Steel, (Ref. 6.1-14) are applied during weld fabrication. in the last paragraph.

6.1-4 6.1.1.2.2 RAI No.612, Question No. 06.01.01-23 Deleted The recommendations of RG 1.50, Control of Preheat Temperature for Welding of Low Alloy Steel, (Ref. 6.1-14) are applied during weld fabrication.

6.1-10 Table 6.1-1 RAI No.612, Question No. 06.01.01-20 Added weld filler material SFA-5.11 and SFA-5.14 6.1-17 Table 6.1-2 RAI No.612, Question No. 06.01.01-20 Added weld filler material SFA-5.11 and SFA-5.14 Page 1 of 1

6. ENGINEERED SAFETY FEATURES US-APWR Design Control Document The construction materials of ESF systems are compatible with core coolant and containment spray solutions. The ESF construction materials that would be exposed to core coolant and containment spray solutions in the event of a DBA are listed in Table 6.1-2.

The requirements from RG 1.44 (Ref. 6.1-4) are followed during the manufacture and construction of the ESF components and structures. The material used to fabricate the safety significant portions of the ESF systems (including supports) is highly resistant to corrosion. Process controls are enforced during all aspects of the component fabrication and construction to minimize the exposure of stainless steel to contaminants that could lead to stress-corrosion cracking. To avoid significant sensitization during fabrication and assembly of austenitic stainless steel components of the ESF, halogens and halogen-bearing compounds (e.g., die lubricants, abrasives, marking compounds, and masking tape) are not used in the welding processes during the construction of ESF components. Austenitic stainless steel base materials for ESF applications are solution annealed to prevent sensitization and stress corrosion cracking. Furnace-sensitized materials are not used in ESF systems. When practical, solution heat-treating includes rapid cooling rates following welding to minimize the formation of carbon deposits in the heat affected zone of the material. Austenitic stainless steel base metal used for the pressure retaining materials has a limited carbon content not exceeding 0.05% (heat analysis) and 0.06% (product analysis) when the standard grade stainless steel is used.

During the detailed design, MHI will determine if there are local areas where flow stagnation may be present resulting in dissolved oxygen content greater than 0.10 ppm in piping and components that have a normal operating temperature above 200°F. For piping and components where the above conditions exist, stainless steel with a carbon content less than or equal to 0.03% will be used.

All ESF components in contact with core coolants and containment spray solutions are either fabricated from or clad with austenitic stainless steel. Cold-worked austenitic stainless steel is not used for pressure boundary applications. If such material is used for other applications when there is no proven alternative available, cold work is controlled, measured and documented during each fabrication process. An augmented inservice inspection (ISI) is conducted to ensure the structural integrity of such components during service, which is described in Section 6.6. Cold-worked austenitic stainless steels have a maximum 0.2 percent offset yield strength of 620 MPa ( 90,000 psi ) to reduce the probability of stress-corrosion cracking in ESF systems.

Operating experience has demonstrated that certain nickel-chromium-iron alloys are susceptible to stress-corrosion cracking. When necessary, nickel-chromium-iron alloys used in the fabrication of ESF components in the US-APWR design is limited to Alloy 690. Alloy 690 was shown to have a high resistance to stress-corrosion cracking.

Fracture toughness properties of the materials used in ESF components are in complete agreement with the ASME Code Section III, Subarticles NC/ND/NE-2300 and this agreement maintained.

The control of welding, heat treatment, welder qualification, and contamination protection for ESF ferritic and austenitic stainless steel material fabrication are described in Chapter 5, Subsection 5.2.3. The minimum preheat temperatures used for welding Tier 2 6.1-2 Revision 23

6. ENGINEERED SAFETY FEATURES US-APWR Design Control Document carbon and low alloy steels in ESF systems will meet the guidelines listed in ASME Code Section III, Appendix D, Article D-1000.

For areas of limited access, welder qualification includes a simulated access mockup equivalent to the physical access and visibility of the production weld, in compliance with Regulatory Guide (RG) 1.71 (Ref. 6.1-5).

The effect of core coolant and containment spray solutions on austenitic stainless steel in a post-LOCA environment has been investigated (Ref. 6.1-6). This report provides test data and concludes that no cracking is anticipated on any equipment (stressed, sensitized or non-sensitized) even in the presence of postulated levels of chlorides and fluorides, provided the emergency core cooling solution is maintained above pH of 7.0.

The recommendations of RG 1.50, Control of Preheat Temperature for Welding of Low Alloy Steel, (Ref. 6.1-14) are applied during weld fabrication.

6.1.1.2 Composition and Compatibility of Core Cooling Coolants and Containment Sprays Controls are instituted to maintain the chemistry of the borated reactor coolant and the borated water in the RWSP. Chlorides and fluorides, which promote intergranular stress-corrosion cracking corrosion, are managed such that their concentrations are below 0.15 ppm. During periods of high temperatures, dissolved oxygen concentrations remain below 0.10 ppm. The controls include the chemical and volume control system (CVCS) and the spent fuel pit cooling and purification system (SFPCS). Details on these control systems are provided in Chapter 9, Subsection 9.3.4, for the CVCS and in Subsection 9.1.3 for the SFPCS.

6.1.1.2.1 Compatibility of Construction Materials with Core Cooling Coolants and Containment Sprays The provision of RG 1.44 (Ref. 6.1-4) are followed during the manufacture and construction of the ESF components and structures. The material used to fabricate the safety, significant portions of the ESF systems (including supports) is highly resistant to corrosion. The sources of corrosion may originate with the fluid (to include air in the ESF air clean-up applications) contained and delivered, as well as from external sources.

Borated reactor coolant, borated emergency make-up water, and a wetting containment spray that combines these fluids with sodium tetraborate decahydrate (NaTB) are important potential sources of such internal and external corrosion.

The pH of the ESF fluids is controlled during a DBA using NaTB baskets as a buffering agent. NaTB baskets are placed in the containment to maintain the desired post-accident pH conditions in the recirculation water. Maintaining the pH in the RWSP avoids stress-corrosion cracking of the austenitic stainless steel components and avoids excessive generation of hydrogen attributable to corrosion of containment metals. The information regarding boric acid in the RWSP water and NaTB in the containment is described in Subsection 6.3.1.3, Subsection 6.3.2.2.5, and Table 6.3-5. Aluminum and zinc are materials within the containment that would yield hydrogen gas by corrosion from the emergency cooling or containment spray solutions in the containment, and their use is limited as much as possiblepracticable.

Tier 2 6.1-3 Revision 23

6. ENGINEERED SAFETY FEATURES US-APWR Design Control Document The materials used in the fabrication of the ESF components are corrosion resistant in normal operation and the post-LOCA environment. General corrosion is negligible with the exception of low-alloy and carbon steels. Some materials within the containment would yield hydrogen gas by corrosion from the emergency cooling or containment spray solutions. Their use is limited as much as practicable (Ref. 6.1-7).

Borated water is used in the RCS and the RWSP. The water quality requirements for the RCS and RWSP are described in Chapter 9, Subsection 9.3.4 and Table 6.1-3, respectively. The pH of the RWSP during a LOCA is adjusted by the NaTB baskets.

The concrete that forms the structure of the RWSP is clad in stainless steel which inhibits the leach-out of chlorides and other contaminants into the RWSP water.

Therefore, the compatibility of the ESF components is preserved in the post-LOCA environment.

The use of particulate based insulation such as Min-K-based pipe insulation is prohibited in containment. Non-metallic (thermal) insulation is controlled in accordance with RG 1.36 (Ref. 6.1-8) to control the leachable concentrations of chlorides, fluorides, sodium compounds, and silicates. Chapter 5, Subsection 5.2.3.2.3, provides further details on the external insulation requirements which are also applicable to ESFs. Close attention to regulatory requirements and guidance ensures material compatibility between US-APWR construction materials and ESF fluids.

6.1.1.2.2 Controls for Austenitic Stainless Steel Chapter 5, Subsection 5.2.3, describes the controls employed during material selection to preclude the severe sensitization of stainless steel materials to be used for fabrication.

For example, cold worked austenitic stainless steel (300 series) typically is solution heat treated. Controls may be based on, but are not limited to, those imposed by Appendix B to 10CFR50, Appendix B part, 50, "Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants", with particular emphasis on Criteria VII, Control of Purchased Material, Equipment, and Services; VIII, Identification and Control of Materials, Parts, and Components; and IX, Control of Special Processes (Ref. 6.1-9).

When using fresh water to flush systems containing austenitic stainless steel components following construction, a chloride stress-corrosion cracking inhibitor is used in the flushing medium. The process of cleaning of materials and components, cleanliness control, and pre-operational flushing for systems that contain austenitic stainless steel components follows RG 1.37 (Ref. 6.1-11) and the quality assurance program complies with the provisions and recommendations provided by ASME NQA 1994, Part II (Ref. 6.1-10). This process includes documentation to verify the compatibility between theof materials used in manufacturing ESF components and thewith ESF fluids.

Chapter 5, Subsection 5.2.3 describes control of welding, heat treatment, welder qualification, and contamination protection for ferritic and austenitic stainless steels material fabrication which are also applicable to ESFs. The ferrite content in stainless steel weld metal will be controlled in accordance with the recommendations of RG 1.31 (Ref. 6.1-13). The recommendations of RG 1.50, Control of Preheat Temperature for Welding of Low Alloy Steel, (Ref. 6.1-14) are applied during weld fabrication.

Tier 2 6.1-4 Revision 23

6. ENGINEERED SAFETY FEATURES US-APWR Design Control Document Table 6.1-1 Principle Engineered Safety Feature Pressure Retaining Material Specifications (Sheet 3 of 3)

ESF Component Material Class, Grade or Type RWSP ASTM A 572 Grade 60 ASTM A 240 Gr. TP304L Fitting / Flange SA-312 Gr. TP304, TP304L SA-358 Gr. 304, 304L SA-182 Gr. F304, F304L SA-479 Type 304, 304L ESF Filter System See Subsection 6.5.1.7 Weld Filler Material SFA-5.1 E6018, E7018, E6016, E7016 SFA-5.4 E308-16, E309-16, E308L-16, E309L-16 SFA-5.5 E9018-B3, E9016-B3 SFA-5.9 ER308, ER309, ER308L SFA-5.11 ENiCrFe-7 SFA-5.14 ERNiCrFe-7 SFA-5.18 ER70S-2, ER70S-3, ER70S-4, ER70S-6, ER70S-G SFA-5.22 E309LT1-1/4, E308LT1-1/4 SFA-5.28 ER90S-B3 ESF Component Material Specification Safety Injection System Safety Injection Pump SA-216 Gr. WCB and WCC SA-217 Gr.WC9 SA-351 GrCF8M and CF3M SA-487 Gr.CA6NM SA-182 TP F304, F304L, F316, F316L Gr.

F11, F22, FXM-19 SA-350 Gr.LF1 and LF2 Accumulator SA-516 Gr.70 SA-182 F304 SA-516 Gr.70 Piping SA-106, Gr. B and C Tier 2 6.1-10 Revision 23

6. ENGINEERED SAFETY FEATURES US-APWR Design Control Document Table 6.1-2 Principle Engineered Safety Features Materials Exposed to Core Coolant Water and Containment Spray (Sheet 2 of 2)

ESF Component Material Class, Grade or Type Piping Class 1 Piping See Table 5.2.3-1 Class 2 Piping SA-312 Gr. TP304, TP304L SA-358 Gr. 304, 304L Valves Class 1 Valves See Table 5.2.3-1 Class 2 Valves The material for Class 2 valves are the same as Class 1. See Table 5.2.3-1 RWSP ASTM A 572 Gr. 60 ASTM A 240 Gr. TP304L Fitting / Flange SA-312 Gr. TP304, TP304L SA-358 Gr. 304, 304L SA-182 Gr. F304, F304L SA-479 Type 304, 304L ESF Filter System See Subsection 6.5.1.7 Weld Filler Material SFA-5.1 E6018, E7018, E6016, E7016 SFA-5.4 E308-16, E309-16, E308L-16, E309L-16 SFA-5.5 E9018-B3, E9016-B3 SFA-5.9 ER308, ER309, ER308L SFA-5.11 ENiCrFe-7 SFA-5.14 ERNiCrFe-7 SFA-5.18 ER70S-2, ER70S-3, ER70S-4, ER70S-6, ER70S-G SFA-5.22 E309LT1-1/4, E308LT1-1/4 SFA-5.28 ER90S-B3 ESF Component Material Specification Safety Injection System, Containment Spray System Piping/Tubing SA-106, Gr. B and C SA-155 Gr. KC70 Class 1 and 70 Class 1 SA-213, TP 304, 304L, and 316 SA-249, TP 304L Tier 2 6.1-17 Revision 23

US-APWR DCD Revision 2 RAI Tracking Report MUAP-09026(R5)

Chapter 9 Mitsubishi Heavy Industries, LTD.

US-APWR DCD Chapter 9 Rev. 2, Tracking Report Rev. 5 Change List Location (e.g., subsection with Page paragraph/ sentence/ Description of Change item, table with row/column, or figure)

RAI 524-4020, Question 12.03-35 9.1-29 Section 9.1.4.2.2.2 Add the function requirement for the reactor cavity drain line.

RAI 633-4857, Question 09.01.04-21 9.1-31 Section 9.1.4.2.2.2 Added the following after the last paragraph.

Plant procedures contain measures to prevent and mitigate inadvertent reactor cavity drain-down events.

Reactor refueling procedures require that valve positions of potential reactor cavity drain paths are verified prior to filling the refueling cavity. Operating procedures direct operators to monitor control room indications for reactor cavity seal leakage during refueling operations.

Maintenance procedures address periodic maintenance and inspection of the permanent cavity seal and other seals and plugs in accordance with vendor recommendations. Emergency response procedures provide direction to operators regarding the proper response to pool drain down events.

RAI 616-4865, Question 09.01.05-18 and Question 9.1-35 Section 9.1.5.1 09.01.05-19 Modified the second paragraph as follows.

The OHLHS cranes are designed to meet the criteria specified in CMAA-70, 2000, Specifications for Top Running Bridge and Gantry Type Multiple Girder Electric Overhead Traveling Cranes (Ref. 9.1.7-25) and Chapter 2-1 of ASME B30.2-2005, Overhead and Gantry Cranes (Ref. 9.1.7-22). The PCCV polar crane main and auxiliary hoist, equipment hatch hoist, and the spent fuel cask handling crane main hoist are designed as single-failure-proof ASME NOG-1 Type l cranes in accordance with NUREG-0554, Single-Failure-Proof Cranes for Nuclear Power Plants, (Ref. 9.1.7-19) and ASME NOG-1, Rules for Construction of Overhead and Gantry Cranes (Top Running Bridge, Multiple Girder) (Ref. 9.1.7-20), to handle the maximum critical loads for the area in which these cranes operate. The single-failure proof cranes each include at least two holding brakes. Each of the two required holding brakes has a torque rating of at least 125% of the rated load hoisting torque at the point of brake application. The reeving design of the single-failure-proof cranes is such that a single rope failure will not result in loss of the lifted load. Note that the suspension hoist and auxiliary hoist of the spent fuel cask handling will not handle critical loads and are not designed as single-failure-proof. However, they meet the electrical performance requirements of Type ll cranes as required by Section 6320 (c) of ASME NOG-1 (Ref. 9.1.7-20).

US-APWR DCD Chapter 9 Rev. 2, Tracking Report Rev. 5 Change List Location (e.g., subsection with Page paragraph/ sentence/ Description of Change item, table with row/column, or figure)

RAI 616-4865, Question 09.01.05-18 9.1-37 Section 9.1.5.2 Modified the first two sentences of the first paragraph as follows.

The primary pieces of equipment used in the OHLHS are the spent fuel cask handling crane in the fuel handling area, equipment hatch hoist in the PCCV and the polar crane in the PCCV. The spent fuel cask handling crane, equipment hatch hoist and the polar crane are designed in accordance with the provisions of NUREG-0554 and ASME NOG-1 as Type l single-failure-proof cranes.

RAI 616-4865, Question 09.01.05-18 9.1-39 Section 9.1.5.2.4 Added the 9.1.5.2.4 Equipment Hatch Hoist subsection after the Subsection 9.1.5.2.3.

RAI 616-4865, Question 09.01.05-18 9.1-40 Section 9.1.5.3 Added the following text as the third bullet of the first paragraph in Subsection 9.1.5.3.

Fabricating and erecting equipment hatch hoist that complies with the requirements of NUREG-0554 (Ref.

9.1.7-19). This is accomplished by designing the crane in conformance with ASME NOG-1 (Ref. 9.1.7-20). All slings are supplied in accordance with ANSI/ASME B30.9 (Ref.

9.1.7-24). The slings are of metallic material and have dual/redundant load paths or are capable of supporting a load twice the weight of the handled load.

RAI 616-4865, Question 09.01.05-19 9.1-40 Section 9.1.5.3 Modified the third sentence and delete the fourth sentence of the relocated third bullet (fourth bullet) of the first paragraph in Subsection 9.1.5.3 as follows.

Administrative control procedures are also required to be used to assure that the auxiliary hoist of the spent fuel cask handling crane does not handle heavy loads that could have adverse consequences for nuclear safety.

RAI 616-4865, Question 09.01.05-18 and Question 9.1-41 Section 9.1.5.3 09.01.05-19 Modified the first sentence of the second paragraph in Subsection 9.1.5.3 as follows.

Except for the OHLHS polar crane main and auxiliary hoist, equipment hatch hoist and main hoist of the spent fuel cask handling crane; miscellaneous cranes and hoists with heavy load capacities as listed in Table 9.1.5-3 are not designed as single-failure-proof.

RAI 616-4865, Question 09.01.05-19 9.1-41 Section 9.1.5.3 Modified the first sentence and delete the second sentence of the first bullet under the second paragraph in Subsection 9.1.5.3 as follows.

The non-single-failure-proof cranes and hoists in Table 9.1.5-3 are not located over or adjacent to fuel

US-APWR DCD Chapter 9 Rev. 2, Tracking Report Rev. 5 Change List Location (e.g., subsection with Page paragraph/ sentence/ Description of Change item, table with row/column, or figure) assemblies.

RAI 616-4865, Question 09.01.05-18 and Question 9.1-42 Section 9.1.5.4 09.01.05-19 The following two paragraphs were added (via RAI 563-4386 question 09.01.05-17) as the third and fifth paragraphs to Subsection 9.1.5.4.

Critical welds to support the polar crane main and auxiliary hoist , the equipment hatch hoist and spent fuel cask handling crane main hoist are identified and subject to non-destructive examination in accordance with Section 7200 and Paragraph 4251.4 of ASME NOG-1.

And No-load testing of the polar crane main and auxiliary hoist, equipment hatch hoist and spent fuel cask handling crane main hoist is performed in accordance with Paragraph 7421 of ASME NOG-1.

RAI 616-4865, Question 09.01.05-19 9.1-53 Table 9.1.5-2 Modified Polar Crane Auxiliary Hook hoisting speed in Table 9.1.5-2.

Middle speed from 6.0m/min to 2.4m/min and first speed from 12.0m/min to 3.0m/min.

RAI 616-4865, Question 09.01.05-18 9.1-54 Table 9.1.5-3 Added the new specification table for the Equipment Hatch Hoist as new Table 9.1.5-3.

RAI 616-4865, Question 09.01.05-18 and Question 9.1-55 Table 9.1.5-4 09.01.05-19 Re-labeled Table 9.1.5-3 as Table 9.1.5-4 and modified specification data for the Polar Crane, Spent Fuel Cask Handling Crane and PCCV Equipment Hatch Hoist.

RAI 532-4019, Question 12.02-27 9.3-24 Section 9.3.4.1.2.3 Add the explanation about the purification flow rate during shutdown.

9.4-13 Section 9.4.3.1.2.1 RAI 634-4845, Question 09.04.03-12 Modified to clarify the areas in which auxiliary building HVAC system keeps dose level due to the airborne radioactivity below the allowable values set by 10 CFR 20.

9.4-21 Section 9.4.3.4.1 RAI 634-4845, Question 09.04.03-13 Added the following after the last paragraph to clarify that confirmation of ventilation flow balancing is performed such that an unmonitored release will not occur under credible worst-case ventilation balance conditions.

The auxiliary building HVAC system ventilation flow balancing also is inspected periodically such that an

US-APWR DCD Chapter 9 Rev. 2, Tracking Report Rev. 5 Change List Location (e.g., subsection with Page paragraph/ sentence/ Description of Change item, table with row/column, or figure) unmonitored release will not occur under credible worst-case ventilation balance conditions.

9.4-46 Section 9.4.6.2.4.1 RAI 615-4816, Question 06.05.01-19 Modified the last paragraph of Section 9.4.6.2.4.1 as follows.

The containment radiation monitors RMS-RE-040 and 041 described in Subsection 11.5.2.2.1 provide a means for detection of unidentified leakage into the containment atmosphere from the reactor coolant pressure boundary (RCPB). When the unidentified leakage rates increases, alarms are initiated in the MCR and a containment purge isolation signal is generated. Upon receipt of the isolation signal, the containment low volume purge system containment isolation valves are automatically close. The radiation monitors are required in normal operation as described in DCD Subsection 5.2.5.4.

9.4-46 Section 9.4.6.2.4.1 RAI 634-4845, Question 09.04.03-11 Added the following to the end of the eighth paragraph of Section 9.4.6.2.4.1 to clarify the containment low volume purge system operation when the exhaust is aligned to the auxiliary building HVAC system.

When exhaust from the auxiliary building HVAC system is filtered by the containment low volume purge exhaust filtration unit, the containment low volume purge system containment isolation valve is manually closed and the containment low volume purge supply fan is manually stopped.

9.4-47 Section 9.4.6.2.4.2 RAI 615-4816, Question 06.05.01-19 Modified the last paragraph of Section 9.4.6.2.4.2 as follows.

The containment radiation monitors RMS-RE-040 and 041 described in Subsection 11.5.2.2.1 provide a means for detection of unidentified leakage into the containment atmosphere from the reactor coolant pressure boundary (RCPB). When the unidentified leakage rates increases, alarms are initiated in the MCR and a containment purge isolation signal is generated. Upon receipt of the isolation signal, the containment high volume purge system containment isolation valves are automatically close. The radiation monitors are required in normal operation as described in DCD Subsection 5.2.5.4.

9.4-53 Section 9.4.8 RAI 642-4770, Question 09.04.01-24 Added the following as a reference.

Safety-Related Air Conditioning, Heating, Cooling and

US-APWR DCD Chapter 9 Rev. 2, Tracking Report Rev. 5 Change List Location (e.g., subsection with Page paragraph/ sentence/ Description of Change item, table with row/column, or figure)

Ventilation Systems Calculations, MUAP-10020-P Rev.0 (Proprietary) and MUAP-10020-NP Rev.0 (Non-Proprietary), November, 2010.

9.4-82 Figure 9.4.3-1 RAI 634-4845, Question 09.04.03-13 Revised Figure 9.4.3-1

9. AUXILIARY SYSTEMS US-APWR Design Control Document A new fuel assembly stored in the new fuel storage racks is transferred to the spent fuel pit to prepare for refueling.

A new fuel assembly stored in the new fuel racks is lifted using the suspension hoist of the spent fuel cask handling crane, and transferred to the new fuel elevator located in the fuel inspection pit. The new fuel assembly is then lowered using the new fuel elevator for access by the fuel handling machine. The new fuel assembly is latched by the spent fuel assembly handling tool on the fuel handling machine, and is lifted using the fuel handling machine mast tube or auxiliary hoist and then transferred to the spent fuel pit for temporary storage in the spent fuel rack.

General arrangement figures for the US-APWR are presented in Subsection 1.2.1.7.

9.1.4.2.2.2 Reactor Refueling Operations Reactor refueling operations are divided into four phases: preparation, reactor disassembly, fuel handling, and reactor assembly. Refueling operations are outlined below and performed in accordance with operating procedures defined in Subsection 13.5.2.

• Phase I - Preparation The reactor is placed into cold shutdown mode as defined in the Technical Specifications, Chapter 16. The refueling water and reactor coolant are borated to assure the core remains approximately 5% below criticality during refueling operations based on the maximum reactivity of the fuel to be cycled through an US-APWR.

The water level in the refueling cavity and the spent fuel handling pit and interconnected pits is maintained at an elevation sufficient to keep radiation levels within personnel access limits when the fuel assemblies are being removed and transported from the core to the spent fuel racks in accordance with RG 1.13.

The radiation and environmental levels are monitored to assure levels do not exceed personnel access limits.

Upon achieving safe radiation and environmental conditions, the LLHS system is tested and the refueling machine overload is verified to be within operable. This is accomplished by using the mockup fuel assembly nozzle attached to the floor of the refueling cavity.

• Phase II - Reactor Disassembly The reactor vessel head assembly is prepared for refueling by disconnecting electrical cabling, seismic support tie rods, in-core instrumentation, and cooling duct work. The refueling cavity is prepared by:

- Closing and locking the reactor cavity drain line

- Removing the blind flange of the fuel transfer tube Tier 2 9.1-29 Revision 32

9. AUXILIARY SYSTEMS US-APWR Design Control Document

- The fuel container is pivoted to the horizontal position. The fuel transfer car is moved back through the transfer tube to the refueling area in R/B. The fuel container is pivoted to the vertical position again.

- The irradiated fuel is grasped by the fuel handling machine. The fuel is then transferred to the spent fuel rack. If needed, the spent fuel is transferred to fuel inspection pit to perform underwater visual inspections before transferring to the spent fuel rack, or inspected after completion the refueling (during normal operation).This process is continued until the core is off loaded. SFP level is maintained at normal throughout the refueling process to assure adequate radiation protection for personnel.

- The rod control clusters, the thimble plugs, and the burnable poison rod assemblies are shuffled in the SFP by using long handled tools on the fuel handling machine bridge.

- Irradiated and new fuel assemblies are individually lifted from a spent fuel rack by using the fuel handling machine, transferred to the up ender, and transferred to inside containment by reversing the core unloading process.

Phase IV - Reactor Assembly The reactor assembly is accomplished by reversing the process described in Phase II - Reactor Disassembly.

Plant procedures contain measures to prevent and mitigate inadvertent reactor cavity drain-down events. Reactor refueling procedures require that valve positions of potential reator cavity drain paths are verified prior to filling the refueling cavity. Operating procedures direct operators to monitor control room indications for reactor cavity seal leakage during refueling operations. Maintenance procedures address periodic maintenance and inspection of the permanent cavity seal and other seals and plugs in accordance with vendor recommendations. Emergency response procedures provide direction to operators regarding the proper response to pool drain down events.

9.1.4.2.2.3 Spent Fuel Storage The spent fuel assemblies are stored in the SFP until fission product activity is low enough to permit shipment from the site or to be placed in dry storage. Spent fuel storage and cooling is discussed in Subsections 9.1.2 and 9.1.3, respectively.

9.1.4.2.2.4 Spent Fuel Shipment The procedure for the spent fuel shipment is as follows:

The spent fuel cask is received into the R/B by way of the refueling area truck access bay at elevation 3 ft - 7 in. The spent fuel cask is raised from the truck using the spent fuel cask handling crane through the access hatch in the floors at elevation 25 ft - 3 in and 76 ft - 5 in the R/B refueling area.

Tier 2 9.1-31 Revision 32

9. AUXILIARY SYSTEMS US-APWR Design Control Document 9.1.5 Overhead Heavy Load Handling System The overhead heavy load handling system (OHLHS) consists of devices used for critical load handling evolutions. A critical load handling evolution is defined as the handling of a heavy load where inadvertent operations or equipment malfunctions, separately or in combination, could:

Cause a significant release of radioactivity Cause a loss of margin to criticality Uncover irradiated fuel in the reactor vessel or spent fuel pool Damage equipment essential to achieve or maintain safe shutdown The OHLHS exists in the reactor building, speficically thefuel storage and handling area, and in the pre-stressed concrete containment vessel (PCCV) of the reactor building. The functional arrangement and design characteristics of the OHLHS are discussed in the subsections procided below.

Heavy loads are defined as a load weighing more than one fuel assembly and its handling device. For the US-APWR, a fuel assembly weighs approximately 2,000 lbs with a handling tool weighing approximately 450 lbs. Therefore, for the US-APWR, a heavy load is defined as any load greater than the combined weight of approximately 2,450 lbs. This definition is established as a threshold for invoking the use of the OHLHS. The OHLHS is not used for the handling of new and spent fuel assemblies.

New and spent fuel assemblies are handled using the light load handling system (light load handling system) defined in Section 9.1.4 9.1.5.1 Design Bases The load that, if dropped, that would cause the greatest damage is a function of the area in which the OHLHS is operating. In the containment, this is defined as the integrated reactor head package/internals being lifted and transported to the lay down area. In the fuel handling area, this is defined as a full spent fuel cask being lifted and transported through the fuel handling area. In the area between the PCCV and the fuel handling area, this would be a reactor coolant pump motor.

The OHLHS cranes are designed to meet the criteria specified in CMAA-70, 2000, Specifications for Top Running Bridge and Gantry Type Multiple Girder Electric Overhead Traveling Cranes (Ref. 9.1.7-25) and Chapter 2-1 of ASME B30.2-2005, Overhead and Gantry Cranes (Ref. 9.1.7-22). The PCCV polar crane main and auxiliary hoist, equipment hatch hoist and the spent fuel cask handling crane main hoist are also designed as single-failure-proof ASME NOG-1 Type l cranes in accordance with NUREG-0554, Single-Failure-Proof Cranes for Nuclear Power Plants, (Ref. 9.1.7-19) and ASME NOG-1, Rules for Construction of Overhead and Gantry Cranes (Top Running Bridge, Multiple Girder) (Ref. 9.1.7-20), to handle the maximum critical loads for the area in which these cranes operate. The single-failure proof cranes each include at least two holidng brakes. Each of the tow required holding brakes has a torque rating of Tier 2 9.1-35 Revision 32

9. AUXILIARY SYSTEMS US-APWR Design Control Document at least 125% of the rated load hoisting torque at the point of brake application. The reeving design of the single-failure-proof cranes is such that a single rope failure will not result in loss of the lifted load. Note that the suspension hoist and auxiliary hoist of the spent fuel cask handling crane and the auxiliary hoists on these cranes will not handle critical loads and are not designed as single-failure-proof. However, they meet the electrical performance requirements of Type ll cranes as required by Section 6320 (c) of ASME NOG-1 (Ref. 9.1.7-20).

The use of the single failure proof crane precludes the need to perform load drop evaluations with the one exception. Single-failure proof cranes are designed so that any credible failure of a single component will not result in the loss of capability to stop and hold a critical load. However, ASME NOG-1 allows a drop of 1 inch for axle failure. It further defines the acceptable stopping distance as not exceeding 5 inches while lowering the maximum critical load at its maximum speed unless specified otherwise by the purchaser. These distances, 1 inch to 5 inch, represent a case where a critical load be lowered to the floor could impose an impact load on the floor and associated structural features, should a failure event occur within this range.

On occasion, the OHLHS may be used to handle non-critical loads of greater weight than the maximum critical load. For those occasions, the maximum non-critical load is the design rated load. The design rated load does not have the safety factor limits of a single-failure-proof crane required by NUREG-0554. The design rated load utilizes standard commercial practice safety factor limits.

One example is the special lifting of heavy loads during construction or plant shutdown conditions. Prior to the lifting of non-critical loads after initial fuel loading, it would be documented that the potential load drops due to inadvertent operations or equipment malfucntions, separatery or in combination, would not jeopardize safe shutdown functions, cause a significant release of radioactivity, a criticality accident, or inability to cool fuel within the reactor vessel or spent fuel pool. Non-critical lifts are those lifts that involve non-critical heavy loads, as defined in Section 9.1.5 above, that, because of theirlocaiton, timing, and the load path could not cause a significant release of radioactivity, cause a loss of margin to criticality, uncover irradiated fuel in the reactor vessel or spent fuel pool, or damage equipment essential to achieve or maintain safe shotdown. Non-critical lifts would be evaluated and documented in a manner similar to a critical heavy load lift, as required by the heavy load handling program to be developed by the COL applicant as required by COL 9.1 (6) and Subsection 9.1.5.3 of this DCD.

The areas of the plant in which the OHLHS is operated are shown in Figures 9.1.5-1 through 9.1.5-4. These figures represent the Fuel Handling Area and the interior of the PCCV. The OHLHS is designed to meet requirements of 10 CFR 50, Appendix A, specifically, GDC 1, 2, 4, and 5.

The operation, testing, maintenance, and inspection of OHLHS are controlled utilizing through the use of safe load paths as defined in Figures 9.1.5-1 through 9.1.5-4 and administrative control procedures.

The administrative control procedures govern the operation, testing, maintenance, and inspection of overhead heavy load handling system. These procedures incorporate the Tier 2 9.1-36 Revision 32

9. AUXILIARY SYSTEMS US-APWR Design Control Document requirements of and follow the recommendations and/or guidelines of the following documents:

Scope Reference Reference Title Chapter 5, General Control of Heavy Loads at Nuclear Power Section 5.1.1, requirements Plants (Ref. 9.1.7-21)

NUREG-0612 Crane Operators Chapter 2-3, Overhead and Gantry Cranes - Top Running (Training, ANSI/ASME Bridge, Single or Multiple Girder, Top Running qualifications, B30.2 Trolley Hoist (Ref. 9.1.7-22) and conduct.)

Inspection, Chapter 2-2, Overhead and Gantry Cranes - Top Running testing, and ANSI/ASME Bridge, Single or Multiple Girder, Top Running maintenance. B30.2 Trolley Hoist (Ref. 9.1.7-22) 9.1.5.2 System Description The primary pieces of equipment used in the OHLHS are the spent fuel cask handling crane in the fuel handling area, equipment hatch hoist in the PCCV and the polar crane in the PCCV. The spent fuel cask handling crane, equipment hatch hoist and the polar crane are designed in accordance with the provisions of NUREG-0554 and ASME NOG-1 as Type l single-failure-proof cranes. Therefore these cranes are designed to retain control of and continue to hold their maximum loads during an SSE. The OHLHS is seismic category II and Equipment Class 5, as described in Section 3.2.

Other than the single-failure-proof OHLHS, miscellaneous hoists and cranes with heavy load capacities are installed in safety-related areas of the US-APWR plant. Descriptions and data for all cranes and hoists that have heavy load capacities andwhich are installed over safe shutdown equipment are given in Table 9.1.5-3. The safety evaluations for those cranes and hoists are discussed in Subsection 9.1.5.3.

The OHLHS also includes equipment accessories (e.g., slings, and hooks, etc.)

instrumentation, physical stops and/or electrical interlocks, and associated administrative controls.

The applicable Codes and Standards are identified in Section 9.1.5.1.

9.1.5.2.1 Physical Arrangement The areas of the plant in which the spent fuel cask handling crane and polar crane operate are shown in Figures 9.1.5-1 through 9.1.5-4. The specifications for the spent fuel cask handling crane and the polar crane are given in Table 9.1.5-1 and 9.1.5-2. As shown, the spent fuel handling crane has three load handling hooks, the main, the auxiliary, and the suspension crane. The suspension crane is only used for new fuel assembly handling between a new fuel container to the new fuel storage area or between the new fuel storage rack and the basket on the new fuel elevator. Because of this Tier 2 9.1-37 Revision 32

9. AUXILIARY SYSTEMS US-APWR Design Control Document The polar crane range of movement is limited, in general, area defined by the hookhoist coverage ranges shown in Figures 9.1.5-4. The limitation is controlled by the configuration of the polar crane and by the fact;, travel is limited by the circumferential rail on which the polar crane travels.

For the heavy loads, polar crane movement is limited to exclude the area bounded by the reactor cavity by way of administrative control procedures.

The polar crane has a seismic restraint system which precludes derailment of the either the hoist trolley or the main bridge box girders during a seismic event.

The polar crane is stored in the parked position during plant operation. The parked position for the polar crane is parallel to the centerline of the C/V running between azimuth 0° and azimuth 180° with the hoist trolley located over the roof of the pressurizer room. The polar crane is designed to be used as a structural component during steam generator (SG) replacement. The driven components are not used during SG replacement.

9.1.5.2.4 Equipment Hatch Hoist During refueling. the equipment hatch is transferred up approximately 30 feet and placed in a secured position. The equipment hatch is guided upward by guides to avoid any unanticipated horizontal movement.

The equipment hatch hoist movement and storage is handled as follows:

The hoist is utilized only to lift the equipment hatch vertically. There are limitations on the speed the equipment hatch may be lifted by the design of the crane as a single failure proof crane and the guides in place to guide the hatch vertically.

The hoist does not move vertically or horizontally.

The equipment hatch hoist is base mounted to its support, which is designed to seismic category II requirements. The hoist supports are supported off the side of the containment.

The hoist is stored in a parked position and is not utilized for any lifts other than the equipment hatch.

9.1.5.3 Safety Evaluation The OHLHS is evaluated as to its ability to, assure there is no unacceptable release of radiation through mechanical damage to fuel, prevent damage that could compromise ability to maintain adequate degree of sub criticality, uncovering of fuel in the reactor vessel or spent fuel pool, and to prevent damage that could result in loss of essential safe-shutdown functions. This is accomplished by the following:

Tier 2 9.1-39 Revision 32

9. AUXILIARY SYSTEMS US-APWR Design Control Document Limiting the travel of the spent fuel cask handling machine to the areas shown in Figures 9.1.5-1 through 9.1.5-3 through the use of by physical stops on the travel rails of the machine and the hoist carriage. The machine is fabricated and erected in accordance with the requirements of NUREG-0554 ,

Single-Failure-Proof Cranes for Nuclear Power Plants, and (Ref. 9.1.7-19).

TThis is accomplished by procuring the machine in conformance with ASME NOG-1, Rules for Construction of Overhead and Gantry Cranes (Top Running Bridge, Multiple Girder), (Ref. 9.1.7-20). All lifting devices used for the spent fuel cask are designed and fabricated in accordance with ANSI N14.6, American National Standard for Special Lifting Devices for Shipping Containers Weighing 10,000 Pounds (4,500 kg) or More for Nuclear Materials, (Ref. 9.1.7-23) with the exception of slings which are supplied in accordance with ANSI/ASME B30.9, Safety Standards for Cableways, Cranes, Derricks, Hoists, Hooks, Jacks, and Slings - Slings, (Ref. 9.1.7-24). The slings are of metallic material and have dual/redundant load paths or are capable of supporting a load twice the weight of the handled load.

Fabricating and erecting a polar crane that complies with the requirements of NUREG-0554, Single-Failure-Proof Cranes for Nuclear Power Plants, (Ref. 9.1.7-19). This is accomplished by designing the crane in conformance with ASME NOG-1, Rules for Construction of Overhead and Gantry Cranes (Top Running Bridge, Multiple Girder), (Ref. 9.1.7-20). All lifting devices are designed and fabricated in accordance with ANSI N14.6, American National Standard for Special Lifting Devices for Shipping Containers Weighing 10,000 Pounds (4,500 kg) or More for Nuclear Materials, (Ref. 9.1.7-23) with the exception of slings which are supplied in accordance with ANSI/ASME B30.9, Safety Standards for Cableways, Cranes, Derricks, Hoists, Hooks, Jacks, and Slings - Slings, (Ref. 9.1.7-24). The slings are of metallic material and have dual/redundant load paths or are capable of supporting a load twice the weight of the handled load.

Fabricating and erecting equipment hatch hoist that complies with the requirements of NUREG-0554 (Ref. 9.1.7-19). This is accomplished by designing the crane in conformance with ASME NOG-1 (Ref. 9.1.7-20). All slings are supplied in accordance with ANSI/ASME B30.9 (Ref. 9.1.7-24). The slings are of metallic material and have dual/redundant load paths or are capable of supporting a load twice the weight of the handled load.

Administrative control procedures to govern operator training, load handling instructions, and equipment inspection. The administrative control procedures are developed in accordance with ANSI/ASME B30.2, Overhead and Gantry Cranes - Top Running Bridge, Single or Multiple Girder, Top Running Trolley Hoist, (Ref. 9.1.7-22). Administrative control procedures are also required to be used to assure that the auxiliary hoists of the spent fuel cask handling crane and polar cranes do does not handle heavy loads that could have adverse consequences for nuclear safety. For example, administrative control procedures may prevent the polar crane auxiliary hoist from being used to handle a reactor coolant pump motor unless the containment is defueled, or other measures are Tier 2 9.1-40 Revision 32

9. AUXILIARY SYSTEMS US-APWR Design Control Document taken to assure there is no potential for jeopardizing nuclear safety in the case of a load drop.

Except for the OHLHS polar crane main and auxiliary hoist, equipment hatch hoist and spent fuel cask handling crane main hoist, miscellaneous cranes and hoists with heavy load capacities as listed in Table 9.1.5-3 are not designed as single-failure-proof.

However, they are designed as seismic category ll equipment to prevent unacceptable structural interaction and failure during an SSE event. The non-single-failure proof cranes and hoists in Table 9.1.5-3 satisfy safety criteria for critical load handling evolutions in the following manner:

The non-sasingle-failure-proof cranes and hoists in Table 9.1.5-3 are not located over or adjacent to fuel assemblies., with the exception of the containment equipment hatch hoist. The hatch hoist is controlled by heavy load handling procedures, such that the hatch is not handled when a postulated load drop could result in unacceptable consequences. Therefore, a load handling incident involving the non-single-failure-proof cranes and hoists would not impact fuel assemblies.

The non-single-failure proof cranes and hoists are located over safe shutdown equipment, but the plant configuration provides redundancy by separation of the components to assure that the effects of a single load drop from these cranes and hoists would not jeopardize the ability to achieve or maintain safe shutdown conditions. The hoists associated with the safety injection pumps, CS/RHR pumps, EFW pumps, CCW pumps, and CCW Heat Exchangers are all located on the basement slab of the R/B at floor elevation 4, and each equipment train has its own room. Similarly, separation for other safe shutdown equipment serviced by non-single-failure proof cranes and hoists is achieved by walls, slabs, and/or adequate physical distance between adjacent equipment trains to assure that redundancy of safe shutdown functions is maintained in the case of a single load drop.

The non-single-failure proof cranes and hoists are dedicated to servicing particular pieces of safe shutdown equipment (such as pumps, valves, heat exchangers, and chillers) or systems that will be out-of-service when the cranes and hoists are used for handling heavy loads over them. The use of these cranes and hoists is administratively controlled by load handling procedures to prevent overhead load handling that could cause unacceptable damage to the dedicated equipment or systems when in service.

Therefore, load handling incidents involving non-single-failure-proof cranes and hoists listed in Table 9.1.5-3 will not jeopardize safe shutdown functions or cause a significant release of radioactivity, a criticality accident, or inability to cool fuel.

To assure proper handling of heavy loads during the plant life, the COL Applicant is to establish a heavy load handling program, including associated procedural and administrative controls, that satisfies commitments made in Subsection 9.1.5 of the DCD, and that meets the guidance of ANSI/ASME B30.2, ANSI/ASME B30.9, ANSI N14.6, Tier 2 9.1-41 Revision 32

9. AUXILIARY SYSTEMS US-APWR Design Control Document ASME NOG-1, CMAA Specification 70-2000, NUREG-0554, NUREG-0612, and NUREG-0800, Section 9.1.5. During the operating life of the plant, it is anticipated that temporarily installed hoists and mobile cranes will also be used for plant maintenance.

The heavy load handling program will include temporary cranes and hoists. The heavy load handling program will adopt a defense-in-depth strategy to enhance safety when handling heavy loads. For instance, the program will restrict lift heights to practical minimums and limit lifting activities as much as practical to plant modes in which load drops have the smallest potential for adverse consequences, particularly when critical loads are being handled. Further, prior to the lifting of heavy loads after initial fuel loading, the program will institute any additional reviews as necessary to assure that potential drops of these loads due to inadvertent operations or equipment malfunctions, separately or in combination, will not jeopardize safe shutdown functions, cause a significant release of radioactivity, a criticality accident, or inability to cool fuel within the reactor vessel or spent fuel pool.

9.1.5.4 Inspection and Testing Requirements The OHLHS components are subjected to various tests and inspections prior to being placed in service and are the subject of an inspection, tests, analyses, and acceptance criteria (ITAAC) program, which is detailed in Chapter 14, Section 14.3.

During fabrication, the quality assurance program of the Manufacturer satisfies the requirements of ASME NQA-1. The manufacturers inspection and testing program conforms to Sections 7100 and 7200 of ASME NOG-1, Rules for Construction of Overhead and Gantry Cranes (Top Running Bridge, Multiple Girder, (Ref. 9.1.7-20).

Critical welds to support the polar crane main and auxiliary hoist, equipment hatch hoist and spent fuel cask handling crane main hoist are identified and subject to non-destructive examination in accordance with Section 7200 and Paragraph 4251.4 of ASME NOG-1.

Prior to operation, the OHLHS is received, stored, and installed in accordance with Sections 7100, 7300, and 7400 of ASME NOG-1, Rules for Construction of Overhead and Gantry Cranes (Top Running Bridge, Multiple Girder, (Ref. 9.1. 7-20). Qualification of the assembled OHLHS is performed in accordance with Section 7500 of ASME NOG-1 (Ref. 9.1.7-20).

No-load testing of the polar crane main and auxiliary hoist, equipment hatch hoist and spent fuel cask handling crane main hoist is performed in accordance with Paragraph 7421 of ASME NOG-1.

Periodic tests and inspections of the OHLHS are performed in accordance with Chapter 2-2 of ANSI/ASME B30.2, Overhead and Gantry Cranes - Top Running Bridge, Single or Multiple Girder, Top Running Trolley Hoist, (Ref. 9.1.7-22).

Inspection and testing of special lifting devices and slings used in conjunction with the polar crane and spent fuel cask handling crane, are performed in accordance with ANSI N14.6 (Ref. 9.1.7-23) and ASME B30.9 (Ref. 9.1.7-24), respectively.

Tier 2 9.1-42 Revision 32

9. AUXILIARY SYSTEMS US-APWR Design Control Document 9.1.5.5 Instrumentation Requirements The OHLHS is equipped with mechanical and electrical limit devices to disengage power to the motors as the load hook approaches its travel limits or to prevent damage to other components when continued operation would potentially damage the OHLHS as required by NUREG-0554, Single-Failure-Proof Cranes for Nuclear Power Plants, (Ref. 9.1.7-19).

In addition to the limit devices, the control system is designed to include safety devices, which will assure the OHLHS returns to and/or maintains a secure holding position of critical loads in the event of a system fault. These safety devices are in addition to and separate from the control devices used for normal operation of the OHLHS. Emergency stop buttons are strategically placed at various locations to de-energize the OHLHS independent of the system controls. The overload sensing system is designed to be reset when switching the OHLHS between maximum critical load operations and design rate load operations. This resetting is performed remotely from the system controls and is governed by the OHLHS administrative control procedures.

The OHLHS driver control systems are designed using a combination of electrical and mechanical components. The control systems take into account the hoisting (raising and lowering) of the complete range of loads from the load hook itself up to and including the rated load in conjunction with the inertia of moving components, such as the motor armature, shafting and coupling, gear reducer, drum, etc. In general, the OHLHS is not contemplated to be used to lift individual spent fuel elements. The control system has been designed to be adaptable to include manual interlocks, which will preclude trolley and/or bridge movement while a spent fuel assembly is being hoisted free of the reactor vessel or a storage rack. The manual interlocks are controlled by administrative control procedures.

Instrumentation is installed within the motor control circuits to detect and react to malfunctions such as excessive electric current, excessive motor temperature, overspeed, overload, and overtravel. Control devices are installed to absorb the kinetic energy of the rotating components and arrest the hoisting movement should the load line or one of the dual revving systems fail, or should an overload and/or overspeed condition occur.

The drives are designed to conform to ASME NOG-1, Rules for Construction of Overhead and Gantry Cranes (Top Running Bridge, Multiple Girder), (Ref. 9.1.7-20) with respect to hoist speed, specifically Section 5331 of ASME NOG-1 (Ref. 9.1.7-20).

The complete operating control system, along with emergency control features is located in the cab on the OHLHS. Additional wireless remote control stations are also provided for remote operations of the OHLHS. The wireless remote control stations have the same control, including emergency, features as the cab mounted controls. The configuration of the controls stations are in accordance with Section 2-1.13 of ANSI/ASME B30.2, Overhead and Gantry Cranes - Top Running Bridge, Single or Multiple Girder, Top Running Trolley Hoist, (Ref. 9.1.7-22). The individual control stations are interlocked to permit only one station to be operable at a time.

9.1.6 Combined License Information COL 9.1(1) Deleted Tier 2 9.1-43 Revision 32

9. AUXILIARY SYSTEMS US-APWR Design Control Document Table 9.1.5-1 Specification of the Spent Fuel Cask Handling Crane

1. Type Overhead bridge crane

2. Operating device Radio remote control unit and cab on crane

3. Component supplied Trolley electric power

4. Electric power supply Power  : 460V ac, 60 Hz, 3 Phase Space Heater  : 230V ac, 60 Hz, Single Phase

5. Bridge Span 47'-3"

6. Top level of the rail Elevation 125'-8" Main HookHoist Auxiliary HookHoist Suspension Hoist

7. Capacity Metric ton 150 20 2

8. Lift ft-in (m) 124'-9" 124'-9" 69'-3" (38.003 m) (38.003 m) (21.0886 m)

9. HookHoist ft-in (m) Refer to Figure 9.1.5-1 and 9.1.5-2 Coverage

10. Hoisting m/min 0.12, 0.6, 1.2 0.45, 1.8, 4.5 2.1, 6.3 Speed

11. Traveling m/min Bridge: 0.6, 1.5, 6.0 Suspension Crane: 3.0, Speed 9.0 Trolley: 0.6, 1.5, 6.0 Hoist: 3.0, 9.0

12. Wire Material Stainless Steel (ATSM A 492 Type 304)

Table 9.1.5-2 Specification of the Polar Crane

1. Type Overhead bridge crane

2. Operating device Portable wireless control box on operating floor, Cab on crane

3. Component supplied electric Trolley power

4. Electric power supply Power  : 460V ac, 60 Hz, 3 Phase Space Heater  : 230V ac, 60 Hz, Single Phase

5. Bridge Span 142'-1"

6. Top level of the rail Elevation 145'-6" Main HookHoist Auxiliary HookHoist

7. Capacity Metric ton 250 50

8. Lift ft-in (m) 67'-9" 119'-1"

( 20.650 m) ( 36.296 m)

9. HookHoist Coverage ft-in (m) Refer to Figure 9.1.5-4

10. Hoisting Speed m/min 0.12, 0.6, 1.2 1.2, 2.46.0, 3.012.0

11. Traveling Speed m/min Bridge: 0.9, 1.8, 18.0 Trolley: 0.6, 3.42, 12.0

12. Wire Material Carbon Steel Tier 2 9.1-53 Revision 32

9. AUXILIARY SYSTEMS US-APWR Design Control Document Table 9.1.5-3 Specification of the Equipment Hatch Hoist

1. Type Base mounted Drum Hoist

2. Operating device control box

3. Component supplied Hoist electric power

4. Electric power supply Power  : 460V ac, 60 Hz, 3 Phase Main Hook

5. Capacity Metric ton 40

6. Lift ft-in (m) 29'-6" (8.99 m)

7. Hoisting m/min 2.1 or less Speed

8. Wire Material Carbon steel Tier 2 9.1-54 Revision 32

9. AUXILIARY SYSTEMS US-APWR Design Control Document Table 9.1.5-34 Cranes and Hoists Installed Over Safe Shutdown Equipment Maximum Single-Crane and ASME Seismic Crane/Hoist Type Location Load Rating Failure-Hoist NOG-1 Type Category (metric tons) proof Top-Running Main hoist 250 l Yes Polar Crane Overhead PCCV ll Bridge Auxiliary 50 I Yes Crane hoist Top- Main hoist 150 I Yes Spent Fuel Running Cask Auxiliary R/B(Fuel Overhead 20 NA No ll Handling hoist handling area)

Bridge Crane Suspension Crane 2 NA No hoist MSIV(main R/B (MS/FW steam Piping Area isolation Underhung overhead crane 10 NA No ll hung from roof valve)room slab) crane PCCV PCCV (above Equipment Base mounted Drum Hoist equipment 40 NAI NoYes ll Hatch Hoist hatch at azimuth 40°)

Safety R/B(SIP Injection Monorail Hoist Rooms, Floor 5 NA No ll Pump(SIP)

EL.-26-4)

Room Hoist R/B(CS/RHR CS/RHR Pump Rooms, Pump Room Monorail Hoist 5 NA No ll Floor Hoist EL.-26-4)

R/B(EFW EFW Pump Pump Rooms, Monorail Hoist 5 NA No ll Room Hoist Floor EL.-26-4)

CCW Pump Monorail Hoist 5 NA No ll Hoist R/B(CCW CCW Heat Rooms, Floor Exchanger Monorail Hoist EL.-26-4) 2 NA No ll Hoist East and West Essential PS/B(Baseme Chiller Unit Monorail Hoist 3 NA No ll nt Floor Hoist EL.-26-4)

Tier 2 9.1-55 Revision 32

9. AUXILIARY SYSTEMS US-APWR Design Control Document combination of the letdown orifices in the letdown flow path to accommodate the various plant operating conditions. The CVCS has sufficient makeup capacity to maintain the minimum required inventory in the event of minor leaks in the RCS, as discussed in Subsection 9.3.4.2.7.4.

9.3.4.1.2.2 Chemical Shim and Chemical Control The CVCS provides the following functions to support the water chemistry and chemical shim requirements of the RCS:

• Means of addition and removal of pH control chemicals for startup and normal operation.

• Means of addition and removal of soluble chemical neutron absorber (boron) and makeup water, to control reactivity changes resulting from the change in reactor coolant temperature between cold shutdown and hot full-power operation, burn-up of fuel and burnable poisons, buildup of fission products in the fuel, and xenon transients. Two boric acid tanks are capable of providing the total boric acid solution for refueling shutdown plus one cold shutdown from full power operation immediately following refueling.

• Means to control the oxygen concentration after venting the RCS prior to startup and suppress the oxygen generated by radiolysis of water in the reactor during power operation.

RCS chemistry changes are accomplished with a feed and bleed operation. The letdown and makeup paths are operated simultaneously and appropriate chemicals are provided at the suction of the charging pumps.

The water chemistry specification for the reactor coolant during normal operation is shown in Table 9.3.4-1.

9.3.4.1.2.3 Purification The CVCS removes radioactive corrosion products, ionic fission products, and fission gases from the reactor coolant to maintain low RCS activity levels. The CVCS purification capability takes into account occupational radiation exposure (ORE) to support ALARA goals.

The purification rate is based on minimizing ORE and providing access to the equipment for maintenance and inspection activities.

The CVCS has sufficient RCS purification and degasification capability to allow the reactor vessel head to be removed expeditiously during a refueling shutdown. In addition, purification during shutdowns has positive impact on reducing the ORE to workers during the outage. The CVCS is capable of providing purification flow up to 400 gpm when using the RHRS for letdown during shutdown. The CVCS supports the plant ALARA goals with its shutdown purification function.

Tier 2 9.3-24 Revision 23

9. AUXILIARY SYSTEMS US-APWR Design Control Document 9.4.3.1.2.1 Auxiliary Building HVAC system The auxiliary building HVAC system is designed to satisfy the following design bases:

• Provide and maintain proper operating environment within the required temperature range (Table 9.4-1) for areas housing mechanical and electrical equipment within the A/B, R/B, PS/B and AC/B during normal plant operation.

• Keep dose levels due to the airborne radioactivity in normally occupied areas below the allowable values set by 10 CFR 20 by supplying and exhausting sufficient airflow.

• Control exhaust fan airflow continuously and automatically at a predetermined value to maintain a slightly negative pressure in the controlled areas within A/B, R/B and AC/B relative to the outside atmosphere. This minimizes exfiltration from the radiological controlled areas during normal plant operation.

• Maintain airflow from areas of low radioactivity to areas of potentially higher radioactivity.

• Provide accessibility to system components for adjustment, maintenance and periodic inspection and testing of the systems equipment and components to assure proper equipment function and reliability and system availability.

• The auxiliary building HVAC system and containment low volume purge system are cross connected to allow the exhaust from the radiological controlled areas to be filtered by the containment low volume purge exhaust filtration units.

• Airborne radioactivity is monitored inside the exhaust air duct from the controlled areas (Subsection 12.3.4.2.8).

9.4.3.1.2.2 Non-Class 1E Electrical Room HVAC System The non-Class 1E electrical room HVAC system is designed to satisfy the following design bases:

• Provide and maintain the room ambient conditions within the required temperature range (Table 9.4-1) to support the continuous operation of the electrical equipment and components.

• Maintain the hydrogen concentration below 1% by volume of battery room.

• Provide accessibility to system components for adjustment, maintenance and periodic inspection and testing of the system equipment and components to assure proper equipment function, reliability and system availability.

9.4.3.1.2.3 Main Steam/Feedwater Piping Area HVAC System Tier 2 9.4-13 Revision 23

9. AUXILIARY SYSTEMS US-APWR Design Control Document The system is provided with adequate instrumentation, temperature, flows, and differential pressure indicating devices to facilitate testing and verification of equipment heat transfer capability and flow blockage.

Preoperational testing of the auxiliary building ventilation system is performed as described in Chapter 14, Verification Programs, to verify that system is installed in accordance with applicable programs and specifications. All HVAC system airflows are balanced in conformance with the design flow, path flow capacity, and proper air mixing temperature throughout the A/B, R/B, PS/B, and AC/B.

The system equipment and components are provided with proper access for initial and periodic inspection and maintenance during normal operation.

Air handling units are factory-tested in accordance with Air Movement and Control Association Standards (Ref.9.4.8-16, Ref.9.4.8-17, Ref.9.4.8-18). Air filters are tested in accordance with the American Society of Heating, Refrigerating and Air-Conditioning Engineers Standards (Ref.9.4.8-19, Ref.9.4.8-20). Cooling coils are hydrostatically tested in accordance with ASME,Section VIII (Ref. 9.4-14) and their performance is rated in accordance with the Air Conditioning and Refrigeration Institute Standards (Ref.9.4.8-21, Ref.9.4.8-22, Ref.9.4.8-23).

Air distribution ductwork is leak-tested in accordance with the Sheet Metal Air Conditioning Contractors National Association (Ref.9.4.8-24, Ref.9.4.8-25) and ASME N510 (Ref.9.4.8-8), and ASME AG-1 Section SA and TA (Ref.9.4.8-2).

System instruments are periodically calibrated and automatic controls are tested for activation at the design setpoints, in conformance with the design sequence of operation at all system operating modes.

9.4.3.4.1 Auxiliary Building HVAC System In addition to the general requirements in Subsection 9.4.3.4, the auxiliary building HVAC system safety-related isolation dampers are inspected periodically and the damper seats are replaced as required and tested in accordance with technical specification surveillance requirement for the annulus emergency exhaust system. The auxiliary building HVAC system ventilation flow balancing also is inspected periodically such that an unmonitored release will not occur under credible worst-case ventilation balance conditions.

9.4.3.4.2 Non-Class 1E Electrical Room HVAC System In addition to the general requirements in Subsection 9.4.3.4, battery fan operation is tested to insure automatic operation of the standby fan upon the airflow failure of the activated fan.

9.4.3.4.3 Main Steam/Feedwater Piping Area HVAC System The general requirements of Subsection 9.4.3.4 apply.

Tier 2 9.4-21 Revision 23

9. AUXILIARY SYSTEMS US-APWR Design Control Document the impacted area, and redirect exhaust airflow to these containment low volume purge exhaust filtration units. This minimizes the potential spread of radioactive contamination for the areas serviced by the auxiliary building HVAC system. When exhaust from the auxiliary building HVAC system is filtered by the containment low volume purge exhaust filtration unit, the containment low volume purge system containment isolation valve is manually closed and the containment low volume purge supply fan is manually stopped.

The containment low volume purge system meets the GDC 60 and 61 requirements based on compliance with RG 1.140 and control of radioactive material releases to environment. However, based on the results of the fuel handling accident analysis presented in DCD Section 15.7.4 with no credit for given for any filtration of released radionuclides, and the calculated offsite doses being remain well within the guideline dose limit values of 10 CFR 50.34, compliance with GDC 60 and 61 is not required for the postulated fuel handling accident condition.

The capacity of the containment low volume purge system is sized to maintain acceptably low levels of radioactivity, including noble gases, during normal plant operation.

The containment area includes four radiation monitors that are part of the Area Radiation Monitoring System (ARMS) described in Section 12.3.4.1. The monitors provide detection of radioactivity due to airborne particulate within the circulating air of the containment.

Following the detection of high levels of radioactivity by any of the four radiation monitors, alarms are indicated in the main control room and the containment isolation valves on the containment low volume purge system is automatically closed.

The containment radiation monitors RMS-RE-040 and 041 described in Subsection 11.5.2.2.1 provide a means for detection of unidentified leakage into the containment atmosphere from the reactor coolant pressure boundary (RCPB). When the unidentified leakage rates increases, alarms are initiated in the MCR and a containment purge isolation signal is generated. Upon receipt of the isolation signal, the containment low volume purge system containment isolation valves are automatically close. The radiation monitors are required in normal operation as described in DCD Subsection 5.2.5.4.

9.4.6.2.4.2 Containment High volume Purge System The containment high volume purge system is shown in Figure 9.4.6-1 and the equipment design data is presented in Table 9.4.6-1. The COL Applicant is to determine the capacity of cooling and heating coils that are affected by site specific conditions.

With the exception of the containment isolation valves, that are safety related and seismic category I, the containment high volume purge system does not serve any safety function.

Therefore, it is not safety-related. Non-safety related equipment and ductwork within, including supports, in areas containing safety-related equipment are supported as seismic Category II to prevent adverse interaction with safety-related systems during a seismic event.

The containment high volume purge system consists of a containment high volume purge air handling unit and an exhaust filtration unit, isolation valves dampers, ductwork and associated instrumentation and controls.

Tier 2 9.4-46 Revision 23

9. AUXILIARY SYSTEMS US-APWR Design Control Document the impacted area, and redirect exhaust airflow to these containment low volume purge exhaust filtration units. This minimizes the potential spread of radioactive contamination for the areas serviced by the auxiliary building HVAC system. When exhaust from the auxiliary building HVAC system is filtered by the containment low volume purge exhaust filtration unit, the containment low volume purge system containment isolation valve is manually closed and the containment low volume purge supply fan is manually stopped.

The containment low volume purge system meets the GDC 60 and 61 requirements based on compliance with RG 1.140 and control of radioactive material releases to environment. However, based on the results of the fuel handling accident analysis presented in DCD Section 15.7.4 with no credit for given for any filtration of released radionuclides, and the calculated offsite doses being remain well within the guideline dose limit values of 10 CFR 50.34, compliance with GDC 60 and 61 is not required for the postulated fuel handling accident condition.

The capacity of the containment low volume purge system is sized to maintain acceptably low levels of radioactivity, including noble gases, during normal plant operation.

The containment area includes four radiation monitors that are part of the Area Radiation Monitoring System (ARMS) described in Section 12.3.4.1. The monitors provide detection of radioactivity due to airborne particulate within the circulating air of the containment.

Following the detection of high levels of radioactivity by any of the four radiation monitors, alarms are indicated in the main control room and the containment isolation valves on the containment low volume purge system is automatically closed.

The containment radiation monitors RMS-RE-040 and 041 described in Subsection 11.5.2.2.1 provide a means for detection of unidentified leakage into the containment atmosphere from the reactor coolant pressure boundary (RCPB). When the unidentified leakage rates increases, alarms are initiated in the MCR and a containment purge isolation signal is generated. Upon receipt of the isolation signal, the containment low volume purge system containment isolation valves are automatically close. The radiation monitors are required in normal operation as described in DCD Subsection 5.2.5.4.

9.4.6.2.4.2 Containment High volume Purge System The containment high volume purge system is shown in Figure 9.4.6-1 and the equipment design data is presented in Table 9.4.6-1. The COL Applicant is to determine the capacity of cooling and heating coils that are affected by site specific conditions.

With the exception of the containment isolation valves, that are safety related and seismic category I, the containment high volume purge system does not serve any safety function.

Therefore, it is not safety-related. Non-safety related equipment and ductwork within, including supports, in areas containing safety-related equipment are supported as seismic Category II to prevent adverse interaction with safety-related systems during a seismic event.

The containment high volume purge system consists of a containment high volume purge air handling unit and an exhaust filtration unit, isolation valves dampers, ductwork and associated instrumentation and controls.

Tier 2 9.4-46 Revision 23

9. AUXILIARY SYSTEMS US-APWR Design Control Document The containment penetration and the containment isolation valves are constructed of fire rated material and as a fire barrier and are equivalent to any fire rated damper. This configuration will prevent the spread of fire from on fire area to another fire area.

This ventilation system contains ductwork in the auxiliary building and reactor building and there will be some penetration through fire barriers. Therefore, fire dampers are installed where ductwork penetrates a fire rated barrier.

The air handling unit consists of, in the direction of airflow, a low efficiency filter, a high efficiency filter, an electric heating coil, a chilled water cooling coil, and a supply fan.

Outside air is drawn and conditioned by the air handing unit and discharged into the containment through the containment high volume purge system penetration.

The exhaust filtration unit consists of, in the direction of airflow, a high efficiency filter, a HEPA filter, and an exhaust fan. The containment air is drawn through the containment high volume purge system penetration by the exhaust filtration unit and discharged to the atmosphere through the plant vent stack. The containment high volume purge system meets GDC 60 and 61 requirements based on compliance with RG 1.140 and control of radioactive material release to environment.

The capacity of the containment high volume purge system is sized to maintain acceptably low levels of radioactivity, including noble gases, during refueling operations.

The containment area includes four radiation monitors that are part of the Area Radiation Monitoring System (ARMS) described in Section 12.3.4.1. The monitors provide detection of radioactivity due to airborne particulate within the circulating air of the containment.

Following the detection of high levels of radioactivity by any of the four radiation monitors, alarms are indicated in the main control room and the containment isolation valves on the containment high volume purge system is automatically closed.

The containment radiation monitors RMS-RE-040 and 041 described in Subsection 11.5.2.2.1 provide a means for detection of unidentified leakage into the containment atmosphere from the reactor coolant pressure boundary (RCPB). When the unidentified leakage rates increases, alarms are initiated in the MCR and a containment purge isolation signal is generated. Upon receipt of the isolation signal, the containment high volume purge system containment isolation valves are automatically close. The radiation monitors are required in normal operation as described in DCD Subsection 5.2.5.4.

9.4.6.3 Safety Evaluation 9.4.6.3.1 Containment Fan Cooler System The containment fan cooler system has no safety-related function and therefore does not require a safety evaluation. However, a part of ductwork in the containment serving the containment fan cooler system are supported in accordance with seismic category II requirements to remain in place during the SSE and preclude damage to any safety-related structures, systems, or components located in the vicinity of the piping or the ductwork. As a further safety feature of the containment ventilation system, the fan Tier 2 9.4-47 Revision 23

9. AUXILIARY SYSTEMS US-APWR Design Control Document 9.4.8-15 Design, Inspection, and Testing Criteria for Air Filtration and Adsorption Units of Normal Atmosphere Cleanup Systems in Light-Water-Cooled Nuclear Power Plants, Regulatory Guide (RG) 1.140-2001, Revision2.

9.4.8-16 Laboratory Methods of Testing Fans for Rating, ANSI/AMCA 210-2007.

9.4.8-17 Laboratory Methods of Testing Air Circulator Fans for Rating, ANSI/AMCA 230-1999.

9.4.8-18 Industrial Process / Power Generation Fans: Establishing Performance Using Laboratory Models, ANSI/AMCA 802-2002 9.4.8-19 Gravimetric and Dust Spot procedures for Testing Cleaning Devices Used in General Ventilation for Removing Particulate Matter, ASHRAE 52.1-1992.

9.4.8-20 Method of Testing General Ventilation Air-Cleaning Devices for Removal Efficiency by Particle Size, ASHRAE 52.2-2007.

9.4.8-21 Forced-Circulation Air-Cooling and Air-Heating Coils, ARI 410-2001.

9.4.8-22 Performance Rating of Room Fan-coils, ARI 440-2005.

9.4.8-23 1999 Standard for Central Station Air-Handling Units, ANSI/ARI 430-1999 9.4.8-24 HVAC Air Duct Leakage Test Manual - First Edition; Technical Research Update - 92, SMACNA 1143-1985.

9.4.8-25 HVAC Systems Testing, Adjusting and Balancing - Third Edition, SMACNA 1780 - 2002 9.4.8-26 International Mechanical Code, 2003 Edition.

9.4.8-27 General Design Criteria for Nuclear Power Plants, NRC Regulation Title 10, Code of Federal Regulations, 10 CFR Part 50, Appendix A.

9.4.8-28 Seismic Design Classification, Regulatory Guide 1.29 Revision 4, March 2007.

9.4.8-29 Minimization of Contamination and Radioactive Waste Generation: Life-Cycle Planning. RG 4.21, Rev.0,U.S. Nuclear Regulatory Commission, Washington, DC, June 2008.

9.4.8-30 Safety-Related Air Conditioning, Heating, Cooling and Ventilation Systems Calculations, MUAP-10020-P Rev.0 (Proprietary) and MUAP-10020-NP Rev.0 (Non-Proprietary), November, 2010.

Tier 2 9.4-53 Revision 23

9. AUXILIARY SYSTEMS US-APWR Design Control Document Figure 9.4.3-1 Auxiliary Building HVAC System Flow Diagram Tier 2 9.4-82 Revision 23

US-APWR DCD Revision 2 RAI Tracking Report MUAP-09026(R5)

Chapter 10 Mitsubishi Heavy Industries, LTD.

US-APWR DCD Chapter 10 Rev. 2, Tracking Report Rev. 5 Change List Location (e.g., subsection with Page paragraph/sentence/ Description of Change Item, table with column/row, or figure) 10.3-15 Subsection 10.3.5.5 RAI 630-5044, Question No.10.04.06-16 The second paragraph Replaced the second paragraph.

10.3-16 Subsection 10.3.5.6 RAI 630-5044, Question No.10.04.06-16 The second paragraph Deleted the second sentence.

10.3-16 Subsection 10.3.5.6 RAI 630-5044, Question No.10.04.06-16 The third paragraph Deleted the third sentence.

10.3-16 Subsection 10.3.5.7 RAI 630-5044, Question No.10.04.06-16 The first paragraph Deleted the first sentence.

10.3-36 Subsection 10.3.5 RAI 630-5044, Question No.10.04.06-16 Table 10.3.5-1 Deleted the Table 10.3.5-1.

10.3-37 Subsection 10.3.5 RAI 630-5044, Question No.10.04.06-16 Table 10.3.5-1 Deleted the Table 10.3.5-1.

10.3-38 Subsection 10.3.5 RAI 630-5044, Question No.10.04.06-16 Table 10.3.5-1 Deleted the Table 10.3.5-1.

10.3-39 Subsection 10.3.5 RAI 630-5044, Question No.10.04.06-16 Table 10.3.5-1 Deleted the Table 10.3.5-1.

10.3-40 Subsection 10.3.5 RAI 630-5044, Question No.10.04.06-16 Table 10.3.5-2 Deleted the Table 10.3.5-2.

10.3-41 Subsection 10.3.5 RAI 630-5044, Question No.10.04.06-16 Table 10.3.5-3 Deleted the Table 10.3.5-3.

10.3-42 Subsection 10.3.5 RAI 630-5044, Question No.10.04.06-16 Table 10.3.5-3 Deleted the Table 10.3.5-3.

10.4-29 Subsection 10.4.6.2.1 RAI 630-5044, Question No.10.04.06-16 The fifth paragraph Deleted as shown in Table 10.4.6-2.

10.4-35 Subsection 10.4.6 RAI 630-5044, Question No.10.04.06-16 Table 10.4.6-2 Deleted the Table 10.4.6-2.

Page 1 of 1

10. STEAM AND US-APWR Design Control Document POWER CONVERSION SYSTEM

• Atmospheric contamination is detected by monitoring the quantity of dissolved oxygen in the condensate pump discharge and the condenser air removal rate.

10.3.5.3 Condensate Polishing A condensate polishing system removes suspended corrosion products and ionic contaminants. The system is capable of handling the condensate design flow rate.

This system is not normally used in all phases of plant operation.

The CFS recirculate water to the condenser prior to and during plant startup. The condensate polishing sysytem is used to remove corrosion products in this phase and thus prevent their ingress into the steam generators.

See Subsection 10.4.6 for a further description of condensate polishing.

10.3.5.4 Chemical Addition US-APWR employs an all volatile treatment (AVT) method to minimize general corrosion in the FWS, SGs and main steam piping. A pH adjusting chemical and an oxygen scavenger are injected into the condensate water downstream of the condensate polisher.

To reduce the general corrosion and FAC rate of ferrous alloys, a volatile pH adjustment chemical is injected to maintain a non-corrosive environment. Feedwater pH of 9.2 or more provides sufficient iron reduction effect.

Hydrazine (or an equivalent oxygen scavenger) is added to scavenge the dissolved oxygen and reduce it within the specified limits in the feedwater for each mode of operation.

10.3.5.5 Action Levels for Abnormal Conditions Appropriate responses to abnormal chemistry conditions provide for the long-term integrity of the secondary cycle components. Remedial actions are taken when chemistry parameters are outside normal operating ranges.

Secondary side water chemistry guidelines are provided in Table 10.3.5-1.The COL applicant will provide secondary side water chemistry threshold values and recommended operator actions for chemistry excursions, or provide a commitment to the latest version of the EPRI PWR Secondary Water Chemistry Guidelines in effect at the time of COLA submittal.

10.3.5.6 Lay Up and Heatup US-APWR anticipates no long-term SG layup under dry conditions. When inspection or Tier 2 10.3-15 Revision 23

10. STEAM AND US-APWR Design Control Document POWER CONVERSION SYSTEM maintenance is required on the secondary side, the SGs are drained hot water under a nitrogen atmosphere. After cooling, the nitrogen is purged and inspection/maintenance is performed.

Wet layup conditions are established for corrosion protection during outages.

Guidelines for this are given in Table 10.3.5-2.

The bulk water in the SGs is generally brought into power operation specifications before heatup to full power. This is done by either draining and refilling or feeding and bleeding.

Guidelines for heatup are provided in Table 10.3.5-3.

10.3.5.7 Chemical Analysis Basis Guidelines for control and diagnostic parameters for chemicals important for corrosion control in feedwater and SGs are listed in Table 10.3.5-1. Each chemical's impact is discussed below:

• Oxygen in the presence of moisture rapidly corrodes carbon steel. The resulting corrosion products may be carried through the FWS and form a sludge pile in the SGs. This sludge creates an ideal environment for localized corrosion on SG tubes. Thus the oxygen concentration should be kept as low as practical in the feedwater system. Dissolved oxygen is controlled at the condenser and deaerating feedwater heater to prevent oxygen transport in the FWS.

• The oxygen concentration is measured by process analyzers and by grab samples and is used as input for the oxygen scavenger injection.

• In the absence of significant impurities, the pH is controlled by the concentration of the volatile pH adjustment chemical and the oxygen scavenger. Maintaining pH within the recommended band results in minimal ferrous material corrosion rates.

The pH is measured in both process and bench instruments.

• Cation conductivity is a measure of the presence of ionic contamination and provisions are made for monitoring conductivity in samples from the condensate, feedwater and the SG blowdown.

• Sodium is an effective indicator of many forms of contaminant ingress. Sodium is measured by process analyzers and is capable of tracing the chemical at sub ppb level. Increased sodium levels are indicative of condenser tube leakage or makeup water contamination.

• Chloride is aggressive to ferrous materials at steam generator operating conditions. It has also been identified to be relevant to inconel 600 pitting.

• Sulfate causes acidic environment of SG crevice pH.

10.3.5.8 Sampling Samples are taken from the condenser hotwell, demineralized water, condensate, Tier 2 10.3-16 Revision 23

10. STEAM AND US- APWR Design Control Document POWER CONVERSION SYSTEM Table 10.3.5-1 Guidelines for Secondary Side Water Chemistry during Power Operation (Sheet 1 of 4)

- Condensate -

Parameter Control Value Control Dissolved Oxygen, ppb(1) 10.0 Note:

1. The condenser serves as an on-stream vacuum region. The possible evolution of oxygen due to makeup water inflow calls for condensate water monitoring to control dissolved oxygen (DO) values below 10 ppb.

Tier 2 10.3-36 Revision 23

10. STEAM AND US- APWR Design Control Document POWER CONVERSION SYSTEM Table 10.3.5-1 Guidelines for Secondary Side Water Chemistry during Power Operation (Sheet 2 of 4)

- Feedwater -

Parameter Control Value Control Hydrazine, ppb(1) 50(2)

Total iron, ppb 5.0(3)

Total copper, ppb 1.0(4)

Dissolved Oxygen, ppb 5.0(5)

Diagnostic pH at 25°C(1) 9.2(6)

Cation Conductivity due to strong acid anions @ (7) 25°C, µS/cm Metal oxide species ECP (8)

Integrated Corrosion Product Transport (9)

Lead (10)

Notes:

1. Feedwater Electrical Corrosion Potential (ECP) measurement suggests that low ECP in SG can be achieved with a feedwater hydrazine/condensate oxygen ratio of > 10. The feedwater hydrazine concentration is set to 50 ppb.

(feedwater hydrazine = condensate oxygen concentration x >10 5 ppb x >10

= 50 ppb).

2. Formation of oxidizing environment in secondary systems can be prevented at feedwater hydrazine concentrations of 50 ppb.

3. Scaling in SG and secondary system components can be controlled by maintaining iron concentrations as low as possible. Iron concentrations are dependent on the secondary system components materials and secondary water chemistry, however, iron concentrations can be maintained by appropriate AVT control.

Tier 2 10.3-37 Revision 23

10. STEAM AND US- APWR Design Control Document POWER CONVERSION SYSTEM Table 10.3.5-1 Guidelines for Secondary Side Water Chemistry during Power Operation (Sheet 3 of 4)

- Feedwater -

4. Copper acts as an oxidizing agent; it is desirable to decrease copper transport to avoid damage to SG tubes.

5. Dissolved oxygen concentrations should be minimized to avoid damage to SG tubes, and a control value of 5ppb based on detection limit from routine analysis must be maintained.

6. To control FAC, feedwater pH of 9.2 is set as a control value which effects a reduction in iron levels.

7. Cation conductivity is a semi-quantitative indicator of organic acid concentrations. Since chemical control during construction varies with plant, diagnostic value is not set up.

8. Setting up of ECP analyzers depend on plant requirements hence, no diagnostic value is set.

9. Periodic assessment of corrosion products mass transport to steam generators is performed using integrated samples.

10. Since chemical control during construction varies with plant, no diagnostic value is set.

Tier 2 10.3-38 Revision 23

10. STEAM AND US- APWR Design Control Document POWER CONVERSION SYSTEM Table 10.3.5-1 Guidelines for Secondary Side Water Chemistry during Power Operation (Sheet 4 of 4)

- Steam Generator Blowdown -

Parameter Control Value Control Sodium, ppb 5.0(1)

Chloride, ppb 10.0(2)

Sulfate, ppb 10.0(3)

Diagnostic Crevice pHt 5-1110(4)

Notes:

1. There are several environments that can cause damage of SG tube, based on the SG corrosion susceptibility diagram for alloys 600MA, 600TT and 690TT (Reference 10.3-21). According to evaluation results for the SG crevice environment estimation code, when the sodium concentration increases and crevice concentration factor is 107, the SG tube crevice pH gradually increases and exceeds 10 at sodium concentrations of more than 5 ppb

2. Chloride causes pitting of SG tube material when present with oxidizing materials. The control value is set at 10 ppb and is 1/10 of the 100ppb standard dose and does not have significant influence on the SG tube material.

3. Sulfate causes an acidic environment of crevice pH.

4. SG tube materials are damaged at alkaline environments of pH 10 and over and acidic environments of pH 4 and below and temperatures at 300°C. The electric potential of the SG tube material increases due to oxidizing materials and may damage susceptibility range. Surveillance and management of the environment of the oxidizing agent carried over in the SG tube crevice is difficult.

Therefore crevice pH during operation is evaluated by calculating the concentration of impurities to determine that they are within the recommended range of 5 to 1110.

Tier 2 10.3-39 Revision 23

10. STEAM AND US- APWR Design Control Document POWER CONVERSION SYSTEM Table 10.3.5-2 Guidelines for Steam Generator Water Chemistry during Cold Shutdown/Wet Layup Parameter Control Value Prior to Heatup (

200°F)

Control pH at 25°C 9.5(1) -

Hydrazine, ppm(2) 75 -

Sodium, ppb 1000(3) 100 Chloride, ppb 1000(4) 100 Sulphate, ppb 1000(5) 100 Diagnostic Dissolved O2, ppb - 100(6)

Notes:

SG water pH exceeding 9.5 indicates that a sufficient amount of hydrazine is present in SG water to form a protective oxide film.

2. Values apply if hydrazine is used for oxygen scavenging. An alternate oxygen scavenger may be used with appropriate concentration limits, which can keep reducing atmosphere in secondary system.

3. Control value of 1000 ppb or less for sodium concentration is maintained in order to quickly reach concentration values of 100 ppb or less during plant heatup.

4. A chloride concentration of 1000 ppb is maintained in order to quickly reach a concentration of 100 ppb or less during plant heatup.

5. A sulphate concentration of 1000 ppb is maintained in order to quickly reach a concentration of 100 ppb during plant heatup.

6. Dissolved oxygen control is required prior to and/or during the water fill/makeup phase. Appropriate compensatory actions should be taken to minimize SG dissolved oxygen content, e.g. addition of oxygen scavenging to water source, in plants without DO control.

Tier 2 10.3-40 Revision 23

10. STEAM AND US- APWR Design Control Document POWER CONVERSION SYSTEM Table 10.3.5-3 Guidelines for Secondary Side Water Chemistry during Heatup (Sheet 1 of 2)

Feedwater ( > 200°F To < 30% Power)

Parameter Control Value Control Value Prior To Power Under 30 %

Dissolved Oxygen, ppb 10.0(1)

Notes:

1. To avoid damage to the SG tubes, it is desirable to minimize oxygen carry over into the SG as much as possible. The control value is set up less than 10 ppb which can usually be attained during plant startup.

Tier 2 10.3-41 Revision 23

10. STEAM AND US- APWR Design Control Document POWER CONVERSION SYSTEM Table 10.3.5-3 Guidelines for Secondary Side Water Chemistry during Heatup (Sheet 2 of 2)

Steam Generator Blowdown ( > 200°F To < 30% Power)

Parameter Control Value Control Value Prior To Power Escalation Under 30 %

Sodium, ppb 50.0(1)

Chloride, ppb 100.0(1)

Sulfate, ppb 100.0(1)

Diagnostic Total Cation Conductivity @ 25°C, 2.0(2)

µS/cm Notes:

1. Listed parameters shall be verified to be within their respective ranges within a reasonable time period prior to escalation above 30 percent power.

2. 100 ppb of chloride concentration, which is a control value of plant startup, is equivalent to 2 µS/cm of total cation conductivity.

Tier 2 10.3-42 Revision 23

10. STEAM AND US- APWR Design Control Document POWER CONVERSION SYSTEM 10.4.6 Condensate Polishing System The condensate polishing system (CPS) is designed to remove dissolved ionic solids and impurities from the condensate. The CPS provides condensate cleanup capability and maintains condensate quality, on as needed basis, through demineralization.

10.4.6.1 Design Bases 10.4.6.1.1 Safety Design Bases The CPS does not serve any safety-related function, and, thus, has no safety design bases.

10.4.6.1.2 Non safety Power Generation Design Bases

• The CPS is designed to remove dissolved ionic solids and impurities from the condensate and assists in the removal of corrosion products.

• With a condenser tube leak of 0.001 gpm CPS is designed to assist normal continuous plant operation until repairs can be made.

• With a condenser tube faulted leak of 0.1 gpm, the CPS is designed to maintain plant operation until an orderly shutdown is achieved.

• The CPS is in a side stream arrangement, and processes one-third of the rated condensate flow during start up and clean up (impurities removal) of condensate .

10.4.6.2 System Description 10.4.6.2.1 General Description The CPS is designed with prefilters to remove corrosion products and with deep bed mixed resin vessels (demineralizers) to remove ionic impurities from the condensate during plant startup, hot standby, shutdown operations, and power operation.

Condensate polishing vessels and prefilters are installed in the 2nd floor of T/B.

The condensate bypass valve is located in the condensate pump discharge header to bypass condensate polishing vessels. The flow rates to condensate polishing vessel are controlled by the condensate bypass valve according to the requirements of the CPS.

The condensate polishing system is shown in Figure 10.4.6-1.

The requirements for the condensate purity of the CPS effluent are determined as shown in Table 10.4.6-2 to satisfy secondary side water chemistry guidelines for feedwater as described in Subsection 10.3.5.

10.4.6.2.2 Component Description Tier 2 10.4-29 Revision 23

10. STEAM AND US- APWR Design Control Document POWER CONVERSION SYSTEM Table 10.4.6-2 Condensate Purity Requirements in CPS Effluent Conductivity (mS/m @25°C) < 0.01 Total iron (ppb as Fe) <1 Total copper (ppb as Cu) <1 Dissolved silicate (ppb as SiO2) < 10 Sodium (ppb as Na) < 0.06 Chloride (ppb as Cl) < 0.15 Tier 2 10.4-35 Revision 23

US-APWR DCD Revision 2 RAI Tracking Report MUAP-09026(R5)

Chapter 11 Mitsubishi Heavy Industries, LTD.

US-APWR DCD Chapter 11 Rev. 2, Tracking Report Rev. 5 Change List Location (e.g., subsection with Page paragraph/ sentence/ Description of Change item, table with row/column, or figure)

RAI 629-4973, Question 11.03-18 11.2-2 Section 11.2.1.2 Add the description about IE bulletin 80-10.

RAI 532-4019, Question 12.02-29 11.2-12 Section 11.2.2.1.2.3 Added the description about RWSAT and RWSP.

RAI 624-4972, Question 11.02-33 11.2-16 Section 11.2.3.1 Corrected the grammatical error.

Added the description about PWR-GALE code version.

RAI 629-4973, Question 11.03-18 11.3-2 Section 11.3.1.2 Added the description about IE bulletin 80-10.

RAI 629-4973, Question 11.03-18 11.3-4 Section 11.3.1.4 Added the description about IE bulletin 80-10.

RAI 624-4972, Question 11.02-33 11.3-11 Section 11.3.3.1 Added the description about PWR-GALE code version.

RAI 629-4973, Question 11.03-18 11.3-18 Section 11.3.8 Added the description about IE bulletin 80-10 as reference 11.3-27.

RAI 629-4973, Question 11.03-18 11.3-25 table 11.3-4 Added the description about holdup time.

RAI 629-4973, Question 11.03-18 11.4-2 Section 11.4.1.2 Added the description about IE bulletin 80-10.

11. RADIOACTIVE WASTE MANAGEMENT US-APWR Design Control Document leakage levels (i.e., leakage from fuel producing 1% of the reactor thermal power level). The processing capabilities are such that the operation of the plant will not be impaired under these conditions.

• The LWMS is designed so that no potentially radioactive liquids can be discharged to the environment unless they have first been monitored and confirmed to be within acceptable limits. Offsite radiation doses measured on an annual basis will be within the limits of 10 CFR 20 (Ref. 11.2-1) and 10 CFR 50, Appendix I (Ref. 11.2-2).

• The LWMS has cross-connections, adequate storage capabilities and the ability to connect to and return from mobile systems to accommodate anticipated waste surge volumes.

• Interconnections between the LWMS and other plant systems are designed so that contamination of non-radioactive systems are precluded and the potential for uncontrolled and unmonitored releases of radiation to the environment from a single failure are minimized. This feature meets the requirements of IE bulletin 80-10 (Ref. 11.2-25).

• Design features minimize maintenance, equipment downtime, and leakage of radioactive liquid into the building atmosphere. Table 11.2-1 details the equipment codes for design and construction as required in Table 1 of RG 1.143 (Ref. 11.2-3). The Equipment Class 6 components are designed in compliance with applicable codes and standards, and guidelines provided in RG 1.143 (Ref.

11.2-3).

• The waste collection and monitor tanks are provided with an overflow connection at least as large as the inlet. The location of the overflow is above the high-level alarm setpoint. Each cell housing these tanks is coated with an impermeable epoxy liner (coating), up to the cubicle wall height equivalent to the full tank volume, to facilitate decontamination of the facility in the event of tank leakage and failure. This design feature, in conjunction with early leak detection, drainage and transfer capabilities, serves to minimize the release of the radioactive liquid to the groundwater and environment in accordance with the BTP 11-6 (Ref.11.2-17) and 10 CFR 20.1406 (Ref.11.2-7).

• The LWMS tanks are provided with a vent piping connected to the heating, ventilation, and air conditioning (HVAC) system. (See Chapter 9, Section 9.4) with the exception of the containment vessel reactor coolant drain tank (CVDT),

which is routed to the vent header in the gaseous waste management system (GWMS).

• The LWMS is designed in compliance with the as low as reasonable achievable (ALARA) principle for occupational doses. Sufficient shielding is provided for all equipment located in the radiological controlled area (RCA) that could cause unacceptable radiation doses.

Tier 2 11.2-2 Revision 23

11. RADIOACTIVE WASTE MANAGEMENT US-APWR Design Control Document instrumentation to monitor the temperature. Prior to transferring the water to the holdup tank (HT) in CVCS via one of two containment vessel reactor coolant drain pumps, the water temperature is decreased below 200 °F by the addition of PMW. The tank is generally maintained at a near constant level to minimize both the amount of gas sent to the GWMS and the amount of nitrogen cover gas required.

11.2.2.1.2.2 Other Anticipated Operations In the event that the liquid collected in the CVDT is either oxygenated or above the specified radiation limits, it is sent to the WHTs for processing.

11.2.2.1.2.3 Maintenance/Refueling Operations During refueling, the containment vessel reactor coolant drain pumps are used to drain water from the reactor coolant loops and the emergency core cooling system ACCsthe reactor cavity and the fuel transfer canal to the refueling water storage auxiliary tank (RWSAT) while the drain water from the refueling cavity is directly sent to the refueling water storage pit (RWSP) by the CS/RHR pumps or gravity. In this case, typically both pumps are used to speed up the transfer of water from these areas. In this mode, the water is transferred directly to the RWSAT without entering the CVDT. During maintenance or outages, any remaining gas is purged from the system to the GWMS using nitrogen.

Recyclable reactor-grade effluents enter this subsystem from various locations inside the containment and are collected without exposure to air (which would degrade the quality of the water). These liquids are collected in the CVDT which is situated inside the containment. The contents of the tank are maintained in a nitrogen-rich environment to minimize the potential for degradation of the water quality and minimize the potential for formation of a flammable mixture. The tank is a cylindrical stainless steel vessel and oriented horizontally. The tank is vented to the GWMS. The tank is equipped with a relief valve, which vents to the containment sump. The purge water head tank in CVCS shares the same vapor space with the CVDT.

The liquid level and temperature within the tank are monitored. The liquid temperature is maintained below 200 °F by the addition of PMW, as necessary, in order to minimize the potential for damage to downstream equipment, including diaphragm valves. The liquid is transferred via one of two reactor coolant drainage system pumps to the CVCS HT.

The liquid can also be routed to the WHT and the RWSAT depending on the quality of the water and plant conditions.

The reactor coolant drainage system pumps are also used for transferring water from the reactor cavity, the emergency core cooling system ACCs, and the fuel transfer canal to the RWSAT after refueling operations.

Tier 2 11.2-12 Revision 23

11. RADIOACTIVE WASTE MANAGEMENT US-APWR Design Control Document particular equipment. The decontamination factors are taken from NUREG-0017 Rev.1(Ref.11.2-13) and are presented in Table 11.2-7. The calculations are made based on the assumption as follows:

• The liquid effluent from the primary system is processed by the LWMS and is discharged without reuse.

• The steam generator blowdown is treated and returned to the condenser.

The release physical location and configuration of the treated effluent is site specific.

Detailed design information such as release point, effluent temperature and flow rate, and size and shape of flow orifices, is to be presented in the site specific detail design.

The COL applicant is responsible for ensuring that the site-specific information of the LWMS, e.g., radioactive release points, effluent temperature, shape of flow orifices, etc.,

is to be provided in the COLA(COL11.2(2)).

The annual average release of radionuclides is estimated by the PWR-GALE Code (Ref.11.2-13) with the reactor coolant activities that is are described in Section 11.1.

The version of the code is a proprietary modified version of the NRC PWR-GALE code reflecting the design specificities of US-APWR design. The parameters used by the PWR-GALE Code are provided in Table 11.2-9, and the calculated effluents are provided in Table 11.2-10. The calculated effluents for the maximum releases are provided in Table 11.2-11.

The calculated effluent concentrations using annual release rates are then compared against the concentration limits of 10 CFR 20 Appendix B (Ref. 11.2-8); see Table 11.2-12 and Table 11.2-13.

The calculation uses 12,900 gpm cooling tower blowdown as dilution water (See Chapter 10, Subsection 10.4.5). The ratios to the concentration limits of 10 CFR 20 Appendix B (Ref 11.2-8) are 8.10E-02 (with expected releases) and 3.09E-01 (with maximum defined fuel defects), and these values are less than the allowable value of 1.0.

The individual doses are evaluated with LADTAP II Code (Ref. 11.2-14). The parameters used in the LADTAP II Code are listed in Table 11.2-14, and calculated doses are listed in Table 11.2-15. Based on these parameters, the dose to total body is 1.98 mrem/yr(Child) and the dose to organ is 2.54 mrem/yr (Childs liver). These values are less than the criteria of 3 mrem/y and 10 mrem/yr, respectively, as specified in 10 CFR 50 Appendix I (Ref. 11.2-2).

The COL applicant is to calculate doses to members of the public following the guidance of RG 1.109 (Ref 11.2-15) and RG 1.113 using site-specific parameters, and compares the doses due to the liquid effluents with the numerical design objectives of Appendix I to 10 CFR 50 (Ref 11.2-10) and compliance with requirements of 10 CFR 20.1302, 40 CFR 190.

Tier 2 11.2-16 Revision 23

11. RADIOACTIVE WASTE MANAGEMENT US-APWR Design Control Document

• Interconnections between the GWMS and other plant systems are designed so that contaminations of non-radioactive systems are precluded and the potential for uncontrolled and unmonitored releases of radiation to the environment from a single failure are minimized. This feature meets the requirements of IE bulletin 80-10(Ref. 11.3-27).

• The GWMS is designed to provide redundancy and storage capacity to process anticipated surge volumes due to equipment downtime and maintenance activities.

• The GWMS design is in compliance with ANSI/ANS-55.4 (Ref. 11.3-6).

• The GWMS is designed to comply with the ALARA principle for occupational doses. Shielding is provided to the equipment cubicles for the equipment containing design basis source terms to keep doses to personnel ALARA.

• The GWMS is designed to meet the requirements of GDC 60, 61, and 64 and the guidance of RG 1.143 (Ref. 11.3-2) so that waste gases are successfully processed. The GWMS includes radiation monitors, which continuously monitor the effluents prior to release into the environment.

• In accordance with ANSI/ANS-55.4 (Ref. 11.3-6), the A/B that houses the GWMS equipment is designed to withstand the effect of OBE.

• The GWMS equipment, piping, monitors, and controls are in compliance with ANSI/ANS 55.4 (Ref. 11.3-6) and RG 1.143 (Ref. 11.3-2). The GWMS is provided with hydrogen and oxygen monitors so that a lower explosive limit is not reached.

• The GWMS is designed to process the waste gases generated from normal operation and including AOOs. Stored gas that is not reused is treated and is discharged, provided that the gas meets release criteria.

• Streams in the GWMS are monitored for both hydrogen and oxygen content so that a flammable mixture is not reached. This feature complies with 10 CFR 50 Appendix A, GDC 3 (Ref. 11.3-7) and RG 1.189 (Ref. 11.3-8).

• The GWMS is designed so that the average annual dose at the site boundary does not exceed the limits of 10 CFR 50, Appendix I (Ref. 11.3-3) during normal operation including AOOs.

• The GWMS equipment is designed, located, and shielded in accordance with RG 8.8 (Ref. 11.3-9).

11.3.1.3 Other Design Considerations In addition to the listed design criteria, the following consideration is satisfied:

Tier 2 11.3-2 Revision 23

11. RADIOACTIVE WASTE MANAGEMENT US-APWR Design Control Document The GWMS uses equipment that is commonly used in the nuclear power industry. Their performances are proven and documented. The equipment is sized to process waste gases using design source term and design conditions that bound normal operation including AOOs. The equipment is also housed in the A/B with sufficient shielding such that the average annual dose at the site boundary from direct radiation from the gaseous sources does not exceed the limits of 10 CFR 50, Appendix I (Ref. 11.3-3). Charcoal beds significantly remove and reduce the fraction of radioactive iodine in the effluent stream. Noble gases can be stored in the gas surge tank or HTs to allow decay prior to release.

GWMS equipment is designed, located, and shielded to comply with the guidance of RG 8.8 (Ref. 11.3-9), thus maintaining occupational doses ALARA.

The GWMS includes radiation monitoring to continuously measure the radioactivity in the effluent stream prior to release into the environment to comply with the requirements of GDC 60 and 64. Additional and redundant radiation monitors are provided in the vent stack to verify the radiation level. Upon detection of radiation levels above the setpoint, the monitor activates an alarm and sends signals to close the GWMS discharge valves.

The GWMS is designed so that interconnection between plant systems precludes the contamination of non-radioactive systems and uncontrolled releases of radioactivity to the environment to meet the requirements of IE bulletin 80-10 (Ref. 11.3-27). At least two isolation valves are located between the clean and contaminated systems to minimize the potential for contamination of clean systems. This feature meets the requirements of 10 CFR 20.1406 (Ref. 11.3-10).

The design standards and materials for all GWMS structures, systems, and components (SSCs) are consistent with the specifications provided in RG 1.143 (Ref. 11.3-2). A list of the major equipment for the GWMS, including the number of units supplied, rates, process conditions, materials, and relevant design codes, is provided in Tables 11.3-2 and Figure 11.2-1 (sheet 3 of 3).

11.3.1.5 Site-Specific Cost-Benefit Analysis The GWMS is designed to be used for any site. This report provides the justification that the design is flexible so that site-specific requirements such as preference and upgrade of technologies, degree of automated operation, and radioactive waste storage, can be incorporated into the design with minor modifications.

RG1.110 provides compliance with 10 CFR 50, Appendix I (Ref. 11.3-3) numerical guidelines for offsite radiation doses as a result of gaseous or airborne radioactive effluents during normal operations, including AOOs. The cost-benefit numerical analysis as required by 10 CFR 50, Appendix I (Ref. 11.3-3),Section II, Paragraph D demonstrates that the addition of items of reasonably demonstrated technology is not favorable or cost beneficial.

The COL applicant is to perform a site-specific cost-benefit analysis to demonstrate compliance with the regulatory requirements.

Tier 2 11.3-4 Revision 23

11. RADIOACTIVE WASTE MANAGEMENT US-APWR Design Control Document

• Only proven and qualified equipment from the nuclear industry is used.

• Steel piping with butt welded construction is used to minimize crud traps. Only qualified welders are employed.

• Cubicles containing radioactive liquid are steel-lined or lined with an epoxy coating to minimize the potential for contamination to the groundwater system and to ease maintenance and decontamination.

• Drains and overflows are routed directly to sumps to minimize the spread of radioactive fluid.

• Non-radioactive auxiliary subsystems are isolated from the radioactive process streams.

In order to reduce leakage and releases of radioactive material, the following design features are included:

• Equipment, piping, and instruments are subject to stricter leak rate testing and inspections

• Sumps are equipped with level switches that activate alarms for prompt operator action to minimize the spread of contamination 11.3.3 Radioactive Effluent Releases 11.3.3.1 Radioactive Effluent Releases and Dose Calculation in Normal Operation The GWMS treats and releases radioactive gaseous waste generated from normal operation, including AOOs. Gaseous release data (isotope and activity) are presented in Tables 11.3-5 through 11.3-7. There are no liquid or solid waste releases from the GWMS.

The GWMS is designed to treat potentially radioactive gas to meet the concentration and dose limits of 10 CFR 20 (Ref. 11.3-4), the dose limits of 10 CFR 50, Appendix I (Ref.

11.3-3), and GDC 64. The main sources of plant radioactive gaseous inputs to the GWMS are the waste gases from the VCT, CVDT, boric acid evaporator, and HTs.

Their flow rates are presented in Figure 11.3-1 (Sheet 3 of 3).

The treated gaseous waste is further diluted by HVAC ventilation flow before the gases are released from the vent stack. The vent stack runs alongside the containment with the release point above the top of the containment. The design information of the vent stack and release point is described in Section 11.3.2.

The release rates and isotopic compositions are calculated using the PWR-GALE Code, NUREG-0017 (Ref. 11.3-1). The version of the code is a proprietary modified version of the NRC PWR-GALE code reflecting the design specificities of US-APWR design. Other parameters for the PWR-GALE Code calculation are listed in Section 11.1 and Section Tier 2 11.3-11 Revision 23

11. RADIOACTIVE WASTE MANAGEMENT US-APWR Design Control Document 11.3-19 Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I, Regulatory Guide 1.109, Rev. 1, October 1977.

11.3-20 Offsite Dose Calculation Manual Guidance: Standard Radiological Effluent Controls for Pressurized Water Reactors. [This NUREG includes Generic Letter 89-01.], NUREG-1301.

11.3-21 U.S. Nuclear Regulatory Commission, Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants. NUREG-0133, Washington, DC.

11.3-22 Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors. Regulatory Guide 1.111, Rev. 1,July 1977.

11.3-23 Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases for the Purpose of Implementing Appendix I. Regulatory Guide 1.113, Rev. 1, April 1977.

11.3-24 Compliance with dose limits for individual members of the public. NRC Regulations Title 10, Code of Federal Regulations, 10 CFR Part 20.1302.

11.3-25 U.S. Environmental Protection Agency, Environmental Radiation Protection Standards for Nuclear Power Operations, Protection of Environment. Title 40, Code of Federal Regulations, Part 190, Washington, DC.

11.3-26 Cost-Benefit Analysis for Radwaste Systems for Light-Water-Cooled Nuclear Power Reactors. Regulatory Guide 1.110, March 1976.

11.3-27 Contamination of Nonradioactive System and Resulting Potential for Unmonitored, Uncontrolled Release of Radioactivity to Environment, U.S.

Nuclear Regulatory Commission, IE Bulletin No. 80-10, May 6, 1980.

Tier 2 11.3-18 Revision 23

11. RADIOACTIVE WASTE MANAGEMENT US-APWR Design Control Document Table 11.3-4 Input Parameters and Calculation Results of Radioactive Effluent Releases and Dose due to the Gaseous Waste Management System Leak or Failures (Sheet 2 of 2)

I. Input Parameters for the Charcoal Bed Leak Reactor coolant activity(Ci/g)(1) Kr-85m 8.15E-01 Kr-85 4.22E+01 Kr-87 5.30E-01 Kr-88 1.52E+00 Xe-133 1.43E+02 Xe-135 4.69E+00 Holdup time for Xe (days) (3) 0.02 Holdup time for Kr (days) (3) 0.02 Other parameters Same as Table 11.2-9 Annual effluent of noble gas (Ci/yr) Kr-85m 3.58E+03 Kr-85 4.10E+05 Kr-87 1.85E+03 Kr-88 6.25E+03 Xe-133 6.70E+05 Xe-135 2.17E+04 Dose factor for total body RG 1.109 X/Q(s/m3)(2) 5.0E-4 II. Result and Criteria Calculated dose(mrem) 2 Dose criteria(mrem) 100 Note:

1. Based on a 300 Ci/g Dose Equivalent Xe-133 as stated in the technical specifications

2. The short-term X/Q at EAB (See Section 2.3.4)

3. Based on BTP 11-5 (Ref. 11.3-18). Other parameters for GALE code calculation are the same as those described in table 11.2-9.

Tier 2 11.3-25 Revision 23

11. RADIOACTIVE WASTE MANAGEMENT US-APWR Design Control Document be fully loaded with suspended solids and dissolved solids using the design basis source term, which is also used to determine the thicknesses of the shield walls.

• The SRSTs are cross-connected so that the failure or maintenance of one component does not impair the system or the plant operation. Table 11.4-5 provides typical failure scenarios. The spent resin storage tanks (SRSTs) are housed in individual cubicles, each with a shield wall thickness commensurate with the projected maximum dose rate of its content. The cubicles that contain significant quantities of radioactive material are coated with an impermeable epoxy liner (coating), up to the cubicle wall height equivalent to the full tank volume, to facilitate decontamination of the facility in the event of tank leakage and failure. This design feature, in conjunction with early leak detection, drainage and transfer capabilities, serves to minimize the release of the radioactive liquid to the groundwater and environment in accordance with the BTP 11-6 (Ref. 11.4-32) and 10 CFR 20.1406 (Ref. 11.4-16). As an additional precaution, the COL Applicant is also required to provide an environmental monitoring system (Section 11.5.5). Other design features addressing release requirements are described in Section 11.2

• Interconnections between the SWMS and other plant systems are designed so that contaminations of non-radioactive systems are precluded and the potential for uncontrolled and unmonitored releases of radiation to the environment from a single failure are minimized. This feature meets the requirements of IE bulletin 80-10(Ref.11.4-29).

• Storage is provided to facilitate radioactive decay of spent resin and to provide adequate holding in the case of a delay in processing due to maintenance and/or a delay in transportation for disposal.

• Any liquids and gases generated from the operation of the SWMS are processed by the LWMS (Section 11.2) and plant ventilation system (Chapter 9, Section 9.4). Based on typical PWR experience, a small quantity of sludge and oily waste is expected to be generated. Sludge is stabilized and transported to a disposal facility. Oily waste is collected and sent to appropriately licensed offsite vendors for processing and disposal.

• Collection, processing, packaging, and storage of radioactive wastes is performed to maintain any potential radiation dose to plant personnel ALARA in accordance with RG 8.8 (Ref. 11.4-2) and within the limits of 10 CFR 20 (Ref. 11.4-3). Some of the design features incorporated to maintain exposure levels ALARA include remote system operation and remotely actuated flushing, and an equipment layout that shields personnel from components containing radioactive materials.

• The SWMS is designed to package radioactive wastes in accordance with 10 CFR 61 (Ref. 11.4-4) and the applicable portions of 10 CFR 60 (Ref 11.4-5) and 10 CFR 63 (Ref. 11.4-6). The containers meet the requirement of 49 CFR 171 (Ref. 11.4-7). Solid wastes are processed and packaged for transportation Tier 2 11.4-2 Revision 23

US-APWR DCD Revision 2 RAI Tracking Report MUAP-09026(R5)

Chapter 12 Mitsubishi Heavy Industries, LTD.

US-APWR DCD Chapter 12 Rev. 2, Tracking Report Rev. 5 Change List Location (e.g., subsection with Page paragraph/sentence/ Description of Change Item, table with column/row, or figure) 12.3-2 12.3.1.1.1.1 Replaced The cobalt content is controlled to be not more than 0.016 mass percent for the US-APWR steam D. SGs generator tubing. with The cobalt content is controlled to be not more than 0.016 mass percent and an average of 0.014 mass percent for the US-APWR steam generator tubing..

RAI No.524-4020 Revision 1 12.3-2 12.3.1.1.1.2 Added a new paragraph which reads: The following subsections describe the general design criteria for several types of plant components. The materials selection criteria described in Subsection 12.3.1.1.1.1 E are also applicable to the various types of equipment described below. The component specifications for the procurement of components for specific applications will be based on the latest relevant national and international industry guidance in order to ensure consistency with regulatory requirements related to ALARA best practices, contamination minimization, and component reliability needs..

RAI No.524-4020 Revision 1 12.3-3 12.3.1.1.1.2 Added a new paragraph which reads: Pumps used for radioactive fluids should utilize seal cartridges, as D. Pumps applicable, to minimize field repair time, reduce exposure times, minimize errors in reassembly, and reduce the consequent leakage potential..

RAI No.524-4020 Revision 1 12.3-4 12.3.1.1.1.2 Added a new last sentence which reads: Tanks containing radioactive particulate material shall include E. Tanks one or more of the features mentioned below:

Purification of radioactive fluids (filtration or ion exchange) prior to entering the tank Sloped or cone-shaped tank bottom Grinding of internal welds Tank flushing capability Agitation by recirculation flow capability Lancing or chemical cleaning capability.

RAI No.524-4020 Revision 1 Page 1 of 3

US-APWR DCD Chapter 12 Rev. 2, Tracking Report Rev. 5 Change List Location (e.g., subsection with Page paragraph/sentence/ Description of Change Item, table with column/row, or figure) 12.3-6 12.3.1.1.1.2 Replaced Metal diaphragm or bellows seat valves are used on those systems where essentially no leakage can H. Valves be tolerated. with Metal diaphragm or bellows seat valves are used on those systems where essentially no leakage can be tolerated such as hazardous gas supply lines or gaseous radioactive waste systems. However, rubber diaphragm valves shall not be used for modulating the flow. Diaphragm valves are only used for isolating the line during maintenance of connected components. Valves 2 inches and under in diameter, except for modulating valves, located in piping carrying radioactive fluids are hermetically sealed (packless),

metal diaphragm or bellows seat valves to preclude radioactive releases to the environment. Valves greater than 2 inches in diameter and all modulating valves, regardless of size, shall include live-loaded packing and graphite packing materials, or leak off connections to reduce the potential for radioactive liquid or gas leakage.

Live-loaded packing shall be used for valves that include cyclic service conditions or critical valves that cannot be retorqued during plant operation..

RAI No.524-4020 Revision 1 12.3-6 12.3.1.1.1.2 Add a new 8th paragraph which reads: Check valves will be used only where necessary. Check valves shall H. Valves be properly located and oriented within the piping system. The type and size of valve selected shall be compatible with the system requirements to minimize flow-induced disk flutter related wear damage..

RAI No.524-4020 Revision 1 12.3-7 12.3.1.1.1.2 Add a new last sentence which reads: Provisions are made in radioactive systems for flushing the piping with I. Piping sufficient water to reduce crud buildup. Welds are made smooth, as much as possible, to prevent crud traps from forming..

RAI No.524-4020 Revision 1 12.3-7 12.3.1.1.1.2 Add a new 2nd paragraph which reads: Connections between piping, fittings, flanges and valves shall be I. Piping welded or flanged only. O-ring type pipe caps, O-ring face seal fittings and threaded joints are not used for radioactive piping..

RAI No.524-4020 Revision 1 12.3-53 Table 12.3-7 Replaced 0.016 with 0.016 (Average: 0.014).

2nd column, 3rd row RAI No.524-4020 Revision 1 Page 2 of 3

US-APWR DCD Chapter 12 Rev. 2, Tracking Report Rev. 5 Change List Location (e.g., subsection with Page paragraph/sentence/ Description of Change Item, table with column/row, or figure) 12.3-53 Table 12.3-7 Replaced 0.10 with 0.05.

2nd column, 6th row RAI No.524-4020 Revision 1 12.3-53 Table 12.3-7 Replaced 0.20 with 0.15.

2nd column, 7th row RAI No.524-4020 Revision 1 12.3-131 Figure 12.3-1 (Sheet Replaced the figures.

thru 12.3- 15 of 34 ~ 18 of 34)

RAI No.524-4020 Revision 1 134 12.3-161 Figure 12.3-3 (Sheet Replaced the figures.

thru 12.3- 1 of 10 ~ 4 of 10)

RAI No.524-4020 Revision 1 164 12.3-171 Figure 12.3-4 (Sheet Replaced the figures.

thru 12.3- 1 of 10 ~ 4 of 10)

RAI No.524-4020 Revision 1 174 12.3-181 Figure 12.3-5 (Sheet Replaced the figures.

thru 12.3- 1 of 10 ~ 4 of 10)

RAI No.524-4020 Revision 1 184 12.3-191 Figure 12.3-6 (Sheet Replaced the figures.

thru 12.3- 1 of 10 ~ 4 of 10)

RAI No.524-4020 Revision 1 194 12.3-205 Figure 12.3-11 (Sheet Replaced the figures.

thru 12.3- 1 of 10 ~ 4 of 10)

RAI No.524-4020 Revision 1 208 Page 3 of 3

12. RADIATION PROTECTION US-APWR Design Control Document C. Reactor Vessel Insulation Insulation, in the area of the reactor vessel nozzle welds, is fabricated in sections with a thin reflective metallic sheet covering and quick disconnect clasps to facilitate the removal of the insulation for the inspection of the welds.

D. SGs The SGs incorporate several design features to facilitate maintenance and inspection in reduced radiation fields. The SGs have the following design aspects:

1. Manways of the channel head are sized to facilitate access for tube bundle inspections and maintenance.

2. The channel head has a cylindrical region just below the tube sheet primary side to enhance the access of tooling to all tubes, including those on the periphery of the tube bundle.

3. Rapid entry/exit nozzle dam systems are provided in both primary nozzles to minimize occupational radiation exposure and to enhance personnel safety.

The specification of low cobalt tubing material for the US-APWR steam generator design is an important feature of the design; not only in terms of reduced exposure relative to the steam generator, but to the total plant radiation source term. The cobalt content is controlled to be not more than 0.016 mass percent and an average of 0.014 mass percent for the US-APWR steam generator tubetubing.

E. Materials Equipment specifications for components exposed to high temperature reactor coolant contain limitations on the cobalt content of the base metal as given in Table 12.3-7. The use of hard facing material with cobalt content such as stellite is limited to applications where its use is necessary for reliability considerations.

Nickel-based alloys in the reactor coolant system (Co-58 is produced from activation of Ni-58) are similarly used only where component reliability may be compromised by the use of other materials. The major use of nickel-based alloys in the reactor coolant system is the inconel steam generator tube.

12.3.1.1.1.2 Balance of Plant Equipment The following subsections describe the general design criteria for several types of plant components. The materials selection criteria described in Subsection 12.3.1.1.1.1 E are also applicable to the various types of equipment described below. The component specifications for the procurement of components for specific applications will be based on the latest relevant national and international industry guidance in order to ensure consistency with regulatory requirements related to ALARA best practices, contamination minimization, and component reliability needs.

Tier 2 12.3-2 Revision 23

12. RADIATION PROTECTION US-APWR Design Control Document A. Filters Filters that accumulate radioactivity are supplied with the means either to back-flush the filter remotely or to perform cartridge replacement with semi-remote tools.

For cartridge filters, adequate space is provided to allow removal, cask loading, and transportation of the cartridge to the solid radwaste area.

Back-flushable filters are designed so that the filter internals may be remotely removed and placed in a shielded cask for offsite shipping and disposal, in the unlikely event that a filter loses its back-flush capability.

Liquid systems containing radioactive cartridge filters are provided with a remote filter handling system for the removal of spent radioactive filter cartridges from their housings and for their transfer to the drumming station for packaging and shipment for burial. The process is accomplished so that exposure to personnel and the possibility of an inadvertent radioactive release to the environment are minimized. Each filter is contained in a shielded compartment and provided with vent and drain valving, and individual compartments have drainage capabilities. The filter handling system has also been designed with a minimum of components susceptible to malfunction.

B. Demineralizers Demineralizers for highly radioactive systems are designed so that spent resins can be remotely and hydraulically transferred to spent resin storage tanks so that fresh resin can be loaded into the demineralizer remotely. The demineralizers and piping are designed with the ability to be flushed with demineralized water. Strainers are installed in the vent lines to prevent the entry of spent resin into the exhaust duct.

C. Evaporators Adequate space and flanged connections for easy removal are provided for the maintenance of evaporator components. Additionally, the evaporator can be operated in an automatic operation mode that can reduce the exposure of the operator to radiation from the equipment.

D. Pumps Wherever practicable, pumps are sealed with mechanical seals to reduce seal servicing time. Pumps and associated piping are arranged to provide adequate space for access to the pumps for servicing. Small pumps are installed to allow easy removal, if necessary. All pumps in the radioactive waste systems are provided with flanged connections for ease of removal. Pump casings are provided with drain connections for draining pumps for maintenance.

Pumps used for radioactive fluids should utilize seal cartridges, as applicable, to minimize field repair time, reduce exposure times, minimize errors in reassembly, and reduce the consequent leakage potential.

Tier 2 12.3-3 Revision 23

12. RADIATION PROTECTION US-APWR Design Control Document E. Tanks Whenever practicable, tanks are provided with sloped bottoms and bottom outlet connections. Overflow lines are directed to the waste collection system to control any minimize the potential for the spread of contamination within plant structures. Tanks containing radioactive fluids are have overboard lines at least equal in size to the largest inlet line. The tank vent line is either equipped with open vents to the cubicle or connected directly to the ventilation system. The spent resin tank vents are equipped with a break-pot, which separates the air from the moisture and any entrained resin, which are subsequently sent to the A/B sump, and vents the air to the exhaust ductwork.

These measures minimize the possible contamination of the area and the ductwork.

Tanks containing radioactive particulate material shall have smooth welds and mixing, flushing and cleaning capabilities to prevent retention of the radioactive particulate material. Tanks containing radioactive particulate material shall include one or more of the features mentioned below:

• Purification of radioactive fluids (filtration or ion exchange) prior to entering the tank

• Sloped or cone-shaped tank bottom

• Grinding of internal welds

• Tank flushing capability

• Agitation by recirculation flow capability

• Lancing or chemical cleaning capability Tank cubicles are coated with non-porous material up to a wall height to contain the entire tank content. The cubicles are equipped with a drainage system to direct any leakage and overflows to sumps with pumps to redirect flow to other tanks. The drainage system is equipped with a liquid detection instrument which can provide early warning for leakage and/or overflow condition to initiate operator actions. The floors of these cubicles containing radioactive fluid are sloped to facilitate faster drainage and to minimize liquid accumulation, and provided with coating with non-porous material to prevent cross contamination.

F. Heat Exchangers Most of the heat exchangers are shell and tube type heat exchangers. These are provided with corrosion-resistant tubes of stainless steel or other suitable materials to minimize leakage. Impact baffles are provided and the tube side and shell side velocities are limited to minimize erosion effects. Wherever possible, the radioactive fluid passes through the tube side of the heat exchanger.

Some heat exchangers, specifically the SFP heat exchangers and the component cooling water heat exchangers, are plate-type heat exchangers constructed of austenitic stainless steel. For the SFP heat exchangers, the SFP water circulates through one Tier 2 12.3-4 Revision 23

12. RADIATION PROTECTION US-APWR Design Control Document Valves in the containment that are expected to exhibit stem leakage are provided with leak-off connections, piped to the reactor coolant drain tank (reactor coolant drain tank or a reduced packing configuration with the valve stem leak-off line capped).

Valves for clean, non-radioactive systems are separated from radioactive sources and are located in readily accessible areas.

For most large valves in lines carrying radioactive fluids, a double set of packing with a lantern ring is provided. A stuffing box with a leak-off connection that is piped to a drain header is also provided. Metal diaphragm or bellows seat valves are used on those systems where essentially no leakage can be tolerated such as hazardous gas supply lines or gaseous radioactive waste systems. However, rubber diaphragm valves shall not be used for modulating the flow. Diaphragm valves are only used for isolating the line during maintenance of connected components. Valves 2 inches and under in diameter, except for modulating valves, located in piping carrying radioactive fluids are hermetically sealed (packless), metal diaphragm or bellows seat valves to preclude radioactive releases to the environment. Valves greater than 2 inches in diameter and all modulating valves, regardless of size, shall include live-loaded packing and graphite packing materials, or leak off connections to reduce the potential for radioactive liquid or gas leakage. Live-loaded packing shall be used for valves that include cyclic service conditions or critical valves that cannot be retorqued during plant operation.

Check valves will be used only where necessary. Check valves shall be properly located and oriented within the piping system. The type and size of valve selected shall be compatible with the system requirements to minimize flow-induced disk flutter related wear damage.

Manually operated valves in the filter and demineralizer valve compartments required for normal operation and shutdown are equipped with reach rods extending through or over the valve gallery wall.

Personnel do not enter the valve gallery during spent resin or cartridge transfer operations. The valve gallery shield walls are designed to minimize personnel exposure during the maintenance of components within or adjacent to the gallery and to protect personnel who remotely operate the valves.

Relief valves are located in an associated equipment compartment or valve gallery.

Check valves are located in the equipment compartment or associated valve gallery unless they are the locking type requiring manipulation during normal operation. In this case, check valves are treated as normal manual valves.

I. Piping The piping in pipe chases is designed for the lifetime of the unit. Wherever radioactive piping is routed through areas where routine maintenance is required, pipe chases are provided to reduce the radiation contribution from these pipes to levels appropriate for the inspection or maintenance requirements. Butt welds are used to the fullest extent possible in radwaste piping utilized for the transport of spent resins or slurries. Piping containing radioactive material is routed to minimize radiation exposure to the unit Tier 2 12.3-6 Revision 23

12. RADIATION PROTECTION US-APWR Design Control Document personnel. Provisions are made in radioactive systems for flushing the piping with sufficient water to reduce crud buildup. Welds are made smooth, as much as possible, to prevent crud traps from forming.

Connections between piping, fittings, flanges and valves shall be welded or flanged only.

O-ring type pipe caps, O-ring face seal fittings and threaded joints are not used for radioactive piping.

J. Floor Drains Floor drains and properly sloped floors are provided for each room or cubicle containing serviceable components containing radioactive liquids. When practicable, shielded pipe chases are used for radioactive pipes. If a radioactive drain line must pass through a plant area requiring personnel access, shielding is provided, as necessary, to ensure that radiation levels are consistent with the required access.

K. Heating, Ventilation, and Air-Conditioning The HVAC system design facilitates the replacement of the filter elements.

L. Sample Stations Proper shielding and ventilation are provided at the local sample stations to maintain radiation zoning in proximate areas and minimize personnel exposure during sampling.

The use of concrete containing fly ash is minimized in the counting room and laboratory areas.

M. Clean Services Whenever possible, clean services and equipment such as compressed air piping, clean water piping, ventilation ducts, and cable trays are not routed through radioactive pipeways.

12.3.1.1.2 Common Facility and Layout Designs for As Low As Reasonably Achievable This subsection describes the design features utilized for the standard type plant process and layout situations. The features used in conjunction with the general equipment described in Subsection 12.3.1.1.1 are discussed in the following paragraphs.

A. Valve Galleries Valve galleries are provided with shielded entrances for personnel protection. Floor drains are provided to recover radioactive leakage. To facilitate decontamination in the valve galleries, concrete surfaces are covered with a smooth surface coating that allows easy decontamination.

Tier 2 12.3-7 Revision 23

12. RADIATION PROTECTION US-APWR Design Control Document Table 12.3-7 Equipment Specification Limits for Cobalt Impurity Levels Application Maximum Mass Percent of Cobalt Inconel and stainless steel components in 0.05 fuel assembly Inconel Tubing in Steam Generator 0.016 (Average: 0.014)

Components in region of high neutron flux such as Neutron Reflector and Lower 0.05 Core Barrel Divider Plate of Steam Generator and weld clad surfaces of Reactor Vessel, 0.05 Pressurizer and Channel Head of Steam Generator Upper Core Plate, Upper/Lower Core 0.100.05 Support Plate and Upper Core Barrel Main Coolant Piping, casings and internals of Rector Coolant Pumps and 0.200.15 Reactor Internals other than listed above Not limited (However, precipitation hardening stainless Bearing and hard-facing materials steel will be used for some valves exposed to severe depressurization conditions, and non-cobalt hard-facing material will be used for Reactor Coolant Pump.)

Auxiliary components such as valves except for listed above, piping instrumentation, tanks, and so on, Not limited including bolting materials in primary and auxiliary components Welding material, except where used as Not limited weld cladding Tier 2 12.3-53 Revision 23

12. RADIATION PROTECTION US-APWR Design Control Document Security-Related Information - Withheld Under 10 CFR 2.390 Figure 12.3-1 Radiation Zones for Normal Operation/Shutdown (Sheet 15 of 34)

Auxiliary Building at Elevation 4 Tier 2 12.3-131 Revision 23

12. RADIATION PROTECTION US-APWR Design Control Document Security-Related Information - Withheld Under 10 CFR 2.390 Figure 12.3-1 Radiation Zones for Normal Operation/Shutdown (Sheet 16 of 34)

Auxiliary Building at Elevation 7 Tier 2 12.3-132 Revision 23

12. RADIATION PROTECTION US-APWR Design Control Document Security-Related Information - Withheld Under 10 CFR 2.390 Figure 12.3-1 Radiation Zones for Normal Operation/Shutdown (Sheet 17 of 34)

Auxiliary Building at Elevation 3-7 Tier 2 12.3-133 Revision 23

12. RADIATION PROTECTION US-APWR Design Control Document Security-Related Information - Withheld Under 10 CFR 2.390 Figure 12.3-1 Radiation Zones for Normal Operation/Shutdown (Sheet 18 of 34)

Auxiliary Building at Elevation 15-9 Tier 2 12.3-134 Revision 23

12. RADIATION PROTECTION US-APWR Design Control Document Security-Related Information - Withheld Under 10 CFR 2.390 Figure 12.3-3 Post Accident Radiation Zone MAP:1hour After Accident (Sheet 1 of 10)

Power Block at Elevation 4 Tier 2 12.3-161 Revision 32

12. RADIATION PROTECTION US-APWR Design Control Document Security-Related Information - Withheld Under 10 CFR 2.390 Figure 12.3-3 Post Accident Radiation Zone MAP:1hour After Accident (Sheet 2 of 10)

Power Block at Elevation 7 Tier 2 12.3-162 Revision 32

12. RADIATION PROTECTION US-APWR Design Control Document Security-Related Information - Withheld Under 10 CFR 2.390 Figure 12.3-3 Post Accident Radiation Zone MAP:1hour After Accident (Sheet 3 of 10)

Power Block at Elevation 3-7 Tier 2 12.3-163 Revision 32

12. RADIATION PROTECTION US-APWR Design Control Document Security-Related Information - Withheld Under 10 CFR 2.390 Figure 12.3-3 Post Accident Radiation Zone MAP:1hour After Accident (Sheet 4 of 10)

Power Block at Elevation 13-6 Tier 2 12.3-164 Revision 32

12. RADIATION PROTECTION US-APWR Design Control Document Security-Related Information - Withheld Under 10 CFR 2.390 Figure 12.3-4 Post Accident Radiation Zone MAP:1day After Accident (Sheet 1 of 10)

Power Block at Elevation 4 Tier 2 12.3-171 Revision 32

12. RADIATION PROTECTION US-APWR Design Control Document Security-Related Information - Withheld Under 10 CFR 2.390 Figure 12.3-4 Post Accident Radiation Zone MAP:1day After Accident (Sheet 2 of 10)

Power Block at Elevation 7 Tier 2 12.3-172 Revision 32

12. RADIATION PROTECTION US-APWR Design Control Document Security-Related Information - Withheld Under 10 CFR 2.390 Figure 12.3-4 Post Accident Radiation Zone MAP:1day After Accident (Sheet 3 of 10)

Power Block at Elevation 3-7 Tier 2 12.3-173 Revision 32

12. RADIATION PROTECTION US-APWR Design Control Document Security-Related Information - Withheld Under 10 CFR 2.390 Figure 12.3-4 Post Accident Radiation Zone MAP:1day After Accident (Sheet 4 of 10)

Power Block at Elevation 13-6 Tier 2 12.3-174 Revision 32

12. RADIATION PROTECTION US-APWR Design Control Document Security-Related Information - Withheld Under 10 CFR 2.390 Figure 12.3-5 Post Accident Radiation Zone MAP:1week After Accident (Sheet 1 of 10)

Power Block at Elevation 4 Tier 2 12.3-181 Revision 32

12. RADIATION PROTECTION US-APWR Design Control Document Security-Related Information - Withheld Under 10 CFR 2.390 Figure 12.3-5 Post Accident Radiation Zone MAP:1week After Accident (Sheet 2 of 10)

Power Block at Elevation 7 Tier 2 12.3-182 Revision 32

12. RADIATION PROTECTION US-APWR Design Control Document Security-Related Information - Withheld Under 10 CFR 2.390 Figure 12.3-5 Post Accident Radiation Zone MAP:1week After Accident (Sheet 3 of 10)

Power Block at Elevation 3-7 Tier 2 12.3-183 Revision 32

12. RADIATION PROTECTION US-APWR Design Control Document Security-Related Information - Withheld Under 10 CFR 2.390 Figure 12.3-5 Post Accident Radiation Zone MAP:1week After Accident (Sheet 4 of 10)

Power Block at Elevation 13-6 Tier 2 12.3-184 Revision 32

12. RADIATION PROTECTION US-APWR Design Control Document Security-Related Information - Withheld Under 10 CFR 2.390 Figure 12.3-6 Post Accident Radiation Zone MAP:1month After Accident (Sheet 1 of 10)

Power Block at Elevation 4 Tier 2 12.3-191 Revision 32

12. RADIATION PROTECTION US-APWR Design Control Document Security-Related Information - Withheld Under 10 CFR 2.390 Figure 12.3-6 Post Accident Radiation Zone MAP:1month After Accident (Sheet 2 of 10)

Power Block at Elevation 7 Tier 2 12.3-192 Revision 32

12. RADIATION PROTECTION US-APWR Design Control Document Security-Related Information - Withheld Under 10 CFR 2.390 Figure 12.3-6 Post Accident Radiation Zone MAP:1month After Accident (Sheet 3 of 10)

Power Block at Elevation 3-7 Tier 2 12.3-193 Revision 32

12. RADIATION PROTECTION US-APWR Design Control Document Security-Related Information - Withheld Under 10 CFR 2.390 Figure 12.3-6 Post Accident Radiation Zone MAP:1month After Accident (Sheet 4 of 10)

Power Block at Elevation 13-6 Tier 2 12.3-194 Revision 32

12. RADIATION PROTECTION US-APWR Design Control Document Security-Related Information - Withheld Under 10 CFR 2.390 Figure 12.3-11 Post Accident Radiation Zone MAP:1week After Accident (Sheet 1 of 10)

Power Block at Elevation 4 Tier 2 12.3-205 Revision 23

12. RADIATION PROTECTION US-APWR Design Control Document Security-Related Information - Withheld Under 10 CFR 2.390 Figure 12.3-11 Post Accident Radiation Zone MAP:1week After Accident (Sheet 2 of 10)

Power Block at Elevation 7 Tier 2 12.3-206 Revision 23

12. RADIATION PROTECTION US-APWR Design Control Document Security-Related Information - Withheld Under 10 CFR 2.390 Figure 12.3-11 Post Accident Radiation Zone MAP:1week After Accident (Sheet 3 of 10)

Power Block at Elevation 3-7 Tier 2 12.3-207 Revision 23

12. RADIATION PROTECTION US-APWR Design Control Document Security-Related Information - Withheld Under 10 CFR 2.390 Figure 12.3-11 Post Accident Radiation Zone MAP:1week After Accident (Sheet 4 of 10)

Power Block at Elevation 13-6 Tier 2 12.3-208 Revision 23

US-APWR DCD Revision 2 RAI Tracking Report MUAP-09026(R5)

Chapter 14 Mitsubishi Heavy Industries, LTD.

US-APWR DCD Chapter 14 Rev. 2, Tracking Report Rev. 5 Change List Location (e.g., subsection with Page paragraph/sentence/ Description of Change Item, table with column/row, or figure) 14.2-65 14.2.12.1.34 Added C.4 to include testing for confirmation that impacts due to water hammer are within acceptable limits.

RAI 585, 09.02.01-36 14.2-130 14.2.12.1.99 Replaced Criterion with Criteria of heading D.

Added new acceptance criterion as D.2.

Added new acceptance criterion as D.3.

RAI 592, 09.04.02-6 RAI 634, 09.04.03-13 14.2-28 Table 14.2-1 (Sheet 4 Newly created to include Equipment Hatch Hoist of 5) Preoperational Test.

14.2-146, 147 14.2.12.1.118 RAI 616, 09.01.05-18 14A-14 Table 14A-1 (Sheet 14 of 21) 14.2-176 14.2.12.4.11 Replaced at or around the design temperatures based on outside air ambient design conditions with within the design temperatures based on the design basis environmental conditions and design basis heat loads.

RAI 558, 06.05.01-18 RAI 615, 06.05.01-20 14.3-21 14.3.4.7 Deleted in combination with mobile processing equipment from 9th bullet.

RAI 629, 11.03-18 Page 1 of 1

14. VERIFICATION PROGRAMS US-APWR Design Control Document Table 14.2-1 Comprehensive Listing of Tests (Sheet 4 of 5)

Section Test 14.2.12.1.118 Equipment Hatch Hoist Preoperational Test 14.2.12.2.1.1 RCS Sampling for Fuel Loading 14.2.12.2.1.2 Fuel Loading Instrumentation and Neutron Source Requirements Test 14.2.12.2.1.3 Initial Fuel Loading 14.2.12.2.1.4 Inverse Count Rate Ratio Monitoring for Fuel Loading 14.2.12.2.1.5 Precritical Test Sequence 14.2.12.2.1.6 Rod Drop Time Measurement Test 14.2.12.2.1.7 CRDM Operational Test 14.2.12.2.1.8 Rod Position Indication Test 14.2.12.2.1.9 Rod Control System Test 14.2.12.2.1.10 Reactor Protection System Test 14.2.12.2.1.11 RCS Final Leak Test 14.2.12.2.1.12 Incore Detector Test 14.2.12.2.1.13 RCS Flow Coastdown Test 14.2.12.2.1.14 Operational Alignment of Process Temperature Instrumentation Test 14.2.12.2.2.1 Initial Criticality Test Sequence 14.2.12.2.2.2 Initial Criticality 14.2.12.2.2.3 Determination of Core Power Range for Physics Testing 14.2.12.2.3.1 Low Power Test Sequence 14.2.12.2.3.2 Boron Endpoint Determination Test 14.2.12.2.3.3 Isothermal Temperature Coefficient Measurement Test 14.2.12.2.3.4 RCCA Bank Worth Measurement at Zero Power Test 14.2.12.2.3.5 Pseudo Rod Ejection Test 14.2.12.2.3.6 Operational Alignment of Nuclear Instrumentation Test 14.2.12.2.3.7 Dynamic Automatic Turbine Bypass Control Test 14.2.12.2.3.8 Pressurizer Heater and Spray Capability and Continuous Spray Flow Verification Test 14.2.12.2.3.9 Natural Circulation Test 14.2.12.2.3.10 Automatic Low Power SG Water Level Control Test 14.2.12.2.4.1 Power Ascension Test Sequence 14.2.12.2.4.2 Power Coefficient Determination Test 14.2.12.2.4.3 Axial Flux Difference Instrumentation Calibration Test and Axial Distribution Oscillation Test 14.2.12.2.4.4 Flux Map Test 14.2.12.2.4.5 RCCA Misalignment Measurement and Radial Power Distribution Oscillation Test 14.2.12.2.4.6 Remote Shutdown Test 14.2.12.2.4.7 Loose Parts Monitoring System Test (Continuation of 14.2.12.1.72) 14.2.12.2.4.8 Automatic Rod Control System Test 14.2.12.2.4.9 Operational Alignment of Process Temperature Instrumentation at Power Test 14.2.12.2.4.10 Thermal Power Measurement and Statepoint Data Collection Test 14.2.12.2.4.11 Ventilation Capability Test 14.2.12.2.4.12 RCS Flow Measurement Test 14.2.12.2.4.13 Process and Effluent Radiation Monitoring System Test Tier 2 14.2-28 Revision 23

14. VERIFICATION PROGRAMS US-APWR Design Control Document

c. Blowdown water system.

d. Demineralized water system.

C. Test Method

1. Verify controls and functions of circulating water pump and cooling tower fans.

2. Verify strokes and functions of motor-operated valves and control valves.

3. Operate circulating water pumps and verify operating condition.

4. Verify indications and alarms.

D. Acceptance Criteria

1. The circulating water system performs as described in Subsection 10.4.5.

2. Indications and alarms operate as described in Subsection 10.4.5.

14.2.12.1.34 Essential Service Water System (ESWS) Preoperational Test A. Objective

1. To demonstrate the operation of the ESWS.

B. Prerequisites

1. Required construction testing is completed.

2. Component testing and instrument calibration is completed.

3. Test instrumentation is available and calibrated.

4. Required support systems are available.

C. Test Method

1. Verify manual and automatic system controls.

2. Verify system flowrates and performance of ESWS pumps.

3. Verify alarms and status indications are functional.

4. Verify the absence of indications of water hammer by re-activating the ESW pump after a simulated LOOP as specified in Subsection 14.2.12.1.45, Class 1E Bus Load Sequence Preoperational Test.

D. Acceptance Criterion

1. The ESWS operates within design limits, as described in Subsection 9.2.1.

Tier 2 14.2-65 Revision 23

14. VERIFICATION PROGRAMS US-APWR Design Control Document

4. Verify penetration and safeguard component area and discharge duct of auxiliary building HVAC system isolation on a simulated ECCS actuation signal.

D. Acceptance CriterionCriteria

1. The auxiliary building HVAC system operates as described in Subsection 9.4.3.

2. The auxiliary building HVAC system maintains the exhaust airflow rates from radiological controlled areas described in Table 12.2-60.

3. Ventilation flow balancing of the auxiliary building HVAC system is performed as described in Subsection 9.4.3.

14.2.12.1.100 Main Steam/Feedwater Piping Area HVAC System Preoperational Test A Objective

1. To demonstrate operation of the main steam/feedwater piping area HVAC system.

B. Prerequisites

1. Required construction testing is completed.

2. Component testing and instrument calibration is completed.

3. Test instrumentation is available and calibrated.

4. Required support systems are available.

C. Test Method

1. Verify manual and automatic controls in the normal operating mode.

2. Verify alarms and indications are functional.

3. Verify design airflow.

D. Acceptance Criterion

1. The main steam/feedwater piping area HVAC system operates as described in Subsection 9.4.3.

14.2.12.1.101 MCR HVAC System Preoperational Test (including MCR Habitability)

A. Objectives

1. To demonstrate operation of the MCR HVAC system in normal, isolation and emergency pressurization modes.

Tier 2 14.2-130 Revision 23

14. VERIFICATION PROGRAMS US-APWR Design Control Document

1. The compressed gas system (nitrogen gas subsystems and hydrogen gas subsystem only) meets design requirements relating to the supply gas pressure.

2. Loads that are a part of (or support the operation of) portions of loads identified in item C.3 above respond to pressure transients in accordance with design.

14.2.12.1.118 Equipment Hatch Hoist Preoperational Test A. Objective

1. To demonstrate operation of the equipment hatch hoist, associated equipment and accessories.

2. To verify the equipment hatch hoist, associated equipment and accessories have completed static testing at 125% and operational testing (dynamic load testing) at 100% of rated load.

B. Prerequisites

1. Required construction testing is completed.

2. Component testing and instrument calibration is completed.

3. Test instrumentation is available and calibrated.

4. Required support systems are available.

C. Test Method

1. Verify control circuitry, limit devices, safety devices, and interlocks for the equipment hatch hoist as described in Subsection 9.1.5.5.

2. Perform static load testing at 125% of rated load for the equipment hatch hoist, associated equipment and accessories, followed by appropriate inspections.

3. Perform operational (dynamic load) testing at 100% of rated load of the equipment hatch hoist, associated equipment and accessories followed by appropriate inspections.

4. Testing and inspection includes testing and inspection requirements specified by NUREG-0554 (Reference 14.2-24), ASME NOG-1 (Reference 14.2-30), and NUREG-0612 (Reference 14.2-21) as applicable.

D. Acceptance Criteria

1. The equipment hatch hoist and its associated interlocks, limit devices, safety devices and control circuits perform as specified in Subsection 9.1.5.

2. The equipment hatch hoist static testing at 125% of rated load and operational (dynamic load) testing at 100% of rated load is completed and the hoist satisfactorily passes inspections in accordance with NUREG-0554 (Reference Tier 2 14.2-146 Revision 23

14. VERIFICATION PROGRAMS US-APWR Design Control Document 14.2-24), ASME NOG-1 (Reference 14.2-30), and NUREG-0612, (Reference 14.2-21).

3. The associated equipment and accessories satisfactorily pass an inspection following static and operational (dynamic load) testing in accordance with NUREG-0612 (Reference 14.2-21) and NUREG-0554 (Reference 14.2-24).

4. Testing and inspection demonstrates compliance with testing and inspection requirements specified by NUREG-0554 (Reference 14.2-24), ASME NOG-1 (Reference 14.2-30) and NUREG-0612 (Reference 14.2-21) as applicable.

Tier 2 14.2-147 Revision 23

14. VERIFICATION PROGRAMS US-APWR Design Control Document

2. Record temperature readings in specified areas while operating the designed minimum number of HVAC components consistent with existing plant conditions.

3. Record surface concrete temperatures adjacent to the high temperature piping penetrations and at selected locations on the concrete shielding primary shield wall, and at selected locations between the reactor vessel support base plates and concrete (at 100% power only).

4. Record outside air ambient environmental conditions.

D. Acceptance Criteria

1. Temperature conditions are maintained in the containment and ESF areas in accordance with Subsections 9.4.5, 9.4.6, and Table 9.4-1. It has been demonstrated through testing and analyses that the temperatures for these areas are being maintained at or around within the design temperatures based on the design basis environmental outside air ambient design conditions and design basis heat loads.

2. Concrete surface temperatures are maintained in accordance with Subsection 9.4.6.1.2.3.

14.2.12.2.4.12 RCS Flow Measurement Test A. Objectives

1. As for RCS flow rate measurement, RCS flow rate is determined based on the correlation between data obtained by measuring RCP motor input power and the differential pressure across the reactor coolant line elbow tap, for the purpose of confirming reactor coolant flow is equal to or greater than the design flow specified in Section 5.1.

2. To perform calorimetric flow measurements at 50%, 75%, and 100% power, for the purpose of confirming RCS flow is equal to or greater than the design flow in Section 5.1.

B. Prerequisites

1. Required instrument calibration is completed.

2. Required support systems are operational.

3. The reactor core is installed, and the plant is at normal operating temperature and pressure prior to initial criticality.

C. Test Method

1. Prior to criticality, operating all RCPs and any combination of them including a single operation, input power of each operating RCP motor and relating RCS line elbow tap differential pressure is measured. RCS flow rate is calculated using Tier 2 14.2-176 Revision 23

14. VERIFICATION PROGRAMS US-APWR Design Control Document Required interfaces with other systems Numeric performance values Verifying the performance of the liquid waste management system (as permanently installed systems or in combination with mobile processing equipment)

Verifying the performance of the gaseous waste management system (as permanently installed systems or in combination with mobile processing equipment)

Verifying the performance of the solid waste management system (as permanently installed systems or in combination with mobile processing equipment)

Verifying the performance of the process and effluent radiological monitoring instrumentation and sampling systems (as permanently installed systems or in combination with portable skid-mounted equipment)

Verifying the performance of the light load handling system and overhead heavy load handling system.

ITAAC for plant piping systems follow NRC guidelines for fluid systems ITAAC in Appendix C.II.1-A of RG 1.206 (Reference 14.3-1), as summarized above.

Table 14.3-6 lists the systems which the design is addressed in Tier 1.

The COL applicant provides the ITAAC for the site specific portion of the plant systems specified in Subsection 14.3.5, Interface Requirements.

14.3.4.8 ITAAC for Radiation Protection Section 2.8 of Tier 1, which addresses radiation protection, is prepared in accordance with the guidance in RG 1.206 (Reference 14.3-1), SRP 14.3 (Reference 14.3-2), and SRP 14.3.8 (Reference 14.3-12). ITAAC related to radiation protection are provided for those SSCs that provide radiation shielding, confinement or containment of radioactivity, ventilation of airborne contamination, or monitoring of radiation (or radioactivity concentration) for normal operations and during accidents. These ITAAC provide for the following:

Verifying the adequacy of as-built walls, structures, and buildings as radiation shields, as applicable Verifying the plant airborne concentrations of radioactive materials through adequate design of ventilation and airborne monitoring systems Verifying the functional arrangement of ventilation systems Tier 2 14.3-21 Revision 23

14. VERIFICATION PROGRAMS US-APWR Design Control Document Table 14A-1 Conformance Matrix of RG 1.68 Appendix A Guidance Versus Typical Test Abstracts (Sheet 14 of 21)

RG 1.68 Section Number Typical Test Appendix A 14.2.12.1.98 Class 1E Electrical Room HVAC System Preoperational Test 14.2.12.1.102 Non-Class 1E Electrical Room HVAC System 1.n.(14) (c)

Preoperational Test 14.2.12.1.111 Turbine Building Area Ventilation System (Electrical Equipment Area) Preoperational Test Not applicable.

1.n.(14) (d) -

Class 1E Gas Turbine Generator contains the function.

14.2.12.1.99 Auxiliary Building HVAC System Preoperational Test 14.2.12.1.100 Main Steam/Feedwater Piping Area HVAC System Preoperational Test 14.2.12.1.102 Non-Essential Electrical Room HVAC System 1.n.(14) (e) Preoperational Test, 14.2.12.1.103 Technical Support Center HVAC System Preoperational Test 14.2.12.1.110 Turbine Building Area Ventilation System (General Mechanical Area) Preoperational Test MCR HVAC System Preoperational Test (including 1.n.(14) (f) 14.2.12.1.101 MCR Habitability) 1.n.(15) 14.2.12.1.66 Reactor Cavity Cooling System Preoperational Test Not applicable.

1.n.(16) -

This is not a design feature of the US-APWR.

Not applicable.

1.n.(17) -

This is not a design feature of the US-APWR.

Not applicable.

1.n.(18) -

This is not a design feature of the US-APWR.

14.2.12.1.105 Vessel Servicing preoperational Test 1.o.(1) 14.2.12.1.118 Equipment Hatch Hoist Preoperational Test 14.2.12.1.105 Vessel Servicing preoperational Test 1.o.(2) 14.2.12.1.118 Equipment Hatch Hoist Preoperational Test 14.2.12.1.105 Vessel Servicing preoperational Test 1.o.(3) 14.2.12.1.118 Equipment Hatch Hoist Preoperational Test Tier 2 14A-14 Revision 32

US-APWR DCD Revision 2 RAI Tracking Report MUAP-09026(R5)

Chapter 16 Mitsubishi Heavy Industries, LTD.

US-APWR DCD Chapter 16 Rev. 2, Tracking Report Rev. 5 Change List Location (e.g., subsection with Page paragraph/sentence/ Description of Change Item, table with column/row, or figure)

Technical Specifications 3.9.6 4 LCO RAI No.628, Question No.19.01-8 ACTIONS LCO: Added , and low-pressure letdown line isolation valve shall be OPERABLE SURVEILLANCE REQUIREMENTS ACTIONS: Added Item B, and Replace B with C SURVEILLANCE REQUIEMENTS: Added SR 3.9.6.3 Bases B3.4.8-4 SR3.4.8.3 RAI No.628, Question No.19.01-8 Added This requirements mean confirmation of OPERABILITY of Instrumentation and its control (Setpoints, Channel Checks, Channel Calibrations) and valve.

B3.9.6 B 3.4.8 RAI No.628, Question No.19.01-8 6

Background:

Added In MODE 6 Low Water Level, low-pressure letdown line isolation valves are automatically closed upon detection of RCS loop low-level signal to prevent loss of RCS inventory. The function is effective to prevent core damage during plant shutdown, based on probabilistic risk assessment.

LCO: Added The LCO requires the low-pressure letdown line isolation valves to be OPERABLE to mitigate the effects associated with loss of RCS inventory.

APPLICABILITY: Added In MODE 6 Low Water Level, low-pressure letdown line isolation valves are automatically closed upon detection of RCS loop low-level signal to prevent loss of RCS inventory. The function is effective to prevent core damage during plant shutdown, based on probabilistic risk assessment.

ACTIONS: Added B.1 If one low-pressure letdown isolation valve is inoperable, the automatic isolation function to prevent loss of RCS inventory is lost. Action must be initiated to restore the valve to OPERABLE status. The immediate Completion Time reflects the importance of maintaining the availability of three paths for heat removal. and replaced B with C SURVEILLANCE REQUIEMENTS: Added SR 3.9.6.3 Page 1 of 1

RHR and Coolant Circulation - Low Water Level 3.9.6 3.9 REFUELING OPERATIONS 3.9.6 Residual Heat Removal (RHR) and Coolant Circulation - Low Water Level LCO 3.9.6 Three RHR loops shall be OPERABLE, and two RHR loops shall be in operation, and low-pressure letdown line isolation valve shall be OPERABLE.

---

NOTES-------------------------------------------

1. All RHR pumps may be removed from operation for 15 minutes when switching from one train to another provided:

a. The core outlet temperature is maintained > 10 degrees F below saturation temperature,

b. No operations are permitted that would cause introduction of coolant into the Reactor Coolant System (RCS) with boron concentration less than that required to meet the minimum required boron concentration of LCO 3.9.1, and

c. No draining operations to further reduce RCS water volume are permitted.

2. One required RHR loop may be inoperable for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for surveillance testing, provided that the other RHR loops are OPERABLE and in operation.

APPLICABILITY: MODE 6 with the water level < 23 ft above the top of reactor vessel flange.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Less than the required A.1 Initiate action to restore Immediately number of RHR loops required RHR loops to OPERABLE. OPERABLE status.

OR US-APWR 3.9.6-1 Revision 23

RHR and Coolant Circulation - Low Water Level 3.9.6 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME A.2 Initiate action to establish Immediately 23 ft of water above the top of reactor vessel flange.

B. One low-pressure B.1 Initiate action to restore Immediately letdown isolation valve low-pressure letdown line inoperable. isolation valve to OPERABLE status.

BC. No RHR loop in BC.1 Suspend operations that operation. would cause introduction of Immediately coolant into the RCS with boron concentration less than required to meet the boron concentration of LCO 3.9.1.

AND BC.2 Initiate action to restore two RHR loops to operation. Immediately AND BC.3 Close equipment hatch and secure with [four] bolts. 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> AND BC.4 Close one door in each air lock. 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> AND US-APWR 3.9.6-2 Revision 23

RHR and Coolant Circulation - Low Water Level 3.9.6 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME BC.5.1 Close each penetrations 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> providing direct access from the containment atmosphere to the outside atmosphere with a manual or automatic isolation valve, blind flange, or equivalent.

OR BC.5.2 Verify each penetration is 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> capable of being closed by an OPERABLE Containment Purge and Exhaust Isolation System.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.6.1 Verify two RHR loops are in operation and [12 hours circulating reactor coolant at a flow rate of 2645 gpm per pump. OR In accordance with the Surveillance Frequency Control Program]

SR 3.9.6.2 Verify correct breaker alignment and indicated [7 days power available to the required RHR pump that is not in operation. OR In accordance with the Surveillance Frequency Control Program]

US-APWR 3.9.6-3 Revision 23

RHR and Coolant Circulation - Low Water Level 3.9.6 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.9.6.3 Perform a complete cycle of each low-pressure [24 months letdown line isolation valve.

OR In accordance with the Surveillance Frequency Control Program]

US-APWR 3.9.6-4 Revision 23

RCS Loops - MODE 5, Loops Not Filled B 3.4.8 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.4.8.3 SR 3.4.8.3 requires a complete cycle of each low-pressure letdown isolation valve. This requirements mean confirmation of OPERABILITY of Instrumentation and its control (Setpoints, Channel Checks, Channel Calibrations) and valve. Operating a low-pressure letdown isolation valve through one complete cycle ensures that the low-pressure letdown isolation valve can be automatically actuated to mitigate the effects from loss of RCS inventory. [The Frequency of 24 months is based on engineering judgment, taking into consideration the conditions required to perform the Surveillance, and is intended to be consistent with expected fuel cycle length. This equipment is not at risk of imminent damage as it is designed to remain functional and in good condition while in operation, thus significant degradation due to a longer surveillance interval should not be of major concern. The design reliability is, therefore, maintained by taking these considerations into account based on sound engineering judgment.

OR The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.]

REFERENCES None.

US-APWR B 3.4.8-4 Revision 23

RHR and Coolant Circulation - Low Water Level B 3.9.6 B 3.9 REFUELING OPERATIONS B 3.9.6 Residual Heat Removal (RHR) and Coolant Circulation - Low Water Level BASES BACKGROUND The purpose of the RHR System in MODE 6 is to remove decay heat and sensible heat from the Reactor Coolant System (RCS), as required by GDC 34, to provide mixing of borated coolant, and to prevent boron stratification (Ref. 1). Heat is removed from the RCS by circulating reactor coolant through the Containment Spray (CS)/RHR heat exchangers where the heat is transferred to the Component Cooling Water System. The coolant is then returned to the RCS via the RCS cold leg(s). Operation of the RHR System for normal cooldown decay heat removal is manually accomplished from the control room. The heat removal rate is adjusted by controlling the flow of reactor coolant through the CS/RHR heat exchanger(s) and the bypass lines. Mixing of the reactor coolant is maintained by this continuous circulation of reactor coolant through the RHR System.

In MODE 6 Low Water Level, low-pressure letdown line isolation valves are automatically closed upon detection of RCS loop low-level signal to prevent loss of RCS inventory.

The function is effective to prevent core damage during plant shutdown, based on probabilistic risk assessment.

APPLICABLE While there is no explicit analysis assumptions for the decay heat removal SAFETY function of the RHR System in MODE 6, if the reactor coolant ANALYSES temperature is not maintained below 200°F, boiling of the reactor coolant could result. This could lead to a loss of refueling cavity water level.

Additionally, boiling of the reactor coolant could lead to a reduction in boron concentration in the coolant due to the boron plating out on components near the areas of the boiling activity. The loss of reactor coolant and the reduction of boron concentration in the reactor coolant will eventually challenge the integrity of the fuel cladding, which is a fission product barrier. Three trains of the RHR System are required to be OPERABLE, and two trains in operation, in order to prevent this challenge.

RHR and Coolant Circulation - Low Water Level satisfies Criterion 4 of 10 CFR 50.36(c)(2)(ii).

LCO In MODE 6, with the water level < 23 ft above the top of the reactor vessel flange, three RHR loops must be OPERABLE. Additionally, two loops of RHR must be in operation in order to provide:

a. Removal of decay heat,

b. Mixing of borated coolant to minimize the possibility of criticality, and US-APWR B 3.9.6-1 Revision 23

RHR and Coolant Circulation - Low Water Level B 3.9.6 BASES LCO (continued)

c. Indication of reactor coolant temperature.

The LCO requires the low-pressure letdown line isolation valves to be OPERABLE to mitigate the effects associated with loss of RCS inventory.

This LCO is modified by two Notes. Note 1 permits the RHR pumps to be removed from operation for 15 minutes when switching from one train to another. The circumstances for stopping all RHR pumps are to be limited to situations when the outage time is short and the core outlet temperature is maintained > 10 degrees F below saturation temperature.

The Note prohibits boron dilution or draining operations when RHR forced flow is stopped.

Note 2 allows one RHR loop to be inoperable for a period of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> provided the other loops are OPERABLE and in operation. Prior to declaring the loop inoperable, consideration should be given to the existing plant configuration. This consideration should include that the core time to boil is short, there is no draining operation to further reduce RCS water level and that the capability exists to inject borated water into the reactor vessel. This permits surveillance tests to be performed on the inoperable loop during a time when these tests are safe and possible.

An OPERABLE RHR loop consists of an CS/RHR pump, a heat exchanger, valves, piping, instruments and controls to ensure an OPERABLE flow path and to determine the low end temperature. The flow path starts in one of the RCS hot legs and is returned to the RCS cold legs.

All RHR pumps may be aligned to the Refueling Water Storage Pit to support filling or draining the refueling cavity or for performance of required testing.

APPLICABILITY Three RHR loops are required to be OPERABLE, and two RHR loops must be in operation in MODE 6, with the water level < 23 ft above the top of the reactor vessel flange, to provide decay heat removal and mixing of the borated coolant. Requirements for the RHR System in other MODES are covered by LCOs in Section 3.4, Reactor Coolant System (RCS), and Section 3.5, Emergency Core Cooling Systems (ECCS). RHR loop requirements in MODE 6 with the water level 23 ft are located in LCO 3.9.5, "Residual Heat Removal (RHR) and Coolant Circulation - High Water Level."

In MODE 6 Low Water Level, low-pressure letdown line isolation valves are automatically closed upon detection of RCS loop low-level signal to prevent loss of RCS inventory.

US-APWR B 3.9.6-2 Revision 23

RHR and Coolant Circulation - Low Water Level B 3.9.6 BASES APPLICABILITY (continued)

The function is effective to prevent core damage during plant shutdown, based on probabilistic risk assessment.

ACTIONS A.1 and A.2 If less than the required number of RHR loops are OPERABLE, action shall be immediately initiated and continued until the RHR loop is restored to OPERABLE status and to operation or until 23 ft of water level is established above the reactor vessel flange. When the water level is 23 ft above the reactor vessel flange, the Applicability changes to that of LCO 3.9.5, and only one RHR loop is required to be OPERABLE and in operation. An immediate Completion Time is necessary for an operator to initiate corrective actions.

B.1 If one low-pressure letdown isolation valve is inoperable, the automatic isolation function to prevent loss of RCS inventory is lost. Action must be initiated to restore the valve to OPERABLE status. The immediate Completion Time reflects the importance of maintaining the availability of three paths for heat removal.

BC.1 If no RHR loop is in operation, there will be no forced circulation to provide mixing to establish uniform boron concentrations. Suspending positive reactivity additions that could result in failure to meet the minimum boron concentration limit is required to assure continued safe operation. Introduction of coolant inventory must be from sources that have a boron concentration greater than that what would be required in the RCS for minimum refueling boron concentration. This may result in an overall reduction in RCS boron concentration, but provides acceptable margin to maintaining subcritical operation.

BC.2 If no RHR loop is in operation, actions shall be initiated immediately, and continued, to restore two RHR loop to operation. Since the unit is in Conditions A and B concurrently, the restoration of three OPERABLE RHR loops and at least two operating RHR loop should be accomplished expeditiously.

US-APWR B 3.9.6-3 Revision 23

RHR and Coolant Circulation - Low Water Level B 3.9.6 BASES ACTIONS (continued)

BC.3, BC.4, BC.5.1, and BC.5.2 If no RHR is in operation, the following actions must be taken:

a. The equipment hatch must be closed and secured with [four] bolts,

b. One door in each air lock must be closed, and

c. Each penetration providing direct access from the containment atmosphere to the outside atmosphere must be either closed by a manual or automatic isolation valve, blind flange, or equivalent, or verified to be capable of being closed by an OPERABLE Containment Purge and Exhaust Isolation System.

With RHR loop requirements not met, the potential exists for the coolant to boil and release radioactive gas to the containment atmosphere.

Performing the actions stated above ensures that all containment penetrations are either closed or can be closed so that the dose limits are not exceeded.

The Completion Time of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> allows fixing of most RHR problems and is reasonable, based on the low probability of the coolant boiling in that time.

US-APWR B 3.9.6-4 Revision 23

RHR and Coolant Circulation - Low Water Level B 3.9.6 BASES SURVEILLANCE SR 3.9.6.1 REQUIREMENTS This Surveillance demonstrates that two RHR loops are in operation and circulating reactor coolant. The flow rate is determined by the flow rate necessary to provide sufficient decay heat removal capability and to prevent thermal and boron stratification in the core. In addition, during operation of the RHR loops with the water level in the vicinity of the reactor vessel nozzles, the CS/RHR pump suction requirements must be met. [The Frequency of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is sufficient, considering the flow, temperature, pump control, and alarm indications available to the operator for monitoring the RHR System in the control room. OR The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.]

SR 3.9.6.2 Verification that the required pump is OPERABLE ensures that an additional RCS or RHR pump can be placed in operation, if needed, to maintain decay heat removal and reactor coolant circulation. Verification is performed by verifying proper breaker alignment and power available to the required pump. [The Frequency of 7 days is considered reasonable in view of other administrative controls available and has been shown to be acceptable by operating experience. OR The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.]

SR 3.9.6.3 SR 3.9.6.3 requires a complete cycle of each low-pressure letdown isolation valve. This requirements mean confirmation of OPERABILITY of Instrumentation and its control (Setpoints, Channel Checks, Channel Calibrations) and valve. Operating a low-pressure letdown isolation valve through one complete cycle ensures that the low-pressure letdown isolation valve can be automatically actuated to mitigate the effects from loss of RCS inventory. [The Frequency of 24 months is based on engineering judgment, taking into consideration the conditions required to perform the Surveillance, and is intended to be consistent with expected fuel cycle length. This equipment is not at risk of imminent damage as it is designed to remain functional and in good condition while in operation, thus significant degradation due to a longer surveillance interval should not be of major concern. The design reliability is, therefore, maintained by taking these considerations into account based on sound engineering judgment. OR The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.]

US-APWR B 3.9.6-5 Revision 23

RHR and Coolant Circulation - Low Water Level B 3.9.6 BASES REFERENCES 1. Subsection 5.4.7.

US-APWR B 3.9.6-6 Revision 23

US-APWR DCD Revision 2 RAI Tracking Report MUAP-09026(R5)

Chapter 19 Mitsubishi Heavy Industries, LTD.

US-APWR DCD Chapter 19 Rev. 2, Tracking Report Rev. 5 Change List Location (e.g., subsection with Page paragraph/sentence/ Description of Change item ,table with column/row, or figure) 19.1-5, 6 19.1.2.4 RAI#639 19-457 and RAI#640 19-470 Fourth paragraph, Deleted that, Insert new paragraph, In accordance with After the last also be concluded. , and revise the sentences.

paragraph of Page 19.1-5, and the third and fourth paragraphs.

19.1-87 19.1.5.2 RAI#639 19-456 First paragraph Inserted new paragraphs, The US-APWR design features located in turbine building.

19.1-91 19.1.5.2.1 RAI#639 19-464 After the item e. Insert new item f. as follows.

f. An evaluation has been performed by assuming total damage of the fire-originating fire PRA compartment in consideration of fire heat influence.

19.1-92 19.1.5.2.1 RAI#639 19-458 After the last Inserted new paragraph, The oil collection paragraph of conservative assumptions.

Subsection 19.1.5.2.1 19.1-93 to 19.1.5.2.1 RAI#641 19-478 19.1-94 After the last Inserted new paragraphs, The fire ignition frequencies, paragraph of compartment fire scenario.

Subsection 19.1.5.2.1 19.1-104 to 19.1.5.2.3 RAI#639 19-455 19.1-105 Before the last two Insert the summary of internal fire risk insights.

paragraphs on page 19.1-104 to 105.

19.1-106 to 19.1.5.3 RAI#640 19-467 19.1-107 First paragraph Insert a new paragraph, The US-APWR design features to the yard.

Page 1 of 4

US-APWR DCD Chapter 19 Rev. 2, Tracking Report Rev. 5 Change List Location (e.g., subsection with Page paragraph/sentence/ Description of Change item ,table with column/row, or figure) 19.1-116 to 19.1.5.3.2 RAI#640 19-469 19.1-117 Before the last two Insert the summary of internal flooding risk insights.

paragraphs on page 19.1-117.

19.1-121 to 19.1.6.1 RAI#621 19.01-3 19.1-122 Bullet of POS4: Inserted the additional description for POS 4.

RHR cooling Inserted the statement for POSs 4-1, 4-2 and 4-3.

(mid-loop operation) 19.1-122 19.1.6.1 RAI#621 19.01-2 Bullet of POS6: No Revised the statement for POS 6.

fuel in the core 19.1-123 19.1.6.1 RAI#621 19.01-5 Bullet of RHR cooling Inserted the additional description for POS 8.

(mid-loop operation Inserted the statement for POSs 8-1, 8-2 and 8-3.

after refueling) 19.1-128 19.1.6.1 RAI#607 19-438 Loss of coolant Revised description in Loss of coolant accident (LOCA) accident (LOCA) to reflect the RAI.

Re-estimated the frequency of LOCA event during POS 8-1.

19.1-129 19.1.6.1 RAI#621 19.01-4 Bullet of GI: Inserted This function is unavailable if the RCS is Gravitational pressurized or if the SG nozzle lid is installed to isolate injection the SG. after the last sentence of GI description.

19.1-133 19.1.6.1 RAI#621 19.01-6 Bullet of AC: Offsite Changed allowable time and LOOP duration probability to power recovery three hours and to 0.17, respectively.

Page 2 of 4

US-APWR DCD Chapter 19 Rev. 2, Tracking Report Rev. 5 Change List Location (e.g., subsection with Page paragraph/sentence/ Description of Change item ,table with column/row, or figure) 19.1-159 to 19.1.6.3.1 RAI#528 19-422 19.1-160 Inserted additional descriptions for seismic at LPSD to reflect the RAIs.

RAI# 621 19.01-7 Updated the above sentences to include additional information.

19.1-173 19.1.9 RAI#607 19-438 Added a new reference An Analysis of Loss of Decay Heat Removal Trends and Initiating Event Frequencies (1989-2000), EPRI 1003113.

19.1-820 Table 19.1-79 RAI#621 19.01-3 Inserted RCS water level RCS full and Reactor cavity full in the line for RCS water level and Remarks regarding over-drain.

19.1-821 Table 19.1-80 RAI#621 19.01-5 Inserted RCS water level RCS full and Reactor cavity full in the line for RCS water lelve and Remarks regarding over-drain.

19.1-822 Table 19.1-81 RAI#621 19.01-2 Sheet 1 Revised reason to exclude POS 6 in LPSD PRA.

19.1-983 Table 19.1-119 RAI#608 19-440 Sheet 20 Replaced at least one of the with both.

19.1-986 to Table 19.1-119 RAI#608 19-455 19.1-987 Sheet 23 Revised one insight and added two insights in Internal fire design features and insights.

19.1-989 to Table 19.1-119 RAI#608 19-455 19.1-990 Internal fire risk Added five insights in Internal fire risk insights.

insights 19.1-990 Table 19.1-120 RAI#528 19-422 and RAI#621 19.01-7 Incorporated new table Initiating Events and Mitigation Systems during LPSD, which is related to seismic.

19.1-991 to Table 19.1-119 RAI#640 19-469 19.1-992 Internal flood insights Revised three insights and added two insights in Internal flood design features and insights.

19.1-994 to Table 19.1-119 RAI#640 19-469 19.1-995 Internal flood risk Added six insights in Internal flood risk insights.

insights Page 3 of 4

US-APWR DCD Chapter 19 Rev. 2, Tracking Report Rev. 5 Change List Location (e.g., subsection with Page paragraph/sentence/ Description of Change item ,table with column/row, or figure) 19.1-1130 Figure 19.1-22 RAI#612 19.01-7 Inserted new figure Loss of RHRS caused by Seismic Event Tree, which is related to seismic.

19.2-8 19.2.3.3.2 RAI#627 19-454 The second to last Inserted For several this operation will inherently paragraph prevent high hydrogen concentration in the RWSP.

19.2-36 19.2.5 (3) RAI#627 19-454 After the last Inserted Firewater can be in the RWSP.

paragraph in (During operations at power)

Page 4 of 4

19. PROBABILISTIC RISK ASSESSMENT US-APWR Design Control Document AND SEVERE ACCIDENT EVALUATION 19.1.2.2 PRA Level of Detail The US-APWR realistically reflects the actual plant design, planned construction, anticipated operational practices, and relevant operational experience. The approach, methods, data, and computer codes that are used, as documented throughout this chapter, are compliant with industry standard codes and practices. The level of detail is sufficient to ensure that the impacts of designed-in dependencies are correctly captured.

The level of detail of the PRA is sufficient to provide confidence in the results such that the PRA may be used in regulatory decision-making to support risk-informed applications in design phase.

19.1.2.3 PRA Technical Adequacy The quality of the methodologies, processes, analyses, and personnel associated with the US-APWR PRA comply with the provisions for nuclear plant quality assurance.

Toward this end, the US-APWR PRA adheres to the recommendations provided in RG 1.200 pertaining to quality and technical adequacy. The US-APWR incorporates the technical elements of an acceptable PRA shown in Table 1 of RG 1.200 (Reference 19.1-9), and is consistent with the technical characteristics and attributes given in Tables 2 and 3 of RG 1.200, entitled Summary of Technical Characteristics and Attributes of a PRA, and Summary of Technical Characteristics and Attributes of an Internal Flood and Fire Analysis and External Hazards Analysis, respectively. The PRA has been developed in accordance with industry consensus standards as described in Section 19.0, and has been subjected to a peer review process as defined in ASME-RA-S-2002 and associated addenda (Reference 19.1-1, 19.1-2, 19.1-3) and as outlined in the Nuclear Energy Institute (NEI) peer review guide (Reference 19.1-14).

19.1.2.4 PRA Maintenance And Upgrade The objective of the PRA maintenance and upgrade program is to ensure that the PRA will be maintained and upgraded so that its representation of the as designed, as-to-be built, and as-to-be operated plant is sufficient to support the applications for which the PRA is being used. The PRA will be under configuration control and the program will contain the following key elements:

• A process for monitoring PRA inputs and collecting new information

• A process that maintains and upgrades the PRA to be consistent with the as-built, as-operated plant

• A process that ensures that the cumulative impact of pending changes is considered when applying the PRA

• A process that evaluates the impact of changes on previously implemented risk-informed decisions that have used the PRA

• A process that maintains configuration control of computer codes used to support PRA quantification

• Documentation of the program Tier 2 19.1-5 Revision 23

19. PROBABILISTIC RISK ASSESSMENT US-APWR Design Control Document AND SEVERE ACCIDENT EVALUATION PRA maintenance involves updating of PRA models to reflect plant changes such as modifications, procedure changes, or plant performance. A PRA upgrade involves the incorporation into the PRA model of new methodologies or significant changes in scope or capability. Those changes could include items such as new human error analysis methodology; new data update methods; new approaches to quantification or truncation; or new treatments of common cause failure (CCF).

In accordance with 10 CFR 50.71(h)(1) (Reference 19.1-15), prior to the scheduled date for initial loading of fuel, a plant-specific PRA that covers initiating events and modes for which NRC-endorsed consensus standards on PRA exist one year prior to the scheduled date for initial loading of fuel shall be developed. The plant-specific PRA will reflect the as-built plant. The plant-specific PRA model will utilize the US-APWR DCD PRA model as a baseline. Any additional modeling changes resulting from the plant-specific design, departures from the design used in the US-APWR DCD PRA, insights from procedure development and operator training, or other PRA modeling changes that are identified subsequent to the completion of the US-APWR DCD PRA will also be utilized. The PRA-based risk insight differences between the plant-specific PRA and the US-APWR DCD PRA will be evaluated. Plant walk-downs to confirm that the assumptions used in the PRA remain valid will also be conducted.

During operation, PRA will be maintained and updated in accordance with approved station procedures on a periodic basis not to exceed two refueling cycles.

Changes in PRA inputs or discovery of new information will be evaluated to determine whether the new or changed information warrants a PRA maintenance or upgrade.

Changes that would impact risk-informed decisions will be prioritized to ensure that the most significant changes are incorporated as soon as practical. Other changes will be incorporated during the next PRA update.

Changes to the PRA due to PRA maintenance and PRA upgrade will meet the risk assessment technical requirements of the NRC-endorsed PRA standards.detailed in Section 4 of ASME RA-S-2002 and associated addenda (Reference 19.1-1, 19.1-2, 19.1-3). Upgrades of the PRA will receive a peer review in accordance with the requirements of the NRC-endorsed PRA standards,detailed in Section 6 of ASME RA-S-2002 and associated addenda, but will be limited to aspects of the PRA that have been upgraded.

The PRA will be updated to reflect plant, operational experience, and PRA modeling changes, consistent with the NRC-endorsed standards, such as those described in Section 19.1, in existence six months prior to the issuance of the maintenance update, which will be scheduled in compliance with 10 CFR 50.71 (Reference 19.1-15) specified criteria and intervals.The PRA will be updated to reflect plant experience, operational experience, and PRA modeling changes, consistent with the NRC-endorsed standards.

These standards are described in Section 19.1 and were in existence one year prior to the issuance of the maintenance update scheduled in compliance with 10 CFR 50.71 specified criteria and intervals.

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19. PROBABILISTIC RISK ASSESSMENT US-APWR Design Control Document AND SEVERE ACCIDENT EVALUATION

5. Depending on whether offsite power is available, different scenarios to trip the reactor are considered. In the case offsite power failed (i.e., a LOOP initiating event), the control rod motor generator sets would be de-energized following LOOP and succeed in the release of control rods into the core even if the reactor trip function failed. Only when the control rod system is failed would the reactor trip be failed. This scenario is considered in this study and the HCLPF value for this event is 0.67 g (dominated by the control rod HCLPF). In case offsite power is available, the failure of the reactor trip function should be considered. However, the HCLPF for the reactor trip system would be higher than 0.67 g determined when offsite power is lost. This is because HCLPFs for electrical equipment and sensors/transmitters to trip the reactor are above 0.67 g. Thus, whether offsite power is available or not, the HCLPF value (i.e., seismic capacity) to trip the reactor is higher than the plant HCLPF of 0.50 g.

6. There are no vulnerabilities for containment performance (i.e., containment integrity, containment isolation and prevention of bypass functions) due to a seismic event.

19.1.5.2 Internal Fires Risk Evaluation The US-APWR design features for reducing fire risk and use of the fire PRA in the design process are as follows.

• The US-APWR has four divisions of safety systems, and each division is segregated with a physical fire barrier so as to protect the safety function of those safety systems from the fire.

The fire PRA considered the US-APWR fire protection design that is effective to protect a fire within a single division, and therefore its risk is negligible in the event of a postulated fire.

• The cable routes connecting yard transformers to the SWGR within the turbine building have a high risk because the fire frequency of the turbine building is high and its fire severity is large. Therefore, the cable route has been designed to pass through the outside of the turbine building.

• In the US-APWR, Alternate AC power sources (AAC) composed of gas turbine generators are back-up power sources for the emergency gas turbine generators and act as countermeasures against common cause failures. The AAC power sources and switchover panes are not located in turbine building.

The following subsections describe the internal fires risk evaluation and its results.

19.1.5.2.1 Description of the Internal Fires Risk Evaluation The fire PRA methodology for the US-APWR is based on NUREG/CR-6850 (Reference 19.1-7). This methodology and related data were developed jointly by EPRI and the NRC. NUREG/CR-6850 provides a state-of-the-art methodology for fire PRAs.

The fire PRA methodology is composed of 16 tasks, described below.

Tier 2 19.1-87 Revision 23

19. PROBABILISTIC RISK ASSESSMENT US-APWR Design Control Document AND SEVERE ACCIDENT EVALUATION Various assumptions and engineering judgments provide a basis for the internal fire analysis. The assumptions and engineering judgments used in this analysis are as follows:

a. All fire doors provided to the fire barriers between the redundant safety train fire compartments are normally closed, but are opened with the barrier failure probability.

b. For the transient combustibles three airline trash bags has been assumed in each fire compartment

c. There can be only one fire barrier failure and/or one fire damper failure at any given time. Cascading effect will be unimportant because the probability of situation beyond such assumption will be low.

d. It is assumed that, in a fire in MCR, any mitigation systems considered in Level 2 PRA are not available when operators must evacuate from the MCR.

e. It is assumed that, for a Level 2 PRA, firewater pumps can be used as mitigation systems such as reactor cavity direct injection and providing water in containment as spray droplet, even when a fire breaks out.

f. An evaluation has been performed by assuming total damage of the fire-originating fire PRA compartment in consideration of fire heat influence.

In first step, fire compartments have been defined through plant partitioning. And, in next step, the internal events PRA model for the US-APWR has been reviewed to identify the accident sequences that should potentially be included in the fire PRA model, and equipment to be included in the fire PRA component list has been identified. Some of the sequences included in the internal events PRA are eliminated from the fire PRA model. The elimination criteria of the sequences are as follows:

• Sequences associated with initiating events involving a passive/mechanical failure that can generally be assumed not to occur as a direct result of a fire.

Therefore, initiating events that are caused by primary or secondary side pipe breaks, vessel failure, and SGTRs can be eliminated from the PRA model.

• Sequences associated with events that, while it is possible that fire could cause the events, a low-frequency of occurrence argument could be justified. For example, the anticipated transient without scram sequence has not been treated in the fire PRA because fire-induced failures will almost certainly remove power from the control rods (resulting in a trip), rather than cause a failure-to-scram condition. Additionally, fire frequencies multiplied by the independent failure-to-scram probability can be seen as small contributors to fire risk.

Table 19.1-57 provides a listing of the initiating events that were included and excluded in the fire PRA.

As a result, the following accident sequences have been eliminated from the fire PRA model.

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19. PROBABILISTIC RISK ASSESSMENT US-APWR Design Control Document AND SEVERE ACCIDENT EVALUATION

• RVR

• SGTR

• Feed water line break

• Anticipated transient without scram Furthermore, cables associated with fire PRA components have been identified in each fire compartment.

In qualitative screening step, screening of fire scenarios has been performed. A fire scenario is classified into three types: (a) single-compartment fire scenario, (b) multi-compartment fire scenario, and (c) MCR fire scenario. In this step, singe compartment fire scenarios have been studied, and following compartments have been screened. :

• The compartment which does not contain any fire PRA components or cables, and

• The compartment of which fires will not lead to:

An automatic reactor trip A manual reactor trip as specified in fire procedure, EOPs, or plant technical specification However, such information as being contained in fire procedures and EOPs does not exist for the US-APWR at the present stage. Therefore, it has been assumed that every compartment within the reactor building, power source building, and turbine building (T/B) might contain cables which would require manual reactor trip-operation in a fire scenario. And, access control building has been screened from further analysis because those buildings do not contain safety equipment.

In next step, fire ignition frequency has been estimated. Plant ignition sources have been classified in the ignition source specified in Table 6-1 of NUREG/CR 6850. The frequencies are based on fire event experience in the U.S. nuclear power plants prior to December 2000, the same frequencies are used in US-APWR fire PRA. NUREG/CR 6850 also presents the modeling method in which self ignition fire of cables should be postulated in unqualified cables. Therefore, self ignition fire of cable runs has been excluded from the ignition source bins because qualified cables will be adopted in US-APWR.

The oil collection system is provided to collect the lubricating oil that leaks from each RCP motor and route it to a collection tank. Therefore, postulation of a fire resulting from leaked lubrication oil is not credible. This fire scenario has been screened out in the quantitative screening process because of its insignificant effect, even though the fire occurrence uses conservative assumptions.

Tier 2 19.1-92 Revision 23

19. PROBABILISTIC RISK ASSESSMENT US-APWR Design Control Document AND SEVERE ACCIDENT EVALUATION The fire ignition frequencies, fire propagation analysis, fire damage modeling and fire-induced failure mode, and plant response analysis and modeling are evaluated as follows.

(1) Estimation process of fire ignition frequencies Fire ignition frequencies of each fire compartment of the US-APWR are estimated using the methodology and generic fire mean frequencies of NUREG/CR-6850. The estimation process for the fire ignition frequency for each fire compartment of the US-APWR is shown below:

• Mapping Plant Ignition Source to Generic Source In this step, the ignition sources of the US-APWR are assigned to the ignition source bins in Table 6-1 of NUREG/CR-6850.

• Fixed Fire Ignition Source Counts In this step, the types, amounts and location of existing fire ignition sources are identified.

• Calculation of Ignition Source Weighting Factor Using the counts of the preceding step, ignition source weighting factors are calculated for each fire compartment. For transient fires, the weighting factors are estimated using best estimates of anticipated conditions of the US-APWR plant.

• Ignition Source and Compartment Fire Frequency Evaluation Using the ignition source weighting factor of the preceding step and the fire mean frequency of Table 6-1 of NUREG/CR-6850, the fire ignition frequencies associated with each compartment are estimated.

At the design certification stage of the US-APWR, no plant-specific fire event data exist: therefore, the following tasks are not applicable to the US-APWR fire PRA.

• Plant Fire Event Data Collection and Review

• Plant Specific Updates of Generic Ignition Frequencies (2) Fire propagation analysis A fire inside the containment vessel (CV) may spread to multiple fire PRA compartments of the CV because the boundaries of each fire PRA compartment in the CV are not composed of fire-resistant barriers. In this analysis, the CFAST code is used to simulate the fire behavior inside the CV and fire effects in the fire origin compartment and adjacent compartments. According to the results of the CFAST simulation, it is confirmed that there is no fire in the CV that spreads to adjacent fire compartments in the CV. Therefore, the multiple compartments fire scenario for fire compartments in the CV is not developed in this evaluation.

Conditions for CFAS simulation:

• Fire origin compartment is FA1-101-18 (A-Accumulator area),

• Adjacent compartments are FA1-101-15 (B-Accumulator area) and FA1-101-17 (D-Accumulator area),

• Heat release rate given in Chapter 11 of NUREG/CR-6850 is used,

• Damage temperature of the thermoplastic cable shown in Appendix H of NUREG/CR-6850 is applied.

Tier 2 19.1-93 Revision 23

19. PROBABILISTIC RISK ASSESSMENT US-APWR Design Control Document AND SEVERE ACCIDENT EVALUATION (3) Fire damage modeling and fire-induced failure mode The fire-induced failure mode of various components is studied by considering the component status in the normal operation mode and the function required for post-accident mode. The fire-induced failure modes considered in the fire PRA of the US-APWR are as follows:

• Spurious Operation

• Fail to Start/Run

• Fail to Close/Open The failure modes of components with fire-damaged cables or circuits are identified by the detailed circuit failure analysis. The process of the detailed circuit failure analysis in this evaluation is shown below:

• Compile and Evaluate Prerequisite Information and Data In this step, the components and their cables subjected to circuit failure analysis are identified.

• Perform Detailed Circuit/Cable Failure Analysis In this step, a circuit analysis per NUREG/CR-6850 is conducted to establish the possibility of spurious actuation due to fire-induced circuit failure. Typical circuit failure modes described in Figure B3.3 of NFPA 805 and associated circuits shown in Figure B.3.4 of NFPA 805 are used as a reference.

• Generic Equipment Failure Response Analysis In this step, a matrix of fire compartments, fire PRA components in each compartment (including associated cables) and damage states of components due to fire are developed.

(4) Plant response analysis and modeling To evaluate plant response to a fire, the following three groups of fire scenarios are developed:

• Single compartment fire scenario

• MCR (Main Control Room) fire scenario

• Multiple compartments fire scenario For a single compartment fire scenario, it is assumed that a fire would have widespread impact within the concerned compartment, and the fire risk is evaluated by identifying the fire-induced initiating event and the fire mitigation function. For the MCR fire scenario, fire risk is evaluated by considering evacuation to the RSC (Remote Shutdown Console) room from the MCR and shutdown from the RSC. The multiple compartments fire scenario is developed following the steps described in Task 11 of NUREG/CR-6850, and the fire risk of each multiple compartments fire scenario is estimated by assuming that a fire would have widespread impact within the compartments concerned similar to the single compartment fire scenario.

19.1.5.2.2 Results from the Internal Fires Risk Evaluation Quantitative screening has been performed to screen some fire compartments from further analysis.

Tier 2 19.1-94 Revision 23

19. PROBABILISTIC RISK ASSESSMENT US-APWR Design Control Document AND SEVERE ACCIDENT EVALUATION from remote shutdown console. As a result, changes in total value of fire induced CDF is small.

In situ combustibles inside containment vessel are not so much, and therefore, it has been conservatively assumed that transient materials are placed. It has been confirmed through the sensitivity analysis that the amount of combustible materials will not affect to the fire circumstances of fire compartments where redundant safety function have been installed.

The total LRF value of internal fire PRA is approximately twice of the total LRF value of internal events PRA. Additionally, although mitigation features are lost by the fire, CCFP (Conditional Containment Failure Probability) value of internal fire PRA remained in about 0.13 slightly larger than the CCFP value of internal events PRA.

Most significant fire scenario is LOOP due to yard fire, and the LRF value of this fire scenario is about 25 percent of total LRF of internal fire PRA. Fire compartments of the US-APWR significant to fire risk of CDF are the Yard, FA6-101-01 (Turbine building (T/B)

Other Floor), and FA6-101-04 (FA6-101-04 zone). A Yard fire is also the dominant scenario to fire risk of LRF. A fire in the fire compartment "Yard" and a fire in the fire compartment "FA6-101-01" have the potential to cause loss of offsite power (LOOP).

Additionally, simultaneous occurrence of the failures of all Class 1E gas turbine generators start-up and the operator actions to switchover a Class 1E bus to an alternate alternating current (AAC) power sources will lead to a station blackout. In these cases, conditional core damage probability (CCDPs) of these fire scenarios are higher than those of other fire scenarios, and this is the reason why the fires of these fire compartments have become the dominant fire scenarios to core damage. FA6-101-01 contains a large amount of combustible materials compared with other fire compartments, and therefore, the fire frequency of this fire compartment is high. This is the reason why the CDF of this fire compartment is higher than the other fire scenarios, though no mitigation system is damaged in this fire scenario. The second significant fire scenario is TRANS (general transient) due to FA1-101-17 (C/V 3F northwestern part floor zone) fire.

The dominant fire-induced initiating events are LOOP and SLBO. In order to cope with these initiating events, the EFWS (Emergency Feed-Water System) is required to cool down the RCS in order to prevent core damage. When EFWS is not available, the actuation of the SDV (Safety Depressurization Valve) is required in order to bleed the inventory of the RCS. Therefore, the functions of the EFWS and the Pressurizer control system are significant against fire risk.

In this analysis, no credit has been given to any fire suppression system of the fire compartments. The CDF of those fire compartments is sufficiently low. The credit to the Yard fire suppression system has not been given because the detailed information about the feature of the suppression system has not been established at the design certification phase. However, the Yard fire risk will be reduced if the Yard area should be separated by fire barrier walls and/or any fire suppression actions should be taken.

Every fire compartment except for the fire compartments in the containment vessel and in the Yard is composed of a fire resistant wall, floor, and ceiling; therefore, all four ESF trains are segregated individually. The fire PRA identified that there is no significant multiple compartments fire scenario to fire risk in the US-APWR and that the fire risk of Tier 2 19.1-104 Revision 23

19. PROBABILISTIC RISK ASSESSMENT US-APWR Design Control Document AND SEVERE ACCIDENT EVALUATION the multiple compartments fire scenario between the MCR and the Class 1E I&C rooms is not significant. In addition, a fire in any fire compartment in the containment vessel will not spread to adjacent fire compartments by the CFAST simulation.

Significant operator actions of the post fire accident derived from the importance analysis are the action of connecting a Class 1E bus to the AAC in case of a start-up failure of all four Class 1E gas turbine generators. This operator action is important because this action is necessary to cope with a station blackout resulting from LOOP events, which is the dominant fire-inducing initiating event.

Sensitivity analysis has been performed about the fire protection water supply system as mitigation feature for severe accident. As a result, it has been confirmed that the LRF value of internal fire PRA is greatly decreased if the fire protection water supply system can be used.

Electrical room in turbine building has been divided to two fire compartments by the fire barrier. It has resulted in the reduction of the fire risk.

19.1.5.3 Internal Flooding Risk Evaluation The following subsections describe the internal flooding risk evaluation and its results.

19.1.5.3.1 Description of the Internal Flooding Risk Evaluation Internal flooding risk was evaluated using qualitative and quantitative methods, as discussed below. The internal flooding analysis was performed to identify, analyze, and quantify the core damage risk contribution as a result of internal flooding. The internal flooding analysis models potential flood vulnerabilities in conjunction with random failures modeled as part of the internal events PRA. Through this process, flood vulnerabilities that could jeopardize core integrity have been identified.

The internal flooding PRA is organized into three phases. In the first phase of the internal flooding PRA, qualitative evaluation, the information that is needed for the IFPRA is collected and the initial qualitative analysis steps are performed. The four key steps are (1) identification of flood areas and SSCs; (2) identification of flood sources and flooding mechanisms; (3) performance of plant walk downs (alternatively, perform tabletop examination at design certification stage and COL phase); and (4) perform qualitative screening by considering flood source and mode, and flood propagation pathways; and screen out areas free of flood sources, critical equipment, and propagation potential. The major outputs of the first phase include screening of plant flood areas based on criteria associated with flood sources, identifying flood propagation pathways, identifying potential impacts of floods on SSCs, and selecting flood areas for quantitative evaluation.

The second phase is the quantitative evaluation. Quantitative evaluations of plant locations that have not been screened out are addressed in six separate steps. These steps are organized around the key steps in defining flood scenarios and quantifying their impacts in the PRA model in terms of their contributions to CDF and LRF, and entail (1) flood scenario characterization; (2) flood initiating events analysis; (3) flood consequence analysis; (4) flood mitigation evaluation; (5) PRA modeling of flood scenarios; and (6)

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19. PROBABILISTIC RISK ASSESSMENT US-APWR Design Control Document AND SEVERE ACCIDENT EVALUATION In this analysis, no credit has been given to any fire suppression system of the fire compartments. The CDF of those fire compartments is sufficiently low. The credit to the Yard fire suppression system has not been given because the detailed information about the feature of the suppression system has not been established at the design certification phase. However, the Yard fire risk will be reduced if the Yard area should be separated by fire barrier walls and/or any fire suppression actions should be taken.

Every fire compartment except for the fire compartments in the containment vessel and in the Yard is composed of a fire resistant wall, floor, and ceiling; therefore, all four ESF trains are segregated individually. The fire PRA identified that there is no significant multiple compartments fire scenario to fire risk in the US-APWR and that the fire risk of the multiple compartments fire scenario between the MCR and the Class 1E I&C rooms is not significant. In addition, a fire in any fire compartment in the containment vessel will not spread to adjacent fire compartments by the CFAST simulation.

Significant operator actions of the post fire accident derived from the importance analysis are the action of connecting a Class 1E bus to the AAC in case of a start-up failure of all four Class 1E gas turbine generators. This operator action is important because this action is necessary to cope with a station blackout resulting from LOOP events, which is the dominant fire-inducing initiating event.

Sensitivity analysis has been performed about the fire protection water supply system as mitigation feature for severe accident. As a result, it has been confirmed that the LRF value of internal fire PRA is greatly decreased if the fire protection water supply system can be used.

Electrical room in turbine building has been divided to two fire compartments by the fire barrier. It has resulted in the reduction of the fire risk.

19.1.5.3 Internal Flooding Risk Evaluation The US-APWR design features for reducing internal flooding risk and use of the internal flooding PRA in the design process are as follows.

The US-APWR is expected to satisfy the NRC safety goal and to reduce or eliminate known weaknesses of existing operating plants that are applicable to the new design by introducing appropriate features and requirements. The US-APWR has safety-related SSCs in the reactor building (R/B) and power source buildings (PS/B). Therefore, the US-APWR introduced the following design requirements to protect the R/B and PS/B against internal flooding:

• Prevent the flood propagation to multiple mitigation systems (more than two out of four trains of safety systems in the R/B and PS/B) by:

- Separation of the R/B into two areas of an east side and a west side.

- Installation of water-tight doors for the safety-related SSC areas, safety-related electric I&C rooms, and main control room.

-Requirement of the isolation of a flooded essential service water pump within 15 minutes to prevent inflow into the R/B.

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19. PROBABILISTIC RISK ASSESSMENT US-APWR Design Control Document AND SEVERE ACCIDENT EVALUATION

• Prevent inflow into the R/B from adjoining buildings, such as the T/B and A/B.

- Install water-tight doors between the R/B and adjoining buildings.

• Install flood relief panels on the T/B exterior walls to drain the flooded water from the circulating water system to the yard.

The following subsections describe the internal flooding risk evaluation and its results.

19.1.5.3.1 Description of the Internal Flooding Risk Evaluation Internal flooding risk was evaluated using qualitative and quantitative methods, as discussed below. The internal flooding analysis was performed to identify, analyze, and quantify the core damage risk contribution as a result of internal flooding. The internal flooding analysis models potential flood vulnerabilities in conjunction with random failures modeled as part of the internal events PRA. Through this process, flood vulnerabilities that could jeopardize core integrity have been identified.

The internal flooding PRA is organized into three phases. In the first phase of the internal flooding PRA, qualitative evaluation, the information that is needed for the IFPRA is collected and the initial qualitative analysis steps are performed. The four key steps are (1) identification of flood areas and SSCs; (2) identification of flood sources and flooding mechanisms; (3) performance of plant walk downs (alternatively, perform tabletop examination at design certification stage and COL phase); and (4) perform qualitative screening by considering flood source and mode, and flood propagation pathways; and screen out areas free of flood sources, critical equipment, and propagation potential. The major outputs of the first phase include screening of plant flood areas based on criteria associated with flood sources, identifying flood propagation pathways, identifying potential impacts of floods on SSCs, and selecting flood areas for quantitative evaluation.

The second phase is the quantitative evaluation. Quantitative evaluations of plant locations that have not been screened out are addressed in six separate steps. These steps are organized around the key steps in defining flood scenarios and quantifying their impacts in the PRA model in terms of their contributions to CDF and LRF, and entail (1) flood scenario characterization; (2) flood initiating events analysis; (3) flood consequence analysis; (4) flood mitigation evaluation; (5) PRA modeling of flood scenarios; and (6)

PRA quantification. These steps include the definition of flood scenarios in terms of flood initiating events, the consequences of the flood on SSCs, and the interfacing of the flood scenario with the PRA event tree and fault tree logic. Once the scenarios have been properly characterized, this phase also addresses the quantification of the flood initiating event frequency, CDF, and LRF. The last phase, which is the documentation phase, is an ongoing effort that is being performed along with each of the steps noted above for the qualitative evaluation and quantitative evaluation phases.

The scope of the internal flooding risk evaluation is during normal power operations as well as low power or shutdown operations. Reviews of operating experience data show that on the order of one-third of recorded significant internal flooding events have occurred during shutdown operations.

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19. PROBABILISTIC RISK ASSESSMENT US-APWR Design Control Document AND SEVERE ACCIDENT EVALUATION CDF and LRF of this scenario are 1.1E-06/RY and 3.1E-08/RY, respectively. These measures are effective to reduce flooding risk.

US-APWR sets several water tight doors to prevent the propagation of floods. As a bounding sensitivity study, assumed all water barrier doors except the controlled barriers such as R/B separations between east side and west side and high energy compartments are invalid. The CDF and LRF of this bounding study are 2.6E-06/RY and 6.1E-07/RY, respectively. Although the several local watertight doors opened, the increasing of risk is not significant.

Assessment of uncertainties of the internal flood PRA model accounts for uncertainty in initiating events. Table A-13 through Table A-52 of EPRI 1013141 [Reference 19.1-40]

addresses uncertainties in pipe failure rates. Uncertainties in the evaluation of different flood isolation strategies implicitly involve accounting for uncertainties in spill rate distributions, and the time to reach a critical flood volume. Uncertainty is calculated using a Monte Carlo process.

The plant CDF uncertainty range is found to be 4.1E-06/RY - 2.3E-07/RY for the 95% to 5% interval. This uncertainty calculation is considered 95% contribute scenarios of CDF.

• 95th percentile 4.1E-06/RY

• Mean 1.3E-06/RY

• Median 8.1E-07/RY

• 5th percentile 2.3E-07/RY The plant LRF uncertainty range is found to be 6.4E-07/RY - 5.2E-08/RY for the 95% to 5% interval. This uncertainty calculation is considered about 90% contribute scenarios of LRF.

• 95th percentile 6.4E-07/RY

• Mean 2.4E-07/RY

• Median 1.8E-07/RY

• 5th percentile 5.2E-08/RY The most significant rooms to internal flood risk are the turbine driven (T/D) emergency feedwater pump rooms (FA2-102-01 and FA2-108-01) which are installed in water tight Tier 2 19.1-116 Revision 23

19. PROBABILISTIC RISK ASSESSMENT US-APWR Design Control Document AND SEVERE ACCIDENT EVALUATION areas. If a large internal flood occurs in a room, the flood water is assumed to propagate to other areas on the R/B east or west side and cause failure of two safety-related systems. In the other rooms, T/D pumps are available because the rooms are protected by water-tight doors. Main steam and feedwater piping rooms (FA2-102-01 and FA2-108-01) are also significant rooms to internal flood risk. The internal flooding frequencies of these rooms are higher than other rooms because of the many water sources in the piping rooms.

The most significant system to internal flood risk is the emergency feedwater (EFW) system. The risk-significant cutsets of internal flooding are human errors to switch-over the EFW water sources. This is because the risk-significant flood scenarios might possibly affect the SSCs on either side of the R/B, and the failure of two EFW pumps of the affected side is assumed. In this scenario, switch-over of the EFW pit or the realignment of the EFW source to the intact side of the EFW lines is required. When detailed plant-specific information is available, there is determined to be sufficient time to perform these actions.

Significant systems to internal flooding frequencies are the following four systems, the emergency feedwater system (EFWS), main feedwater system (MFWS), main steam system (MSS) and circulating water system (CWS). These systems contain longer runs of piping in the R/B. For EFWS and MFWS, flood frequencies per piping lengths from EPRI 1013141 (Reference 19.1-40) are relatively higher (on the order of 1E-6/yr-ft) than other systems. For CWS and MSS, the numbers of pipes and lengths of piping are relatively higher than for other systems in the R/B, though the flood frequency (on the order of 1E-7/yr-ft) is lower than that for the EFWS and MFWS.

Except for the flood source in the break room adjoining the main control room, the isolation of flood sources by operators is not considered in this assessment. All floors in the R/B are divided into two areas, east and west, by concrete walls and/or water-tight doors. This design mitigates the impact of flooding from one area to safety-related systems in other areas, impacting no more than two of the four trains.

Except for the flood source in the break room adjoining the main control room, flood source isolation actions by operators is not expected in this assessment. The most significant operator action for internal flooding is switch-over of the EFW pit or the realignment of the EFW source to the intact side of the EFW lines. This case occurs when major flooding due to failure of two trains of the EFW system propagates into the R/B east side or west side.

The major contributors to the uncertainty associated with risk estimates are that available specific information--such as pipe routing, pipe lengths, and flooding isolation actions--are limited at the design certification (DC) phase. The risk assessment of US-APWR internal flooding is performed under some conservative assumptions as a bounding analysis. It is expected that the internal flooding risk will be reduced with the plant-specific detailed information.

Based on these risk insights, safety-related equipment is separated as following, so that the risk due to internal flooding is significantly reduced.

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19. PROBABILISTIC RISK ASSESSMENT US-APWR Design Control Document AND SEVERE ACCIDENT EVALUATION C x x x x x The LPSD PRA has estimated that an outage type C "Refueling shutdown" is a representative outage type.

The LPSD operation modes are characterized in 13 plant operation states (POS).

These POSs are identified considering plant configuration, potential of initiating events, and plant responses. The followings are identified POSs for LPSD PRA.

• POS 1: Low power operation POS 1 is a low power operation state. Normal plant shutdown is gradually decreasing a reactor power. The control mode of control rods is switched from automatic operation mode to manual operation mode. The turbine bypass control is also switched from Tavg control mode to steam pressure control mode, and the main feed water control is switched to the bypass control mode. When the turbine output decreases to 5% lower, the turbine is tripped and the control rods are inserted in the reactor fully. The end of POS 1 is defined as the time at which a control rod insertion into the core to shift to a hot standby state.

• POS 2: SG cooling without the RHR cooling POS 2 is a hot standby state transitioning to hot shutdown with core cooling by use of the SGs. Using the turbine bypass valves (and/or the main steam release valve), the RCS is cooled down and de-pressurized from hot standby to hot shutdown. If the RCS is below a pressure of 400psig and a temperature of 350ºF, The RHRS can be used as the RCS cooling system. Therefore, the end of POS 2 is defined as the time of RCS temperature reaching 350ºF.

• POS 3: RHR cooling (RCS is filled with coolant)

POS 3 is a hot shutdown and a cold shutdown state with cooling provided by the RHRS. When the RCS is below a pressure of 400 psig and a temperature of 350ºF, the RHRS starts and cools the RCS. The end of POS 3 is defined as the timing of initiation of a draindown of the RCS because the change of RCS inventories level is the important factor for LPSD PRA.

• POS 4: RHR cooling (mid-loop operation)

POS 4 is a mid-loop operation state with cooling by the RHRS before refueling.

The POS begins at the initiation of the drain down process to the mid-loop water level, which is controlled by the CVCS. To perform the aeration of the RCS and the eddy current test on the SGs, the SG nozzle lids are installed and the upper lid on the RV is removed. The RCS water level is decreased to near the center of the reactor nozzle. Because the RCS inventory is decreasing, the possibility of the RHR pump failure due to the pump cavitations is considered. Also, the time Tier 2 19.1-121 Revision 23

19. PROBABILISTIC RISK ASSESSMENT US-APWR Design Control Document AND SEVERE ACCIDENT EVALUATION required for loss of inventory and subsequent fuel damage is less than for other states in the event of loss of decay heat removal. At the end of POS4, the reactor cavity is filled with water for refueling.

POS 4 or a mid-loop operation is further divided according to the plant states.

The subdivided POSs are shown in Table 19.1-79 and Figure 19.1-13 to Figure 19.1-15.

• POS 4-1: This POS begins at the initiation of the drain down process from the RCS full level (top of the main coolant piping) to the mid-loop water level. The end of POS 4-1 is the time at which the SG manway covers (SG manhole lid) are opened. Decrease of the RCS inventory and maintaining water level are controlled by the CVCS. . In POS 4-1, since the RCS is closed, the reflux cooling by the SGs is available as a heat sink under the vented condition, but the gravitational injection is unavailable because the RCS is not at atmospheric pressure.

• POS 4-2: This POS begins at the end of POS 4-1 and continues until the time that the SG nozzle lid is installed. In POS 4-2, the RCS inventory is kept at the mid-loop water level by the CVCS. Since the RCS is open, the reflux cooling by the SGs is unavailable as a heat sink, but the gravitational injection is available because the RCS is at atmospheric pressure.

• POS 4-3: This period begins at the end of POS 4-2 and continues until the time at which the refueling cavity is filled with water. The RCS inventory is controlled by the CVCS and increased by the RWR and CS/RHR pumps.

The pressurizer safety valves are removed at the beginning of the POS and the RV head is removed during the POS. In the case of POS 4-3, since the RCS is isolated from the SGs, the decay heat removal by the SGs is unavailable.

• POS 5: Refueling cavity is filled with water (refueling)

POS 5 is period when the refueling cavity is filled with water. To offload fuel from the reactor, the refueling cavity is filled with water. If a loss of decay heat removal were to occur, there is considerable time before the reactor core is exposed due to the boil down of coolant. Therefore, the state in which the refueling cavity is filled with water is identified as one of the states of the plant.

The end of POS 5 is defined as the time at which the reactor core is empty.

• POS 6: No fuel in the core or the fuel is partially offloaded POS 6 is the state at which there may be either no fuel in the reactor core or the fuel is partially offloaded. For refueling and examination of fuel, fuel is transported from the RV to the SFP during this POS. This state is excluded from the analysis because there is either no fuel in the reactor, or if the fuel is partially offloaded, there is considerable time before the reactor core is exposed given a loss of decay heat removal event. The end of POS 6 is defined as the time at which fuel is loading into the reactor core.

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19. PROBABILISTIC RISK ASSESSMENT US-APWR Design Control Document AND SEVERE ACCIDENT EVALUATION subsequent fuel damage is less than for other states in the event of loss of decay heat removal.

POS 4 or a mid-loop operation is further divided according to the plant states.

The subdivided POSs are shown in Table 19.1-79 and Figure 19.1-13 to Figure 19.1-15.

• POS 5: Refueling cavity is filled with water (refueling)

POS 5 is period when the refueling cavity is filled with water. To offload fuel from the reactor, the refueling cavity is filled with water. If a loss of decay heat removal were to occur, there is considerable time before the reactor core is exposed due to the boil down of coolant. Therefore, the state in which the refueling cavity is filled with water is identified as one of the states of the plant.

The end of POS 5 is defined as the time at which the reactor core is empty.

• POS 6: No fuel in the core or the fuel is partially offloaded POS 6 is the state at which there ismay be either no fuel in the reactor core or the fuel is partially offloaded. For refueling and examination of fuel, fuel is transported from the RV to the SFP during this POS. This state is excluded from the analysis because there is either no fuel in the reactor., or if the fuel is partially offloaded, there is considerable time before the reactor core is exposed given a loss of decay heat removal event. The end of POS 6 is defined as the time at which fuel is loading into the reactor core.

• POS 7: Refueling cavity is filled with water (refueling)

POS 7 is the state at which the refueling cavity is filled with water. To load new fuel in the reactor, the refueling cavity is filled with water which defines this POS.

If a loss of decay heat removal were to occur, there would be considerable time before the reactor core is exposed by the boiling of coolant. Therefore, the state in which the refueling cavity is filled with water is one of the states of the plant.

The end of POS 7 is defined as the time at which the RCS is drained. The change of RCS inventory level is an important factor for LPSD PRA.

• POS 8: RHR cooling (mid-loop operation after refueling)

POS 8 is a mid-loop state with cooling by the RHRS after refueling. In order to install the upper lid on the RV, and to remove the SG nozzle lids, the RCS water level is decreased to near the center of the reactor nozzle. Because the RCS inventory is decreased, there is a possibility of the RHR pump failure by cavitation and this is considered. Also the time to act to avoid reactor core damage in this state is less than in other states because the RCS inventory is decreased.

POS 8 or a mid-loop operation is further divided according to a plant states. The subdivided POSs are shown in Table 19.1-80 and Figure 19.1-13 to Figure 19.1-15.

• POS 9: Cold shutdown with RHR cooling (RCS is filled with water)

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19. PROBABILISTIC RISK ASSESSMENT US-APWR Design Control Document AND SEVERE ACCIDENT EVALUATION

• POS 7: Refueling cavity is filled with water (refueling)

POS 7 is the state at which the refueling cavity is filled with water. To load new fuel in the reactor, the refueling cavity is filled with water which defines this POS.

If a loss of decay heat removal were to occur, there would be considerable time before the reactor core is exposed by the boiling of coolant. Therefore, the state in which the refueling cavity is filled with water is one of the states of the plant.

The end of POS 7 is defined as the time at which the RCS is drained. The change of RCS inventory level is an important factor for LPSD PRA.

• POS 8: RHR cooling (mid-loop operation after refueling)

POS 8 is a mid-loop state with cooling by the RHRS after refueling. The POS begins at the initiation of the drain down process from the cavity full to the mid-loop water level. The RCS is drained by the CS/RHR pumps, RWR pumps, and CVCS. In order to install the upper lid on the RV, and to remove the SG nozzle lids, the RCS water level is decreased to near the center of the reactor nozzle. Because the RCS inventory is decreased, there is a possibility of the RHR pump failure by cavitation and this is considered. Also the time to act to avoid reactor core damage in this state is less than in other states because the RCS inventory is decreased. At the end of POS 8, the water level in the RCS is at the top of the main coolant piping.

POS 8 or a mid-loop operation is further divided according to a plant states. The subdivided POSs are shown in Table 19.1-80 and Figure 19.1-13 to Figure 19.1-15.

• POS 8-1: This POS begins at the initiation of the drain down process from the cavity full level to the mid-loop water level. The end of POS 8-1 is the time at which the SG nozzle lid is removed. The RV head is removed at the beginning of this POS and the pressurizer safety valves are removed during this POS. The RCS is drained by the CS/RHR pumps and the RWR pump at the early stage of this POS. After the installation of the RV head, the RCS inventory is drained or controlled by the CVCS. In the case of POS 8-1, since the RCS is isolated from the SGs, decay heat removal by the SGs is unavailable.

• POS 8-2: This POS begins at the end of POS 8-1 and continues until the time that the SG manhole lid is closed. In POS 8-2, the RCS inventory is kept at the mid-loop water level by the CVCS. Since the RCS is open, the reflux cooling by SGs is unavailable as a heat sink, but gravity injection is available because the RCS is at atmospheric pressure.

• POS 8-3: This period begins at the end of POS 8-2 and continues until the time at which the RCS inventory is increased up to the top of the main coolant piping. The RCS inventory is controlled by the CVCS. In POS 8-3, air in the RCS is removed and transferred from the SG to the top of the RV head by operating the RCPs. Removal of the air from the RCS results in vacuum refill of the RCS. In POS 8-3, since the RCS is closed, reflux cooling by the SGs is available as a heat sink, Tier 2 19.1-123 Revision 23

19. PROBABILISTIC RISK ASSESSMENT US-APWR Design Control Document AND SEVERE ACCIDENT EVALUATION During shutdown, the RCS is under low or atmospheric pressure. LOCA caused by pipe rupture are unlikely to occur. This initiating event is considered to be caused by loss of cooling water (other than those due to total loss of offsite power), loss of flow and heat exchanger fouling. Only LOCA events that occur by operator error are considered in the PRA of LPSD - an event that would result from the inadvertent transfer of reactor coolant out of the RCS. In this evaluation, inadvertent transfer to the RWSP from the RHR which is caused by operator failure to close the isolation valve (RHS-MOV-025A/B/C/D) after draining the refueling cavity and full-flow test of the RHR pump, is assumed. This diversion can happen if the containment spray/residual heat removal pump full-flow test line stop valves (RHS-MOV-025A/B/C/D) is opened. This event is defined as a loss of all RHR trains.

The frequency of LOCA is evaluated as follow:

• Frequency of plant shutdown for the typical analysis case is 1 shutdown / 2 years =

0.5 events per year assuming a refueling shutdown scheduled every 24 months.

• The frequency of a LOCA is estimated as 5.4E-06/hr, which is referred from Reference 19.1-48.

• The frequency is evaluated for human error. The assumed human errors are either an omission error or a commission error. The failure probability of an omission error, obtained using THERP methodology, is 1.9E-04. The failure probability of a commission error using THERP methodology is 1.3E-05. based on a POS 8-1 duration of 55.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> (Table 19.1-82), the probability of a LOCA during POS 8-1 is:

5.4E-06 x 55.5 = 3.0E-04 Therefore, the frequency of a LOCA during POS 8-1 is:

[0.5 x (1.9E-04 + 1.3E-05)]3.0E-04 = 1.01.5E-04/Y.

The event tree (ET) for the LOCA is shown in Figure 19.1-16. Each top event of this ET is described as follows:

• LOA: Isolation of CS/RHR hot leg suction valves Following a LOCA, isolation of CS/RHR pump hot leg suction by motor-operated valves is expected. Two normally closed motor-operated valves are aligned in series in each of four RHR train suction lines between the RCS and the CS/RHR pump. The failure of this event tree heading is a failure of isolation by manual operation at the MCR.

• MC: RCS makeup by charging pump This mitigation measure represents the RCS inventory makeup by using the charging pumps. When a loss of RCS inventory event occurs, RCS water level is expected to be recovered by charging injection pump. The suction of this pump is VCT. When the level of VCT becomes low signal level, the suction of this pump Tier 2 19.1-128 Revision 23

19. PROBABILISTIC RISK ASSESSMENT US-APWR Design Control Document AND SEVERE ACCIDENT EVALUATION

• SI: High head Injection If injection using the CVCS fails, the borated water in the RWSP is injected into the RCS using the SI pumps to maintain the RCS inventory. It is assumed that loss of this function occurs if the SI pumps fail to start manually or fail to run for the mission time. The SI pumps have to be started manually because the safety injection signal is blocked during shutdown.

• GI: Gravitational injection Gravity injection from the SFP to the RCS is expected if the other mitigation systems fail. The RCS must be at atmospheric pressure. In order for gravity injection to be initiated, it is necessary to operate valves in the injection line and to supply RWSP water to SFP using the refueling water recirculation pumps. This function is unavailable if the RCS is pressurized or if the SG nozzle lid is installed to isolate the SG.

Loss of RHR due to over-drain (OVDR)

This category is loss of RHR operation during mid-loop operation caused by loss of coolant inventory. Two sub-categories are considered. One is OVDR and another is failure to maintain water level (FLML).

The over-drain occurs if the operator fails to stop the drain down process while the RCS is being drained to mid-loop level. It occurs at the beginning of the mid-loop operation POS (POS 4-1 and POS 8-1). This event is defined as loss of all RHR trains.

For the US-APWR, low-pressure letdown line isolation valves are installed. One normally closed air-operated valve is installed in each of two low-pressure letdown lines that are connected to two of four RHR trains. During normal plant cooldown operation, these valves are opened to divert part of the normal RCS flow to the CVCS for purification and the RCS inventory control.

These valves are automatically closed and the CVCS is isolated from the RHRS by the RCS loop low-level signal to prevent loss of RCS inventory at mid-loop operation during plant shutdown.

The initiating frequency of loss of RHR due to OVDR is evaluated as follow:

• Frequency of plant shutdown for the typical analysis case is 1 shutdown / 2 years =

0.5 events per year assuming a refueling shutdown scheduled every 24 months.

• The human error rate for OVDR is evaluated by THERP methodology. The failure probability is 3.0E-03.

• The automatic isolation failure of the low-pressure letdown line is estimated by fault tree (FT) analysis. Two failures are taken into consideration for automatic isolation failure. One is failure of the RCS loop low-level signal, and the other is failure of an air-operated valve to close. The failure probability obtained by quantifying this FT is 2.5E-03.

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19. PROBABILISTIC RISK ASSESSMENT US-APWR Design Control Document AND SEVERE ACCIDENT EVALUATION supply equipment. Following the LOOP, gas turbines, or AAC power attempt to start up and supply ac power. If the gas turbines or AAC power fail to start or run for the required mission, decay heat removal is lost.

• The frequency of a LOOP is estimated as 1.96E-01/Y. This is the frequency of the LOOP per reactor year as described in Reference 19.1-41.

• Based on a POS 8-1 duration of 56 hours6.481481e-4 days <br />0.0156 hours <br />9.259259e-5 weeks <br />2.1308e-5 months <br /> (Table 19.1-82), the probability of a LOOP during POS 8-1 is:

1.96E-01 / 8760 x 56 = 1.25E-03

• The frequency of plant shutdown for the typical analysis case is 1 shutdown / 2 years = 0.5 events per year assuming a refueling shutdown scheduled every 24 months.

Therefore, the frequency of a LOOP during POS 8-1 is:

1.25E-03 x 0.5 = 6.2E-04/Y The ET for the LOOP is shown in Figure 19.1-20. The ET top events are described as follows:

• GT: Power supply by the gas turbine generators The automatic start up of the gas turbine generators is initiated with blackout sequence after the LOOP, and the gas-turbine generators supply electricity to components important for RHR operation.

• SP: Power supply by the gas turbines or AAC power If operation of the gas turbine generators fails, alternate power supply can supply the emergency power. The operation time of the alternate power supply is longer than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. If this function succeeds, it is assumed that sufficient time has elapsed for offsite power to be recovered.

• AC: Offsite power recovery The recovery of the LOOP within an allowable time is considered. The allowable time is assumed to be 1 hourthree hours, based on time until uncover of reactor core by MAAP analysis. The probability that the LOOP duration exceeds sixthree hours is taken as 0.910.17 from Reference 19.1-41.

• PR: CCW pumps / essential service water pumps restart Following blackout sequence, CCW pumps and essential service water pumps automatically start (or re-start) up after power supply to the safety bus is re-established. If this function fails, the mitigation systems to require CCWS are unavailable.

Tier 2 19.1-133 Revision 23

19. PROBABILISTIC RISK ASSESSMENT US-APWR Design Control Document AND SEVERE ACCIDENT EVALUATION 19.1.6.3.1 Seismic at LPSD For seismic consideration, SSCs for LPSD has been involved in Subsection 19.1.5.1 Seismic Risk Evaluation and confirmed that the HCLPFs are greater than or equal to RLE.

On the min-max approach, plant level HCLPF is evaluated considering HCLPFs of initiating events and HCLPFs of event sequences. For the LPSD Seismic SMA, a simplified assumption is made which is that any initiating event will occur as a result of a seismic event. As shown in Table 19.1-120, if the safety injection (SI) system is available, it does not result in core damage for all POSs and initiating events during LPSD. The dominant seismic cutsets for the SI system are as follows:

No. Seismic Cutsets (Description) :HCLPF

1. SE-EPSDLFFGTABCD (Emergency Gas Turbine Generators GTA, B, C, D)

0.50g

2. SE-HPIPMFFSIPABCD (Safety Injection Pumps SIPA, B, C, D)  : 0.62g

3. SE-EPSEPFFBCPABCD (Battery Charger Panels BCP-A, B, C, D)  : 0.75g

4. SE-EPSEPFFIBDABCD (Instrument Power Distribution Panels IBD-A, B, C, D)

0.75g

5. SE-EPSIVFFINVABCD Inverters INVA, B, C, D (Instrument Power Panels)

0.75g Using the min-max method, the HCLPF for SI system is 0.50g.

Key random failures/human errors during LPSD are reviewed. For POS 8-1 and the initiating event LORH, only the SI system is expected to be functional after a seismic event, as noted in Figure 19.1-22.

Dominant random failures/human errors that lead to SI system failure are as follows:

No. Dominant random failures/human errors (Description) :Prob.

1. HPIOO02S (Operators fail to start standby SI pumps) :4.9E-3

2. EPSCF3DLLRDG-ALL (GTG A,B,C fail to load and run after 1hr operation(CCF)*) :1.1E-3

(*: GTG-D is out of service during POS 8-1)

The dominant mixed cutsets are the combinations of seismic failures of non seismic category 1 SSCs and random failures/human errors.

Opening the pressurizer depressurization valve (SDV) in conjunction with SI system operation will be needed for POSs where the RCS is closed, such as POS 3. The need for opening the SDV will not affect the sequence HCLPF since the HCLPF of the SDV is 0.8g, which is greater than the HCLPF for the SI system.

SSCs for LPSD mitigation systems are involved in the list of SSCs for at-power SMA and the HCLPFs of the SSCs are not less than 0.5g.

Tier 2 19.1-159 Revision 23

19. PROBABILISTIC RISK ASSESSMENT US-APWR Design Control Document AND SEVERE ACCIDENT EVALUATION Only during the case of loss of offsite power or loss of CCW events by a seismic event will mitigation systems possibly be unavailable. However, HCLPFs of Class 1E gas turbine generators (GTGs) and the CCWS are also greater than the SI system HCLPF of 0.5g.

Therefore, plant level HCLPF for LPSD will not be less than RLE.

19.1.6.3.2 Internal Fire at LPSD The scope of the internal fire PRA for LPSD at design certification phase focused on mid-loop operations since during these states the plant would be most vulnerable fire such as maintenance-induced fire. POS 8-1(mid-loop operation) is risk significant for the internal event LPSD PRA. For internal fires, risk significant POS 8-1 of LPSD has been estimated using the same methodology at power though the transient fire due to welding and cutting works and access for maintenance works have been specially reflected. The primary focus of the fire scenario development is the potential of fire damage to Yard transformers, RHRS, CVCS and its support system. Possible initiating events by internal fire at LPSD are as follows:

• LOCA

• OVDR (Loss of RHR due to over drain)

• LOOP (Loss of offsite power)

Standby states of mitigation systems for those initiators are shown in Table19.1-83. The states of out of services of POS 8-1 are similar to other POSs so that there are not more severe other POSs than POS 8-1 related to conditions of available mitigation systems.

Therefore POS 8-1 is selected for internal fire at LPSD PRA.

LOCA and LOOP initiating events are potentially significant for all POSs. On the other hand, OVDR and FLWL are initiating events only considered in POSs representing mid-loop operation. Accordingly, LOCA and LOOP are significant in POSs where the RCS is full, while for POS of mid-loop operation, OVDR and/or FLWL are significant event other than LOCA and LOOP. In internal fire PRA for at-power operation, fire in the compartments (e.g. switchyard) that cause LOOP are significant fire scenarios. Similar events are considerably significant during low power and shutdown (Internal events).

The fire-induced pathways and the method for isolating them against LOCA, OVDR and FLML are as below.

(1) The fire-induced pathways of LOCA and the method for isolation The pathways and the method for isolation are as follows.

a. Spurious open of a safety depressurization Valve (SDV)

RCS operating conditions in POS 3 and 11 are under high pressure and high temperature. Therefore, the spurious open of a safety depressurization valve (SDV) due to fire will result in LOCA.

Tier 2 19.1-160 Revision 23

19. PROBABILISTIC RISK ASSESSMENT US-APWR Design Control Document AND SEVERE ACCIDENT EVALUATION 19.1-41 Reevaluation of Station Blackout Risk at Nuclear Power Plants, NUREG/CR-6890, U.S. Nuclear Regulatory Commission, Washington, DC, December 2005.

19.1-42 Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, NUMARC 93-01, Nuclear Energy Institute, July 2000.

19.1-43 10 CFR 50.69 SSC Categorization Guideline, NEI 00-04, Rev. 0, Nuclear Energy Institute, Washington, DC, July 2005.

19.1-44 Risk-Informed Technical Specifications Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies, NEI 04-10, Rev. 1, Nuclear Energy Institute, Washington DC, April 2007.

19.1-45 Policy, Technical, and Licensing Issues Pertaining to Evolutionary and Advanced Light-Water Reactor Designs,. SECY-93-087, U.S. Nuclear Regulatory Commission, Washington, DC, letter issued April 2, 1993 and Staff Requirements Memoranda issued July 21, 1993.

19.1-46 Rates of Initiating Events at U.S. Nuclear Power Plants, 1987-1995.

NUREG/CR-5750, U.S. Nuclear Regulatory Commission, Washington, DC, February 1999.

19.1-47 US-APWR Probabilistic Risk Assessment, MUAP-07030-P Rev. 2 (Proprietary), Mitsubishi Heavy Industries, December 2009.

19.1-48 An Analysis of Loss of Decay Heat Removal Trends and Initiating Event Frequencies (1989-2000), EPRI 1003113 Electric Power Research Institute, Inc., November 2001.

Tier 2 19.1-173 Revision 23

19. PROBABILISTIC RISK ASSESSMENT US-APWR Design Control Document AND SEVERE ACCIDENT EVALUATION Table 19.1-79 Subdivided State of POS 4 (Mid-Loop Operation) for LPSD PRA Open S/G manhole lid Install S/G nozzle lid Remarks RCS Reactor RCS water level Mid-loop (nozzle center) full cavity full POS (POS4-1) (POS4-2) (POS4-3)

RCS open RCS conditions RCS close RCS open SG Isolated Mitigating systems SG and secondary x N/A N/A systems Gravitational N/A x N/A injection Initiating events Over-drain x N/A N/A Over-drain includes the failure Fail to maintain N/A x x to maintain water level water level in POS 4-1.

Tier 2 19.1-820 Revision 23

19. PROBABILISTIC RISK ASSESSMENT US-APWR Design Control Document AND SEVERE ACCIDENT EVALUATION Table 19.1-80 Subdivided State of POS 8 (Mid-Loop Operation) for LPSD PRA Remove S/G nozzle lid Close S/G manhole Lid Remarks Reactor RCS RCS water level Mid-loop (nozzle center) cavity full full POS (POS 8-1) (POS 8-2) (POS 8-3)

RCS open RCS conditions RCS open RCS close SG Isolated Mitigating systems SG and secondary N/A N/A x systems Gravitational N/A x N/A injection Initiating events Over-drain x N/A N/A Over-drain includes failure to Fail to maintain N/A x x maintain water level water level in POS 8-1.

Tier 2 19.1-821 Revision 23

19. PROBABILISTIC RISK ASSESSMENT US-APWR Design Control Document AND SEVERE ACCIDENT EVALUATION Table 19.1-81 Disposition of Plant Operating States for LPSD PRA (Sheet 1 of 2)

POS POS Description Reason for model exclusion modeled?

This POS is a low power shutdown state and SI signal is still available. Further, all Low power components will not be planned to be 1 No operation maintenance in this POS. Therefore, the risk of this POS will be included in full power PRA This POS is a hot standby state before RHR cooling and SI signal is still available.

Hot standby Further, all components will not be 2 No condition planned to be maintenance in this POS.

Therefore, the risk of this POS will be included in full power PRA.

RHR cooling 3 Yes N/A (RCS full)

RHR cooling 4 (mid-loop Yes N/A operation)

This POS is the state that refueling cavity is filled with water. Since there is large Refueling cavity inventory water in the cavity, there would 5 is filled with water No be sufficient time by core exposure and (refueling) operator action will be more reliable. CDF during this POS is considered negligible.

This POS is the state of no fuels in the reactor core. Fuels are transported from the RV to the SFP during this POS. In the case of loss of SFP cooling, sufficient time to recover SFP cooling is available because of large coolant inventory in the pool. Therefore, this POS is excluded from the analysis. This POS is the state at which there is either no fuel in the reactor No fuels in the core or the fuel is partially offloaded. For core, or the core 6 No refueling and examination of the fuel, the is partially fuel is transported from the RV to the spent offloaded.

fuel pit during this POS. This state is excluded from the analysis because there is either no fuel in the reactor, or if the fuel is partially offloaded, there is considerable time before the reactor core is exposed given a loss of decay heat removal event.

The end of this POS is defined as the time at which fuel is fully loaded into the reactor core.

Tier 2 19.1-822 Revision 23

19. PROBABILISTIC RISK ASSESSMENT US-APWR Design Control Document AND SEVERE ACCIDENT EVALUATION Table 19.1-119 Key Insights and Assumptions (Sheet 20 of 28)

Key Insights and Assumptions Dispositions

15. Surge line flooding may occur if decay heat removal 5.4.7.2.3.6 function is lost during plant operating states where the 19.2.5 pressurizer manway is the only vapor release pass from the COL 19.3(6)

RCS. Water held up in the pressurizer can erroneous COL 13.5(7) readings of water level indicators measured with reference to the pressurizer. This phenomenon can also prevent gravity injection from the SFP. Measures to prevent accident evolution caused by surge line flooding are important. Adoption of at least one of theboth measures listed below can reduce risk from surge line flooding event.

- Installation of an temporary RCP water level sensor that measure the MCP water level with reference to pressure at the reactor vessel head vent line and cross over leg when the RCS is vented at a high elevation.

- Operational procedures to perform continuous RCS injections when loss of RHR occurs under conditions where the pressurizer manway is the only vapor release pass from the RCS.

The temporary water level will satisfy the following specifications.

Water level can be read outside the containment vessel (CV) in order to be effective during events which involve harsh environment in the CV Tygon tubing monometer will not be used Instrumentation piping diameter will be sufficient enough to prevent delay in response Tier 2 19.1-983 Revision 23

19. PROBABILISTIC RISK ASSESSMENT US-APWR Design Control Document AND SEVERE ACCIDENT EVALUATION Table 19.1-119 Key Insights and Assumptions (Sheet 20 23 of 2328)

Key Insights and Assumptions Dispositions Internal fire design features and insights

1. Fire protection seals are provided for walls, floors, and ceilings, which compose the fire area boundaries divided by four train areas.

1. The US-APWR has four divisions of safety systems and a 9.5.1 physical fire barrier separates each division to prevent a total loss of safety systems against the fire. As a result of the fire PRA, it has been confirmed that functions of the safety system for plant safe shutdown will be ensured in the event of any postulated fire.

2. Turbine building electric rooms are segregated into two 9.5.1 groups by qualified fire barriers. This feature is possible to prevent loss of offsite power by a turbine building fire.

3. In case of LOCA or loss of RHR caused by over drain or 5.4.7.2.2.3 failure of water level maintain by a fire during LPSD, the flow pathway could be isolated by automatic closing of the low pressure letdown line isolation valve.

4. The following design features of the US-APWR will 9.5.1 contribute to reduce the risk of MCR fire.

• The Class 1E I&C system of the US-APWR is digital; therefore, fire damage of Visual Display Units due to a MCR fire would not cause spurious operation of equipment.

• The automatic start-up circuits of Engineered Safety Features (ESF) necessary for plant safe shutdown are installed in safety I&C rooms. Even though a MCR fire might damage the equipment installed in the MCR, ESF equipment could be started automatically by a start-up signal (e.g., ECCS actuation signal) whose circuits will pass through the safety I&C room but not the MCR.

• The MCR and four divisions of the safety I&C rooms are separated individually by a physical fire barrier. Therefore, the fire in the MCR will not propagate to multiple Class 1E I&C rooms simultaneously even though there could be a single failure on the physical separation barrier of those rooms.

• The US-APWR is designed to have a RSC (Remote Shutdown Console) as a backup system to the MCR. It is possible to shutdown the reactor from the RSC in the event that a severe MCR fire forces the operator to evacuate the MCR.

5. It is important to provide an oil collection system for the 9.5.1 RCP motor in order to prevent a fire from occurring. Even if fire could occur on the oil pan of the oil collection system, the effects of fire to the adjoining oil pan area would be Tier 2 19.1-986 Revision 23

19. PROBABILISTIC RISK ASSESSMENT US-APWR Design Control Document AND SEVERE ACCIDENT EVALUATION Key Insights and Assumptions Dispositions negligible because the leaked oil in the oil-pan is collected into a closed tank.

Tier 2 19.1-987 Revision 23

19. PROBABILISTIC RISK ASSESSMENT US-APWR Design Control Document AND SEVERE ACCIDENT EVALUATION Table 19.1-119 Key Insights and Assumptions (Sheet 24 of 28)

Key Insights and Assumptions Dispositions Internal fire risk insights

1. Fire compartments of the US-APWR significant to fire risk 19.1.5.2.2 of CDF are the Yard, turbine building (T/B) other floor, and FA6-101-04 zone. A Yard fire is also the dominant scenario to fire risk of LRF. A fire in the fire compartment "Yard" and a fire in the T/B fire compartment have the potential to cause loss of offsite power (LOOP). Additionally, simultaneous occurrence of the failures of all Class 1E gas turbine generators start-up and the operator actions to switchover a Class 1E bus to an alternate alternating current (AAC) power sources will lead to a station blackout.

In these cases, conditional core damage probability (CCDPs) of these fire scenarios are higher than those of other fire scenarios, and this is the reason why the fires of these fire compartments have become the dominant fire scenarios to CD. T/B contains a large amount of combustible materials compared with other fire compartments, and therefore, the fire frequency of this fire compartment is high. This is the reason why the CDF of this fire compartment is higher than the other fire scenarios, though no mitigation system is damaged in this fire scenario.

2. The dominant fire-induced initiating events are LOOP and 19.1.5.2.2 SLBO. In order to cope with these initiating events, the EFWS (Emergency Feed-Water System) is required to cool down the RCS in order to prevent CD. When EFWS is not available, the actuation of the safety dpressurization valve (SDV) is required in order to bleed the inventory of the RCS. Therefore, the functions of the EFWS and the Pressurizer control system are significant against fire risk.

3. In this analysis, no credit has been given to any fire 19.1.5.2.2 suppression system of the fire compartments. The CDF of those fire compartments is sufficiently low. The credit to the Yard fire suppression system has not been given because the detailed information about the feature of the suppression system has not been established at the DC phase. However, the Yard fire risk will be reduced if the Yard area should be separated by fire barrier walls and/or any fire suppression actions should be taken.

4. Every fire compartment except for the fire compartments in 19.1.5.2.2 the containment vessel and in the Yard is composed of a fire resistant wall, floor, and ceiling; therefore, all four ESF trains are segregated individually. The fire PRA identified that there is no significant multiple compartments fire scenario to fire risk in the US-APWR and that the fire risk of Tier 2 19.1-989 Revision 23

19. PROBABILISTIC RISK ASSESSMENT US-APWR Design Control Document AND SEVERE ACCIDENT EVALUATION Key Insights and Assumptions Dispositions the multiple compartments fire scenario between the MCR and the Class 1E I&C rooms is not significant. In addition, a fire in any fire compartment in the containment vessel will not spread to adjacent fire compartments by the CFAST simulation.

5. Significant operator actions of the post fire accident derived 19.1.5.2.2 from the importance analysis are as follows:

• The operator action of connecting a Class 1E bus to the AAC in case of a start-up failure of all four Class 1E gas turbine generators. This operator action is important because this action is necessary to cope with a station blackout resulting from LOOP events, which is the dominant fire-inducing initiating event.

• The operator action of the "feed and bleed operation".

Tier 2 19.1-990 Revision 23

Table 19.1-120 Initiating Events and Mitigation Systems during LPSD Tier 2 Initiating Event Mitigation Systems Identifier POS Description LO MC RH SG SI CV GI Loss of RHRS due to OVDR X (1) (2) (3) X (4) POS 4-1 and POS 8-1 Over-drain Loss of RHRS Caused by FLML Failing to Maintain Water X (1) (2) (3) X (4) (5) POS 4 and POS 8

19. PROBABILISTIC RISK ASSESSMENT Level LOCA Loss of Coolant Accident X (1) (2) (3) X (4) (5) All POSs Loss of RHRS Caused by LORH (3) X (4) (5) All POSs Other Failures LOCS Loss of CCWS/ESWS (7) (5) All POSs X (3) X (4) (5)

LOOP Loss of Offsite Power All POSs (6) (6) (6) (6) (6)

(Notes)

AND SEVERE ACCIDENT EVALUATION X: The system would be functional during and after a seismic event.

19.1-990 (1) MC is assumed to be non-functional due to a seismic event since the refueling water auxiliary tank is not Seismic Category I.

(2) Failure of MC would lead to loss of RH.

(3) SG is not available during POS4-2, 4-3, 8-1and 8-2.

(4) CV is assumed to be non-functional due to a seismic event since the refueling water auxiliary tank is not Seismic Category I.

(5) GI is assumed to be non-functional due to a seismic event since the refueling water recirculation pumps to provide boric water from RWSP to the spent fuel pits are not Seismic Category I.

(6) In order to operate mitigating systems, GT/G is required to start and run after loss of offsite power.

(7) The plant has a seismic margin for seismically induced loss of CCWS/ESWS since the seismic capacity of CCWS/ESWS is higher than review level earthquake (Acronyms)

LO (Isolation of Letdown Line), MC (RCS Makeup by Charging Pumps), RH (Decay Heat Removed from the RCS by the RHRS on Standby), SG(Decay Heat Removed from the RCS via SGs), SI (High Head Injection), CV (Injection by Chemical and Volume Control System), GI (Gravitational Injection)

US-APWR Design Control Document Revision 23

19. PROBABILISTIC RISK ASSESSMENT US-APWR Design Control Document AND SEVERE ACCIDENT EVALUATION Table 19.1-119 Key Insights and Assumptions (Sheet 21 25 of 2328)

Key Insights and Assumptions Dispositions Internal flood design features and insights

1. East side and west side of reactor building are physically 3.4.1.3 separated by flood propagation preventive equipment and the connections are kept closed and locked.

1. Four redundant safety systems are located in the reactor 1.2.1.7.2.1 building (R/B). Each safety system is separated into four divisions by physical barriers to assure that the functions of the systems are maintained in the event of a postulated incident.

2. Areas between the reactor building and the turbine building are physically separated by flood propagation prevention equipment.

2. The R/B consists of a radiological controlled area (RCA) 3.4.1.3 and a non-radiological controlled area (NRCA) separated physically by concrete barrier walls. These concrete barrier walls are designed to preclude flooding between the RCA and the NRCA. Piping, instrumentation, HVAC duct, conduit, and cable trays are installed through the flood barrier wall above the maximum flood level or are provided with water-tight seals.

3. The flood barriers that separate the reactor building 3.4.1.3 between east side and west side are important to safety for 19.2.5 the operation of the facility. These doors should be COL 19.3(6) monitored and controlled during plant operation and COL 19.5(1) maintenance. COL 13.5(7)

3. All floors in the RCA of the R/B are divided into two areas, (RAI 19-207) east and west, by concrete walls with water-tight doors. 3.4.1.5.2.1 The equipment in the east area of the RCA is the A and B train SI pumps and the A and B train CS/RHR pumps with heat exchangers. The equipment in the west area are the C and D train SI pumps and the C and D train CS/RHR pumps with heat exchangers. The concrete walls and the water-tight doors prevent flood water migration from one safety train to another. The floor drains of the east area are connected and drain into the A-R/B sump tank, and the floor drains of the west area are connected and drain into the B-R/B sump tank. There is no cross-connection between the east area drains and the west area drains.

4. All floors in the NRCA of the R/B are divided into the two 3.4.1.5.2.2 areas, east and west, by concrete walls with water-tight doors. The equipment on the east side includes two trains (A and B) of the CCW (HX and pump rooms), two trains (A and B) of the EFW (pump rooms), and the A and B train Class 1E electrical panels. The equipment on the west side includes two trains (C and D) of the CCW (HX and pump Tier 2 19.1-991 Revision 23

19. PROBABILISTIC RISK ASSESSMENT US-APWR Design Control Document AND SEVERE ACCIDENT EVALUATION Key Insights and Assumptions Dispositions room), two trains (C and D) of the EFW (pump rooms) and the C and D train Class 1E electrical panels. The Class 1E electrical panel rooms are isolated from the corridor by concrete walls and water-tight doors. The floor drains of the east areas are connected and drain into the A-R/B non-radioactive sump, and the floor drains of the west areas are connected and drain into the B-R/B non-radioactive sump. There is no cross-connection between the east area drains and west area drains.

5. The T/B adjoins the NRCA of the R/B. The T/B is 1.2.1.7.2.1 independent of the R/B to prevent internal hazards in the 3.4.1.3 T/B from propagating to the R/B. Water-tight doors are installed in the doorways at ground level between the T/B and the R/B. In addition, a flood relief panel system is built into the T/B exterior walls. Actuation of the flood relief panels allows the flood water to drain out to the yard area to prevent it from affecting R/B equipment.

Tier 2 19.1-992 Revision 23

19. PROBABILISTIC RISK ASSESSMENT US-APWR Design Control Document AND SEVERE ACCIDENT EVALUATION Table 19.1-119 Key Insights and Assumptions (Sheet 26 of 28)

Key Insights and Assumptions Dispositions Internal flood risk insights

1. The risk significant flooding areas are A-turbine driven 19.1.5.3.2 (T/D) emergency feedwater (EFW) pump room, D-T/D EFW pump room, and main steam and feedwater piping room (east), and main steam and feedwater piping room (west).

Those four areas contribute about 80% of the total CDF.

The T/D EFW pump rooms which are installed in water tight areas . If a large internal flood occurs in the room, the flood water is assumed to propagate to other areas on the R/B east or west side and cause failure of two safety-related systems. In the case of flooding in other rooms, T/D pumps are available because the rooms are protected by water-tight doors. Main steam and feedwater piping rooms are also significant rooms to internal flood risk. The internal flooding frequencies of these rooms are higher than other rooms because of the many water sources in the piping rooms.

2. The most significant system to internal flood risk is the 19.1.5.3.2 emergency feedwater (EFW) system. The risk-significant cutsets of internal flooding are human errors to switch-over the EFW water sources. This is because the risk-significant flood scenarios might possibly affect the SSCs on either side of the R/B, and the failure of two EFW pumps of the affected side is assumed. In this scenario, switch-over of the EFW pit or the realignment of the EFW source to the intact side of the EFW lines is required. When detailed plant-specific information is available, there is determined to be sufficient time to perform these actions.

3. Significant systems to internal flooding frequencies are the 19.1.5.3.2 following four systems, the emergency feedwater system (EFWS), main feedwater system (MFWS), main steam system (MSS) and circulating water system (CWS). These systems contain longer runs of piping in the R/B.

For EFWS and MFWS, flood frequencies per piping lengths from EPRI 1013141 are relatively higher (on the order of 1E-6/yr-ft) than other systems. For CWS and MSS, the numbers of pipes and lengths of piping are relatively higher than for other systems in the R/B, though the flood frequency (on the order of 1E-7/yr-ft) is lower than that for the EFWS and MFWS.

4. Except for the flood source in the break room adjoining the 19.1.5.3.2 main control room, the isolation of flood sources by operators is not considered in this assessment. All floors in the R/B are divided into two areas, east and west, by concrete walls and/or water-tight doors. This design Tier 2 19.1-994 Revision 23

19. PROBABILISTIC RISK ASSESSMENT US-APWR Design Control Document AND SEVERE ACCIDENT EVALUATION Key Insights and Assumptions Dispositions mitigates the impact of flooding from one area to safety-related systems in other areas, impacting no more than two of the four trains.

5. Except for the flood source in the break room adjoining the 19.1.5.3.2 main control room, flood source isolation actions by operators is not expected in this assessment. The most significant operator action for internal flooding is switch-over of the EFW pit or the realignment of the EFW source to the intact side of the EFW lines. This case occurs when major flooding due to failure of two trains of the EFW system propagates into the R/B east side or west side.

6. The major contributors to the uncertainty associated with 19.1.5.3.2 risk estimates are that available specific information--such as pipe routing, pipe lengths, and flooding isolation actions--are limited at the design certification phase. The risk assessment of US-APWR internal flooding is performed under some conservative assumptions as a bounding analysis. It is expected that the internal flooding risk will be reduced with the plant-specific detailed information.

Tier 2 19.1-995 Revision 23

Tier 2

19. PROBABILISTIC RISK ASSESSMENT (POS8-1)

LORH SG SI CV GI 1 OK 2 OK 1.0 (Note 1) 3 OK Note 1 - SG is not available during POS 8-1.

1.0 (Note 2) 4 OK 1.0 (Note 3) Note 2 - CV is assumed to be non-functional due to a seismic event AND SEVERE ACCIDENT EVALUATION 5 CD since the refueling water auxiliary tank is not Seismic Category I.

19.1-1130 Note 3 - GI is assumed to be non-functional due to a seismic event Event Description LORH Loss of RHRS caused by other failures since the refueling water recirculation pumps to provide boric SG  : Decay heat removed from the RCS via SGs water from RWSP to the spent fuel pits are not Seismic SI  : High head injection CV  : Injection by the CVCS Category I.

GI  : Gravitational injection Figure 19.1- 22 Loss of RHRS caused by Seismic Event Tree US-APWR Design Control Document Revision 23

19. PROBABILISTIC RISK ASSESSMENT US-APWR Design Control Document AND SEVERE ACCIDENT EVALUATION Accident progression analyses for hydrogen generation and control utilizing the hydrogen ignition system have been performed using GOTHIC code. In the developed GOTHIC model, hydrogen igniters are located at 20 locations in the containment and are modeled to initiate hydrogen burning when hydrogen concentration becomes greater than 8% by volume except under steam inert condition.

Hydrogen concentration in each compartment is either lower than 10% or the compartment is inerted by steam. The pressure in containment vessel is kept below 68 psia, and this pressure is much lower than the containment ultimate pressure 216 psia described in Subsection 19.2.4. For several sequences especially when RWSP water is not utilized for decay heat removal, it is identified that the hydrogen concentration in the RWSP may accumulate greater than 10% long after the initiation of an accident. Under such situation, operator action to inject firewater into the containment is performed. The RWSP will be filled with water, and accordingly this operation will inherently prevent high hydrogen concentration in the RWSP. Therefore, the containment integrity is maintained against hydrogen combustion events, and the requirements of 10 CFR 50.44(c)(1),

10 CFR 50.34(f)(2)(ix), 10 CFR 50.44(c)(2), 10 CFR 50.34(f)(3)(v) (A)(1), and 10 CFR 50.44(c)(5) are therefore met.

The maximum pressure in the containment vessel under the adiabatic isochoric complete combustion condition is127 psia. This pressure is lower than the containment ultimate pressure 216 psia and the requirement of 10 CFR 50.44(c)(5) is met.

19.2.3.3.3 Core Debris Coolability The fundamental design concept of the US-APWR for severe accident termination is reactor cavity flooding and cool down of the molten core by the flooded coolant water.

Therefore, dependable systems are provided to properly flood the reactor cavity during a severe accident. The US-APWR provides a diverse reactor cavity flooding system, which consists of the CSS with a drain line from the SG compartment to the reactor cavity and firewater injection to the reactor cavity. The CSS is automatically activated when the high-high containment pressure is detected and P-signal is transmitted. This containment spray water flows into the reactor cavity from the SG compartment through the drain line by gravity. The fire protection water supply system is provided outside of containment and in stand-by status during normal operation. The system line-up is modified for emergency operation during a severe accident and provides firewater from outside to the reactor cavity. These two systems are independent and thus provide high reliability reactor cavity flooding.

MCCI is a phenomenon that occurs when the temperature of core debris exceeds the melting temperature of concrete, and concrete is gradually eroded by high-temperature core debris resulting in potential basemat melt-through. Therefore, the primary mitigation of MCCI is cool down of core debris that has been relocated from RV to the reactor cavity. The US-APWR provides a highly reliable reactor cavity flooding system as discussed above, and coolant water is continuously supplied during a severe accident.

The reactor cavity floor concrete, which has a thickness of 40 in., provides a protection against direct attack to the steel liner plate by the relocated core debris. This steel liner plate underneath the reactor cavity floor concrete is the pressure boundary between containment and the environment.

Tier 2 19.2-8 Revision 23

19. PROBABILISTIC RISK ASSESSMENT US-APWR Design Control Document AND SEVERE ACCIDENT EVALUATION spray header and provides water as spray droplet. This operation temporarily depressurizes containment however the fire protection water supply system does not contain a heat exchanger, and thus has no ability to remove heat from containment to terminate the containment pressurization.

• Accident management of prevention of early containment failure is through prevention of containment bypass, HPME and hydrogen detonation. RCS depressurization is in order for prevention of HPME and temperature-induced SGTR. When core damage is detected, severe accident dedicated depressurization valve is opened and if necessary safety depressurization valve is opened. In case water supply to SG is available, main steam depressurization valve is opened to enhance primary system cooling and depressurization if needed. Water supply to SG is recovered or controlled to avoid FP release due to temperature induced SGTR through secondary system, also to depressurize RCS. Main feedwater system or emergency feedwater system are employed for this function and operation is required when SG water level decreases below a criterion if available. Combustible gas control is in order to prevent containment failure especially due to hydrogen detonation. Although the combustible gas control is automatically achieved by hydrogen ignition system, in case CSS fails and containment vessel atmosphere is kept inerted for certain duration, CSS recovery or operation of alternate containment cooling may lead containment vessel atmosphere to combustible condition under high hydrogen concentration.

In such case containment depressurization is suspended at a relatively high containment pressure. It is widely known that the low inert limit of steam concentration is approximately 55% and the low flammability limit of hydrogen concentration is approximately 4%. Hydrogen impact when depressurizing containment is evaluated and a material, such as a map of hydrogen concentration vs. containment pressure to show if hydrogen burn is safe or potential danger, is prepared to support the containment depressurization operation. MCR alarm for hydrogen concentration is also provided through the containment hydrogen monitoring system when the hydrogen concentration reaches 4% and 8%. The control room operators are required to carefully monitor the condition of containment.

• Firewater can be utilized to fill the RWSP in the case no decay heat removal function is available. This will eliminate the possibility of high hydrogen concentration in the RWSP.

(During LPSD operations)

It is likely that containment is not isolated during LPSD operations in order for various maintenance activities. The accident management functions to maintain containment integrity during LPSD include firstly recovery of containment isolation from the environment, and secondary heat removal from the isolated containment. However, the ability to close the containment and to recover heat removal without ac power is minimal and may not be possible. It is evaluated for the LPSD PRA that the losses of offsite power contribute approximately 30% of shutdown risk in total. As a result any period in which the RCS level is low should be planned to be undertaken with maximum confidence in offsite and onsite power reliability. Maintenance activities in the switchyard are minimal or Tier 2 19.2-36 Revision 23

US-APWR DCD Revision 2 RAI Tracking Report MUAP-09026(R5)

Tier 1 Mitsubishi Heavy Industries, LTD.

US-APWR DCD Tier 1 Rev. 2, Tracking Report Rev. 5 Change List Location (e.g., subsection with Page paragraph/sentence/it Description of Change em, table with column/row, or figure) 2.6-43 Table 2.6.5-1, ITAAC Revised to add a site acceptance test for as-built AAC power

1. 12 sources.

RAI 582, 09.04.01-20 2.7-259, 260 2.7.6.5.1, Key Design Revised to more specifically address single-failure-proof Features features.

RAI 563, 09.01.05-16 RAI 616, 09.01.05-18 2.7-260 2.7.6.5.1, System Added the following text to the end of the last bullet.

Operation The equipment hatch hoist lifts the containment equipment hatch vertically to secure the hatch while the containment is open.

RAI 616, 09.01.05-18 2.7-261 2.7.6.5.1, Numeric Deleted the description for the limits on drop distance due to Performance Values axle failure and maximum stopping distance.

Added equipment hatch hoist in the first sentence.

RAI 563, 09.01.05-16 RAI 616, 09.01.05-18 2.7-264, 265, Table 2.7.6.5-1, Revised to identify specific design features and actions to 266 ITAAC #2.c demonstrate the as-built PCCV polar crane main and auxiliary hoist, equipment hatch hoist and spent fuel cask crane main hoist are single-failure proof.

RAI 563, 09.01.05-16 RAI 616, 09.01.05-18 2.7-266, 267 Table 2.7.6.5-1, Revised to add the 150% acceptance test and NDE of critical ITAAC #2.d welds in accordance with ANSI N14.6 for the special lifting devices.

RAI 563, 09.01.05-16 RAI 616, 09.01.05-18

2.6 ELECTRICAL SYSTEMS US-APWR Design Control Document Table 2.6.5-1 AAC Systems Inspections, Tests, Analyses, and Acceptance Criteria (Sheet 2 of 2)

Inspections, Tests, Design Commitment Acceptance Criteria Analyses

8. The operation (e.g. start, 8. An inspection of the as-built 8. The operation (e.g. start, stop and synchronization) of MCR will be performed. stop and synchronization) the AAC power sources are of the AAC power sources provided in the MCR. are provided in the as-built MCR.

9. Each AAC power source is 9. A test will be performed to 9. Each as-built AAC power capable of providing power verify that the as-built AAC source is capable of at the set voltage and power source can reach set providing power at the set frequency to the non Class voltage and frequency. voltage and frequency to 1E 6.9kV buses within the the non Class 1E 6.9kV maximum allowable time buses within 100 seconds from receiving a start signal. from receiving a start signal.

10. Each AAC power source 10. An inspection of the as-built 10. Each as-built AAC power status and the breaker MCR will be performed. source status and the status of each Class 1E breaker status of each 6.9kV breaker are displayed Class 1E 6.9kV breaker in the MCR. are displayed in the as-built MCR.

11. The functional arrangement 11. An inspection of the 11. The as-built AAC fuel oil of the AAC fuel oil storage functional arrangement of storage and transfer and transfer system is as the as-built AAC fuel oil system conforms to the described in Subsection storage and transfer system functional arrangement as 2.6.5.2. will be performed. described in Subsection 2.6.5.2.

12. The reliability of the AAC 12.i An analysis of the reliability 12.i The reliability of the as-power sources meet or of the as-built AAC power built AAC power sources exceed 95 percent and the sources will be performed. meet or exceed 95 AAC power sources provide percent.

the required capability.

12.ii A site acceptance test will 12.ii The as-built AAC power be performed to sources provide the demonstrate the capability required capability.

of the as-built AAC power sources to perform required function.

Tier 1 2.6-43 Revision 32

2.7 PLANT SYSTEMS US-APWR Design Control Document 2.7.6.5 Overhead Heavy Load Handling System 2.7.6.5.1 Design Description System Purpose and Functions The purpose and function of the overhead heavy handling system (OHLHS) is to move heavy loads. For the US-APWR, a heavy load is defined as any load greater than approximately 2450 lbs. The OHLHS is non-safety related.

Location and Functional Arrangement The OHLHS exists in the reactor building, specifically the fuel storage and handling area, and in the pre-stressed concrete containment vessel (PCCV) of the reactor building.

The functional arrangement and design characteristics of the OHLHS are discussed below.

Key Design Features Key design features of the OHLHS include:

The primary equipment used in the OHLHS are the spent fuel cask handling crane in the fuel handling area, equipment hatch hoist in the PCCV and the polar crane in the PCCV. The main hoist of the spect fuel cask handling crane, the equipment hatch hoist, and the main and auxiliary hoists of the polar crane are designed as single-failure-proof. The hoisting systems consist of reeving, wire rope, hoisting mechanisms, and hooks. Reeving systems of the single-failure-proof cranes are designed such that a single rope failure will not result in a load drop. Each single-failure-proof crane is provided with at least two holding brakes.

Special lifting devices and slings used for critical load handling operations in conjunction with these cranes have dual load paths or double safety factors.

The spent fuel cask handling crane has three load handling hookshoists: the main, the auxiliary, and the suspension hoist.

The suspension hoist is only used for new fuel assembly handling between a new fuel container to the new fuel storage area or between the new fuel storage rack and the basket on the new fuel elevator. Because of this limitation, the suspension hoist is considered part of the Light Load Handing System (LLHS)

(Subsection 2.7.6.4).

The polar crane has a seismic restraint system which precludes derailment of either the hoist trolley or the main bridge box girders during a seismic event.

The main hooks of the PCCV polar crane main and auxiliary hoist, equipment hatch hoist and the spent fuel cask handling crane main hoist are designed as single-failure-proof cranes, and are subject to the following:

Tier 1 2.7-259 Revision 32

2.7 PLANT SYSTEMS US-APWR Design Control Document

1. Static load testing at a minimum of 125 % rated load

2. Dynamic testing to lift, transport, lower, stop and hold a test load of at least 100% of rated load. Each holding brake is capable of stopping and holding a minimum of 100 % rated load.

3. No-load testing to verify limit switches, interlocks and stops are properly adjusted and set.

4. Non-destructive examination of critical welds.

Special lifting devices and slings used for critical load handling operations in conjunction with these cranes have dual load paths or double safety factors.

Special lifting devices used in conjunction with the PCCV polar crane and spent fuel cask handling crane main hoist during critical load handling operations are subject to a load test followed by NDE of critical welds.

The equipment hatch hoist is base mounted to its support, which is designed to seismic category II requirements. The hoist supports are supported off the side of the containment.

Seismic and ASME Code Classifications The OHLHS is seismic Category II.

System Operation The OHLHS operation includes:

A spent fuel cask filled with spent fuel assemblies is lifted and transferred using the main hoist of the spent fuel cask handling crane and the spent fuel cask lift rig.

During refueling, the reactor vessel head assembly and the upper and lower reactor internals are transferred using the polar cranes main hook and a lifting rig.

Reactor coolant pump motors and other similar sized equipment are transferred using the polar cranes auxiliary hook.

The equipment hatch hoist lifts the containment equipment hatch vertically to secure the hatch while the containment is open.

Alarms, Displays, and Controls There are no main control room alarms, displays, or controls associated with the OHLHS.

Tier 1 2.7-260 Revision 32

2.7 PLANT SYSTEMS US-APWR Design Control Document Logic Not applicable.

Interlocks The OHLHS is equipped with mechanical and electrical limit devices to disengage power to the motors as the load hook approaches its travel limits or to prevent damage to other components when continued operation would potentially damage the OHLHS.

The control system includes safety devices which assure that the OHLHS returns to and/or maintains a secure holding position of critical loads in the event of a system fault.

Class 1E Electrical Power Sources and Divisions Not applicable.

Equipment to be Qualified for Harsh Environments Not applicable.

Interface Requirements There are no safety-related interfaces with systems outside of the certified design.

Numeric Performance Values The PCCV polar crane, equipment hatch hoist and the spent fuel cask handling crane are designed as single-failure-proof to prevent uncontrolled lowering of heavy loads.

Therefore, no load drop accident analysis is required. Crane axle failure may result in limited slip of the lifted load, causing impact on the floor, which has been accounted for in the structural design.

Tier 1 2.7-261 Revision 32

2.7 PLANT SYSTEMS US-APWR Design Control Document Design Commitment Inspections, Tests, Analyses Acceptance Criteria 2.c.i The PCCV polar crane and 2.c.i A combination of 2.c.i A report exists and the spent fuel cask handling inspection, tests and/or concludes that the as-built crane main hook are analyses will be PCCV polar crane and the designed asis single-failure- performed on the as-built spent fuel cask handling proof cranes. polar craneOHLHS. crane main hook areis The PCCV polar crane: single failure proof.

1. reeving system The as-built PCCV polar design precludes a crane:

load drop in the event 1. can tolerate a single of a single rope reeving system rope failure. failure without load

2. is equipped with at drop least two holding 2. is equipped with two brakes. holding brakes, each of

3. will be static load which are set and rated tested at a minimum at a minimum torque of of 125% of rated load. 125 % of rated hoisting torque at the point of

4. will be dynamically brake application.

tested at a minimum of 100% of rated load. 3. can withstand a static load of at least 125% of

5. will be no-load tested rated load.

to include verification of limit switch, 4. can lift, transport, interlock and stop lower, stop and hold a settings. test load of at least 100% of rated load.

1.6. critical welds will be Each polar crane hoist subject to non- holding brake is destructive capable of stopping examination (NDE). and holding a minimum of 100% rated load.

5. limit switches, interlocks and stops are properly adjusted and set.

1.6. critical welds meet ASME NOG-1 criteria for NDE.

Tier 1 2.7-264 Revision 32

2.7 PLANT SYSTEMS US-APWR Design Control Document Design Commitment Inspections, Tests, Analyses Acceptance Criteria 2.c.ii The spent fuel cask handling 2.c.ii A combination of 2.c.ii A report exists and crane main hoist is single- inspection, tests and concludes that the as-built failure-proof. analyses will be spent fuel cask handling performed on the as- crane main hoist is single built spent fuel cask failure proof.

handling crane main The as-built spent fuel hoist. cask handling crane main The spent fuel cask hoist:

handling crane main 1. can tolerate a single hoist: reeving system rope

1. reeving system failure without load design precludes a drop load drop in the event 2. is equipped with two of a single rope holding brakes, each of failure. which are set and rated

2. is equipped with at at a minimum torque of least two holding 125 % of rated hoisting brakes. torque at the point of

3. will be static load brake application.

tested at a minimum 3. can withstand a static of 125% of rated load. load of at least 125% of

4. will be dynamically rated load.

tested at a minimum 4. can lift, transport, of 100% of rated load. lower, stop and hold a 1.5. will be no-load tested test load of at least to include verification 100% of rated load.

of limit switch, Each polar crane hoist interlock and stop holding brake is settings. capable of stopping and holding a minimum

6. critical welds will be of 100% rated load.

subject to non-destructive 1.5. limit switches, examination (NDE). interlocks and stops are properly adjusted and set.

6. critical welds meet ASME NOG-1 criteria for NDE.

Tier 1 2.7-265 Revision 32

2.7 PLANT SYSTEMS US-APWR Design Control Document Design Commitment Inspections, Tests, Analyses Acceptance Criteria 2.c.iii The equipment hatch hoist is 2.c.ii A combination of 2.c.ii A report exists and single-failure-proof. inspection, tests and concludes that the as-built analyses will be equipment hatch hoist is performed on the as- single failure proof.

built equipment hatch The as-built equipment hoist. hatch hoist:

The equipment hatch 1. can tolerate a single hoist: reeving system rope

1. reeving system failure without load design precludes a drop load drop in the event 2. is equipped with two of a single rope holding brakes, each of failure. which are set and rated

2. is equipped with at at a minimum torque of least two holding 125 % of rated hoisting brakes. torque at the point of

3. will be static load brake application.

tested at a minimum 3. can withstand a static of 125% of rated load. load of at least 125% of

4. will be dynamically rated load.

tested at a minimum 4. can lift, transport, of 100% of rated load. lower, stop and hold a

5. will be no-load tested test load of at least to include verification 100% of rated load.

of limit switch, Each polar crane hoist interlock and stop holding brake is settings. capable of stopping and holding a minimum

6. critical welds will be of 100% rated load.

subject to non-destructive 5. limit switches, examination (NDE). interlocks and stops are properly adjusted and set.

6. critical welds meet ASME NOG-1 criteria for NDE.

2.d.i Special lifting devices and 2.d.i A combination of 2.d.i A report exists and slings used in conjunction inspection, tests and/or concludes that tThe as-built with the PCCV polar crane analyses will be special lifting devices and main and auxiliary hoist, performed on the as-built slings used in conjunction equipment hatch hoist and OHLHS. with the PCCV polar crane the spent fuel cask handling main and auxiliary hoist, crane main hoisthook during equipment hatch hoist and critical load handling the spent fuel cask handling operations have dual load crane main hoisthook paths or double safety during critical load handling factors. operations have dual load paths or double safety factors.

Tier 1 2.7-266 Revision 32

2.7 PLANT SYSTEMS US-APWR Design Control Document Design Commitment Inspections, Tests, Analyses Acceptance Criteria 2.d.ii Special lifting devices used in 2.d.ii A combination of 2.d.ii As-built special lifting conjunction with the PCCV inspection, tests and/or devices used in polar crane main and analyses will be conjunction with the PCCV auxiliary hoist, equipment performed on the as- polar crane main and hatch hoist and spent fuel built OHLHS. auxiliary hoist, equipment cask handling crane main hatch hoist and spent fuel hoist during critical load cask handling crane main handling operations are hoist during critical load subject to a load test handling operations satisfy followed by NDE of critical ANSI N 14.6 criteria for a welds. 150% load test for a minimum of 10 minutes followed by NDE of critical welds.

Tier 1 2.7-267 Revision 32