dcd markups associated with revised design completion …mua 0011 p coolant lazp branh piping,...

AtMITSUBISHI HEAVY INDUSTRIES, LTD.

16-5, KONAN 2-CHOME, MINATO-KUTOKYO, JAPAN

June 28, 2013

Document Control DeskU.S. Nuclear Regulatory CommissionWashington, DC 20555-0001

Attention: Mr. Jeffrey A. CioccoDocket No. 52-021

MHI Ref: UAP-HF-13044

Subject: DCD Markups Associated with Revised Design Completion Plan forUS-APWR Piping Systems and Components

Reference: 1) Letter (ML12346A448) from Y Ogata ("MHI") to U.S. NRC, "RevisedDesign Completion Plan for US-APWR Piping Systems and Components",UAP-HF-12322, dated December 7, 2012

2) Letter from Y. Ogata ("MHI") to U.S. NRC, "MHI's 2 nd Amended Responseto US-APWR DCD RAI No. 945-6452 Revision 3 (SRP14.03)", dated June7, 2013

With this letter, Mitsubishi Heavy Industries, Ltd. ("MHI") transmits to the U.S. NuclearRegulatory Commission ("NRC") DCD markups associated with the "Revised DesignCompletion Plan for US-APWR Piping Systems and Components" (Reference 1).

In Reference 1, MHI proposed to withdraw Technical Reports on stress analysis which hadbeen submitted in support of Sections 3.6 and 3.9 of the DCD. Enclosed are the DCDmarkups that show changes necessitated by the withdrawal of the Technical Reports. Theapplicable DCD Sections and the reasons for the changes are identified in Attachment 1. InReference 1, MUAP-09010, MUAP-09013, and MUAP-11003 were included in the list ofTechnical Reports to be withdrawn. However, based on discussion with the NRC staff onJune 10, 2013, MHI decided to keep those Technical Reports on the docket. Enclosure 1already reflects the discussion, and Reference 1 will be updated accordingly. The DCDmarkups for Subsection 3.6.3.4.13 and 3.6.3.5 in Enclosure 1 supersede the markups for thesame subsections submitted along with MHI's response to RA1945-6452 (Reference 2).

Please contact Mr. Joseph Tapia, General Manager of Licensing Department, MitsubishiNuclear Energy Systems, Inc. if the NRC has questions concerning any aspect of this letter.His contact information is provided below.

- o8

Sincerely,

Yoshiki Ogata,Executive Vice PresidentMitsubishi Nuclear Energy Systems, Inc.On behalf of Mitsubishi Heavy Industries, LTD.

Enclosure:

1. DCD Markups Associated with Revised Design Completion Plan for US-APWR PipingSystems and Components

CC: J. A. CioccoJ. Tapia

Contact InformationJoseph Tapia, General Manager of Licensing DepartmentMitsubishi Nuclear Energy Systems, Inc.1001 19th Street North, Suite 710Arlington, VA 22209E-mail: joseph tapia@m nes-us.comTelephone:(703) 908-8055

Docket No. 52-021MHI Ref: UAP-HF-13044

Enclosure 1

UAP-H F-13044Docket No. 52-021

DCD Markups Associated with Revised Design CompletionPlan for US-APWR Piping Systems and Components

June 2013

Markups 7US-APWR Design Control Document1. INTRODUCTION AND GENERAL

DESCRIPTION OF THE PLANT

Table 1.6-2 Material Referenced as Technical Reports (Sheet 3 of 6)

Report Number- Title DCD Section NumberO

MUAP-09002-P Summary of Seismic and Accident Load Conditions 3.7.2, 3.8.5. 3.9.2. 3.9.3MUAP-09002-NP for Primary Comonents and Piping. Revision 2.

December 2010.

M-AP 0004- P Summ,, of Strocc An.l6c• Rccultc for Corc 3.9.3, a.94MA4P 0.A0004 NPQ.... Suppz-t Stuo*tur. , R... io.i 1, Jnur; 20144

M-UAP-0•00-5 Su.'mmor; cf trect A~noly'zi Rzcul:'tc fcr Rentore R &-39.-3,-30.4

r4U.AP 00005 NP =c..;.:2 ., F.011.,20

MUAP-001006P Summary of Stress Analysis Results for Reamto-Q-.,4 . 32MUAP 09006 NP CVesats , RLopiizn2, 1Meonoh 2011

MUAP-09003-P Summary of Stress Analysis Results for MainSteam.r, .93..24MUAP 09007 NP RPiina 1eC isi o March 2011,MAAIAP 0-9000Q5 P SuwmmFarFy e f Strcocr Analy6alo Rcou-;11ltc- fcr, PRA O -FZ 8 F 3.. a .9.4-4

MA41IAP 090008 NIP Coo-lant Pump, Reyision 2, Mraeh 2011.

MUAP 090-0 P SuThmalr; of rauc Analysis Re, UAs fPR Cento Red 39.-9-.-94MUAP 090-0 NP FDeRa MckhRnisim, Rcaon 1, FJebrU; 2011. 1MUAP-09010-P Summary of Stress Analysis Results for Reactor a.6.3, 31.3, 3.6.423Z32_4MUAP-09010-NP Coolant Loop Piping. Revision 3. March 2011.

MUAP 0901- P Summary of Meto Analyoic ReoUt foator Ve-l 9 .4.-4MUA 0011 P Coolant Lazp Branh Piping, Reyieien 2, Deeembwr

MUAP 000121 P Summar; of Strccc Mnalys's Reeulte for 3-9.4-4-9.4-UA- 0601-12 NP Aocu-1Mula8tor, ReVOIGio 1, January 2011.

MUAP-0901 3-P Summary of Stress Analysis Results for Main Steam. 3.6.3, 3.0.3, 3.9.43B.3.2.4MUAP-09013-NP Pipino inside Containment, Revision 2. March 2011.

MUAP-09014-P Thermal-Hydraulic Analysis for US-APWR Spent 9.1.2MUAP-09014-NP Fuel Races. Revision 0. June 2009. 1MUAP-09016 US-APWR Reactor Vessel Pressure and 5.3.2. 16 (5.6.4)

Temperature Limits Report. Revision 2_Jufe-214 168Februar. 2013.

MUAP-09017-P Justification for 20 Years Inspection Interval for 5.4.1MUAP-09017-NP Reactor Coolant Pump Flyheel. Revision 0. July

MUAP-0901 8-P Calculation Methodology for Reactor Vessel 4.3.2MUAP-09018-NP Neutron Flux and Fluence. Revision 1. October

2009.

MUAP-09019-P HSI Desion. Revision 0 June 2009. 18.1.1. 18.1.2. 18.1.3.MUAP-09019-NP 18.1.4. 18.15.518.3.3.

18.4.3, 18.6.1. 18.6.3.18.7.2. 18.7.3

MUAP-09020-P Function Assignment Analysis for Safety Logic 7.3.1MUAP-09020-NP System. Revision 2. May 2011.

MUAP-09021-P Resoonse Time of Safety I&C System. Revision 2. 17.9.2. 16(B3.3.1).MUAP-09021-NP April 2011. ý116(B3.3.2)

MIC-03-01-00002MIC-03-01-00002 S1

MIC-03-03-00058

Tier 2 1.6-6 RSVOGOOR 3

1. INTRODUCTION AND GENERALDESCRIPTION OF THE PLANT

US-APWR Design Control Document

Table 1.6-2 Material Referenced as Technical Reports (Sheet 4 of 6)

Regort Numberf- Titl DCD Section Numberal

MUAP-09022-P US-APWR Instrument Setpoint Methodology. 7.2.1, 7.2.2. 7.3.2, 7.5.1MUAP-09022-NP Revision 2. May 2011.

MUAP-09023-P Onsite AC Power System Calculation, Revision 0. 8.31MUAP-09023-NP March 2010.

MUAP-09025-P CFD Analysis for Advanced Accumulator, Revision 6.3,2MUAP-09025-NP 2. August 2011.

M4IJAP 19991 Szicmis Design Bases ef the WS APAR Standord 3.741, 3.7.2, 3483, 346.Pjant, Rey:osier' 3, junz 21. Appenidi* 3H.(3H+.3

MUAP-10002-P Damping Ratio of SC Structure, Revision 0. March 3.7.1. 3.7.2, 3.7.3MUAP-10002-NP 2011.

MUAP-10003 US-APWR Physical Security Hardware ITAAC 14.3.4Abstracts. Revision 1. March 2011.

MUAP-1 0006 Soil-Structure Interaction Analyses and Results for 3.7.1. 3.7.2, 3.8.3 3.8.5

the US-APWR Standard Plant, Revision-+3, JaAowy Appendix 3H (3H.2.3,1.20- 4-November 2012. 3H.3'

Appendix 31 (31.1)

MUAP-1 0008 Staffing and Qualifications Implementation Plan, 18.5.1Revision 0. April 2010.

MUAP-1 0009 HSI Design Implementation Plan, Revision 0. April 18.1.1. 18.1.5. 18.6.32010.

MTWA,,Q Prc1,r De c'lepmct ,mplem o.tatin P41+7 18.1., 18.1.5, 18.6.3Reviion 0, April 2010.

MUA4Q44 Troining ProgmmF Pecelepment Plan., Rc':ision 0, 18.1.1, 15.1.5, 15A.6.Ap~l-2919-

MUAP-10012 Verification and Validation Implementation Plan, 18.1.1, 18.1.5, 18.6.3Revision 0, April 2010.

MUAP-10013 Design Implementation Plan, Revision 1, April 2010. 18.1.1. 18.1.5, 18.6.3.18.11.2

MUAP-10014 Human Performance Monitoring Implementation 18.1.1, 18.1.5, 18.6.31Plan, Revision 0, April 2010. 18.12.2

MUAP 10015 R Summray of EV.rIOnmcnts Fatiguc AnalycicMUAP 10015 NP RccutAc far thp US APAR ClQoz 1 C.mpen..t.,

Rc'yieon 1, Octebcr 2011

MAP 10016•P2 SummRay of ,AnircFnmcnt-l F-tiu-c A••lyscc14U A p 10016 NP Results fer the US APWR ReaotOr Ceooon L69W

Bronoh Piping, Revision 0, July 2091.

MUAP-10017-P US-APWR Methodology of Pipe Break Hazard 3.6.2MUAP-10017-NP Analysis, Revision 2. September 2011.

MUAP-1001•-P US-APWR Containment Performance for Pressure 38.1MUAP-10018-NP Loads. Revision 0. June 2010.

MUAP-10019-P Calculation Methodoloav for Radiological 11,2.3. 11.3.3MUAP-10019-NP Consequences in Normal Operation and Tank

Failure Analysis, Revision 1. March 2011.

MIC-03-01-00002MIC-03-01-00002 S1

MIC-03-01-00022

MIC-03-01-00022

MIC-03-01-00023

MIC-03-01-00023

MIC-03-03-00058

Tier 2 1.6-7 Tie 21.-7RevoeeR

1. INTRODUCTION AND GENERALDESCRIPTION OF THE PLANT


Table 1.6-2 Material Referenced as Technical Reoorts (Sheet 5 of 6)

Report Number lift DCD Section NMumber1MUAP-10020-P Safety-Related Air Conditioning, Heating. Cooling. 6.29. 6.5.7, 9.4.8MUAP-10020-NP and Ventilation Systems Calculations. Revision 42,

Ap~p- 2144-March 2013.

MUAP-10022 Evaluation on Jet Impingement Issues Associated 3.6.2with Postulated Pipe Rupture. Revision 1.September 2011.

MUAP-10023-P Initial Tvye Test Results of Class 1-E Gas Turbine 1.5.2. 3.10.2. 8.3.1MUAP-10023-NP Generator System. Revision 3, Septembr, 201.6

August 2013.

MUAP-10024 Structural Design Criteria for US-APWR Access 3.7.2Building, Revision 1. November 2011.

MWIAP 4,00, Auxir;"' B •uilding Med^ Properticc, 66' A.al'yse, 3..2An-d StRueturl lcgnty .&a I.ti-c4. forkh IhIR U APAR

Standard Plant, Rc'i io 1, Junc 2011.

MUAP-11002 Turbine Building Model Properties, SSI Analyses. 3.7.2and Structural Integrity Evaluation, Revision 02-' .... .February2013.I

MUAP-11003-P Summary of Stress Analysis Results for Pressurizer &.643B.3.2.4MUAP-11003-NP Surqe Line. Revision 1, March 2011.

MUAP-11005 Research Achievements of SC Structure and 383Strenath Evaluation of US-APWR SC StructureBased on 1/10th Scale Test Results Revision 01F•.-•e'- 2011 December 2012.

MWAP 4496 FEMzdclDz .. Iop mcntrd,,ifictie., Reiefo. , ;.;.2, 8H.4, -H.2.4,3 H..9,

MUAP-11007 Rcsults of E6.'alwtien UiR.g LSM; fcr R'B- 3.7.2. 3.8.5,-mple, R'.;.icien 0, J... 2011.Ground WaterEffects on SSI. Revision 2. November 2012.

141AP 41044 Effzatc of Str Ghurc Sail StRutc IANt'zrWaztAo (SSSU) 34;72onR Standard Soismie Dosign of US APWVR Plant,Revislen 0, Juno 2011.

MUAP-11012-P US-APWR RCCA Insertion Limit Load Test Report. 3.9.5MUAP-11012-NP Revision 0. March 2011.

MUAP-11013 Design CGntotna faor SC_ M ,dul, Roeyiciz 4, August 3.8.3

2Q4 4.Containment Internal Structure Design and

Validation Methodology. Revision 2. February 2013

MUAP-11014-P Over Temperature AT and Over Power AT Trip 7.2.1MUAP-11014-NP Function and Setooint Determination Process,

Revision 0, June 2011.

MUAP-11017-P Hydraulic Test of the Full Scale US-APWR Fuel 4.2MUAP-11017-NP Assembly, Revision 0. May 2011

MIC-03-01-00002MIC-03-01-00002 S1MIC-03-01-00026

DCD 03.10-17 S01

MIC-03-01-00022

MIC-03-03-00057

MIC-03-03-00058

MIC-03-01-00022

MIC-03-01-00022

MIC-03-01-00022

MIC-03-01-00022

MIC-03-03-00057

Tier 2 I .6-8 R~on4Tier 2 1.6-8 Rawmanan 2

3. DESIGN OF STRUCTURES, SYSTEMS, US-APWR Design Control DocumentCOMPONENTS, AND EQUIPMENT

3.6.3.4.9 Evaluation of Piping System Using BAC

This information will be pv,-idcd in the Tc., cr,^,al Rcpot (R•,f-r•nc• 3.6 21)developed as I DCD 14.03-9described in Appendix 3B. Sol

3.6.3.4.10 Bounding Analysis Results

Bounding analysis results will be p-.Vid.d in the Tc.hni..l Report (R•f.r.n.. 3.6 DCD 14.039244documented as described in Subsection 3.6.3.4.13. Sol

3.6.3.4.11 Differences in Inspection Criteria for Class I and 2 Systems

Class 1 and 2 systems are subjected to ISI requirements from ASME Code, Section Xl(Reference 3.6-11). For Class 1 piping, terminal ends and dissimilar metal welds arevolumetrically inspected, along with other locations, to total 25% of the welds. For Class 2piping, the requirement is to volumetrically inspect the terminal ends and other locationsto total 7.5% of the welds. These requirements were developed by ASME Code, SectionXl consistent with the different safety classes of these systems.

The LBB evaluations are based on the ability to detect a potential leaking crack; not theability to find cracks by ISI. The criteria or methods of the LBB evaluations are the samefor ASME Code, Section III, Class 1 and 2 (References 3.6-12 and 3.6-9).

3.6.3.4.12 Differences in Fabrication Requirements of ASME Code, Section IIIClass I and Class 2 Piping

The significant difference among ASME Code, Section III, Class 1 and 2 seamless pipesoccurs in the nondestructive examination requirements. The Class I seamless pipeexamination requirements include an ultrasonic testing examination, whereas Class 2does not. In addition, the Class 1 examination requirements for a circumferential buttwelded joint include radiographic testing and magnetic particle or liquid penetrantexamination where Class 2 does not.

For the fabrication of welds in the ASME Code, Section III, Class 1 and Class 2 pipes,there are no significant differences.

The differences in fabrication and nondestructive examination requirements do not affect

the LBB analyses assumptions, criteria, or methods.

3.6.3.4.13 Documentation of LBB Evaluation

Documentation of the LBB evaluation will be providod in the Tochnei! Report (Refrenc. DCD 14.03-93.)624developed as pt of the eloccout d'. umc.t.tio. for the r"p.... cystem Sol01ev'aluati•, ITA AC doccribed in Tier ! ,for the as-built piping systems to which LBB criteria MIC-03-03-

00058are applied and include the information described in subsection 3.6.3.5.

3.6.3.5 Technical Report

The following information will be pro.ided in the Tchnisl Rcpo-t (Refcrcn.. 3.66 DCD 14.03-924)developed as part of the coceout ..... me..... ÷ •Fr tho• . e.p. tive system L, So

MIC-03-03-00058

Tier 2 3.6-36

3. DESIGN OF STRUCTURES, SYSTEMS,COMPONENTS, AND EQUIPMENT


-va.uwtfin ITAAC dzcrFibcd in Tier 1 LBB evaluation of as-built piping systems to which MIC-03-03-

LBB criteria are applied. 100058

" Representative and bounding material properties.

" Design piping isometric drawings showing location of supports and theircharacteristics and location and weights of components such as valves.

" Evaluation of piping system using BAC and bounding analysis results.

3.6.4 Combined License Information

COL 3.6(1)

COL 3.6(2)

COL 3.6(3)

COL 3.6(4)

COL 3.6(5)

COL 3.6(6)

The COL Applicant is to identify the site-specific systems orcomponents that are safety-related or required for safe shutdown thatare located near high-energy or moderate-energy piping systems, andare susceptible to the consequences of these piping failures. The COLApplicant is to provide a list of site-specific high-energy and moderate-energy piping systems, which includes a description of the layout of allpiping systems where physical arrangement of the piping systemsprovides the required protection, the design basis of structures andcompartments used to protect nearby essential systems orcomponents, or the arrangements to assure the operability of safety-related features where neither separation nor protective enclosuresare practical. Additionally, the COL Applicant is to provide the failuremodes and effect analyses that verifies the consequences of failures insite-specific high-energy and moderate-energy piping does not affectthe ability to safely shut down the plant. The COL ADplicant is touodate the as-design piDe hazards analysis report to include theimnact of all site snecific hich and moderate ninina svstems

DCD_03.06.01-9

Deleted

Deleted

The COL Applicant is to implement the criteria for defining break andcrack locations and configurations for site-specific high-energy andmoderate-energy piping systems. The COL Applicant is to identify thepostulated rupture orientation of each postulated break location forsite-specific high-energy and moderate-energy piping systems. TheCOL Applicant is to implement the appropriate methods to assure thatas-built configuration of site-specific high-energy and moderate-energypiping systems is consistent with the design intent and provide as-builtdrawings showing component locations and support locations andtypes that confirms this consistency.

Deleted

Deleted

Tier 2 3.6-37 Reyme4eR


3.6-10 Nuclear Power Plant Components. ASME Section III, and Subarticle NE-1120for Containment Penetrations, American Society of Mechanical Engineers.

3.6-11 In-Service Examination of Pipe Welds. ASME Section Xl, IWA-2400,American Society of Mechanical Engineers.

3.6-12 ASME, Section III, Division 1, Class I Piping, NB-3653. American Society ofMechanical Engineers.

3.6-13 Instrument Lines Penetrating Primary Reactor Containment Safety Guide 11,Supplement to Safety Guide 11, Backfitting Considerations. Regulatory Guide1.11, U.S. Nuclear Regulatory Commission, Washington, DC, May 1971.

3.6-14 Design Bases for Protection of Light Water Nuclear Power Plants AgainstEffects of Postulated Pipe Rupture. ANSI/ANS-58.2-1988, American NationalStandards Institute/American Nuclear Society.

3.6-15 RELAP-5, Transient Hydraulic Analysis Program, MOD 3.2, Idaho NationalEngineering and Environmental Laboratory, Idaho Falls, Idaho, USA.

3.6-16 GOTHIC Containment Analysis Package User Manual, Version 7.2a(QA), NAI8907-02, Rev. 17, Numerical Applications Inc., Richland, WA, January 2006.

3.6-17 Stevenson, J.D. et. al., Structural Analysis and Design of Nuclear Power PlantFacilities. American Society of Civil Engineers.

3.6-18 Roemer, R.E. et al., Evaluation of Pipe Whip Impact on Concrete Barriers-ASimplified Approach. Proceeding of Second ASCE Conference on CivilEngineering and Nuclear Power, Volume IV (Impactive and Impulsive Loads),1980.

3.6-19 Enis, R.O. et. al., A Design Guide for Evaluation of Barriers for Impact fromWhipping Pipes. Proceeding of Second ASCE Conference on CivilEngineering and Nuclear Power, Volume IV (Impactive and Impulsive Loads),1980.

3.6-20 Report of the ASCE Committee on Impactive and Impulsive Loads. SecondASCE Conference on Civil Engineering and Nuclear Power, Volume V, 1980.

3.6-21 Guidance on Monitoring and Responding to Reactor Coolant SystemLeakage. Regulatory Guide 1.45 Rev. 1, U.S. Nuclear RegulatoryCommission, Washington, DC, May 2008.

3.6-22 Control of the Use of Sensitized Stainless Steel. Regulatory Guide 1.44,U.S. Nuclear Regulatory Commission, Washington, DC, May 1973.

3.6-23 Evaluation of Potential Pipe Breaks, NUREG-1061, Vol. 3, U.S. NuclearRegulatory Commission Piping Review Committee, November 1984.

3.6-24 . , ''y e S•e . . . . , ,' s R es-. t6 cI r the US .AI p . Rc atr CceA.Dt L e p I MIC-03-03-Piping, MUAP 00010, P and NP., Re%. 3, Mit.subichi Hcav ,ndkuctriec, Ltd., 00058MaFrh 2044.1

Tier 2 3.6-39 Tr236-9ReyeR


. . . . .. . ... .. e.., ,', AReult th USA12000 MIC-03-03-

Inside CRntan•n...t V ol..., MUAP poo 13, P and NP. , Rc. 2, Mitsubishi Heavy 00058

lnidust~es, Ltd., March 2011.

MUAP 11003, P and NPL, Rev. 1, Mitsubishi Heavy Indutricc, Ltd., March204. Deleted.

3.6-25 US-APWR Methodology of Pipe Break Hazard Analysis, MUAP-1 0017, Rev.43, Mitsubishi Heavy Industries Ltd., Drecmbcr, 2G01 Ma 2012. DCD_-03.06.

02-40

3.6-26 Kitade, K., Nakatogawa, T., Nishikawa, H., Kawanishi, K., and Tsuruto, C.,Experimental Study of Pipe Reaction Force and Jet Impingement Load at thePipe Break, Trans. 5th Int. Conf. on SMiRT, F6/2, 1979.

3.6-27 Kitade, K., Nakatogawa, T., Nishikawa, H., Kawanishi, K., and Tsuruto, C.,Experimental Studies on Transient Water-Steam Impinging Jet, Vol. 22 No. 5,pp. 403-409, Journal of Atomic Energy Society of Japan, 1980 (in Japanese).

3.6-28 Kitade, K., Nakatogawa, T., Nishikawa, H., Kawanishi, K., and Tsuruto, C.,Experimental Studies on Steam Free Jet and Impinging Jet, Vol. 22 No. 9, pp.634-640, Journal of Atomic Energy Society of Japan, 1980 (in Japanese).

3.6-29 Masuda, F., Nakatogawa, T., Kawanishi, K. and Isono, M., Experimental Studyon an Impingement High-Pressure Steam Jet, Nuclear Engineering andDesign 67-2, pgs 273-285, 1982.

3.6-30 Masuda, F., Nakatogawa, T., Kawanishi, K. and Isono, M., Experimental Studyon Jets Formed Under Discharges of High-Pressure Subcooled Water andSteam-Water Mixture, Trans. 7th Int. Conf. on SMiRT, Fl/6, 1983.

3.6-31 Isozaki, T. and Miyazono, S., Experimental Study of Jet Discharging TestResults under BWR and PWR Loss of Coolant Accident Conditions, NuclearEngineering and Design 96, 1986.

3.6-32 Evaluation on Jet Impingement Issues Associated with Postulated PipeRupture, MUAP-10022, Rev. 02, Mitsubishi Heavy Industries Ltd., JaR,-, I DCD-03.06.

20-l4May 2012. 102-40

Tier 2 3.6-40


reconciliation of the physical plant with the specified requirements of the ASMECode, Section III (Reference 3.9-1), the certified design requirements areassured.

10 CFR 52.47(b)(1) (Reference 3.9-31), requires that a design certificationapplication contain the proposed inspections, tests, analyses, and acceptancecriteria (ITAAC). The ITAAC are necessary and sufficient to provide reasonableassurance that, if the inspections, tests, and analyses are performed and theacceptance criteria met, a plant that incorporates the design certification is builtand will operate in accordance with the design certification, the provisions of theAtomic Energy Act, and the NRC's regulations; through requirements of "as-built"reconciliation of the physical plant with the specified requirements of the ASMECode, Section III (Reference 3.9-1), including ITAACs, the certified design isassured.

Proposed inspections, tests, and analyses which satisfy 10 CFR 52.80(a)(Reference 3.9-32) are discussed in Subsection 14.3, including those applicableto emergency planning as discussed in Subsection 13.3 that the licensee is toperform, and the acceptance criteria that are necessary and sufficient to providereasonable assurance that, if the inspections, tests, and analyses are performedand the acceptance criteria are met, the facility is constructed and will operate inconformity with the provisions of the Atomic Energy Act and the NRC'sregulations. The proposed ITAAC will assure that the facility is constructed,operates, and will continue to operate to the certified design conditions.

This subsection further describes the application of the ASME Code, Section III(Reference 3.9-1) to safety-related components and core support structures. The designand installation criteria applicable to over-pressure protection devices are presentedalong with the requirements for operability assurance related to maintaining structural andleak tight integrity, pressure retaining capability, and required functionality of pumps andvalves.

In order to assure that ASME components meet the service level stress requirements andfunctionality requirements, the ASME Code, Section III, NCA-2000 (Reference 3.9-1)requires that a design specification be prepared for ASME Code, Section III, Class 1, 2,and 3 components. The design specifications for ASME Code, Section III, Class 1, 2 and3 components, supports, and appurtenances are prepared under administrativeprocedures that meet or exceed the ASME Code, Section III rules. These specificationsconform to and are certified to the requirements of ASME Code, Section III depending onthe component classification. The Code also requires a design report for safety-relatedcomponents, to demonstrate that the component design meets the requirements of therelevant ASME design specification and the applicable ASME Code, Section III(Reference 3.9-1).

The seismic and accident load conditions for primary components and piping design aresummarized in Reference 3.9-58 and the c•t... BRnly'ic .. sult. for o...p9n9ent6 a•nd MIC-03-03-

piping-" r c.... ... .. in.^. - R fcz. 3.0 -6 . 100058

Tier 2 3.9-36 Re~o~4Tier 2 3.9-36 RA40WAR 2


The three coils are required to provide magnetic force to work latch assemblies. If thecoils fail, there is then no magnetic force and the RCCA drops into the core, shuttingdown the reactor. Therefore, failure of the coils is not a related safety issue. Coils arefabricated, in adherence to standard industrial quality assurance standards, and notIEEE, Class 1E standards.

Temperature of the operating coils is maintained below 392°F by forced air cooling.

3.9.4.3 Design Loads, Stress Limits, and Allowable Deformations

The pressure housing is required to comply with ASME Code, Section III(Reference 3.9-1) Subsection NB. The pressure housing is evaluated under the loadcombinations prescribed by the code. The loading combination and stress limits aredescribed in Subsection 3.9.3 and shown in Tables 3.9-9 and 3.9-10. This includesseismic loading. The allowable rod travel housing deflection during the seismic event is1.18 inch, which will still allow the RCCA to be inserted into the core. This is quantified byanalysis.

The clearances in the CRDM latch assembly, the latch arm, and the coil assembly arecontrolled to avoid a stuck rod condition. The thermal expansion of each part is evaluatedto determine and ensure the clearances. This design is the same as L-1 06A, whichreflects operationally-proven design and experience.

3.9.4.4 CRDS Operability Assurance Program

The functional performance of the CRDMs must be qualified both statically, as RCSpressure boundary components, and dynamically as functional mechanisms. To fulfillthese requirements, performance assurance programs are provided.

The structural integrity as a RCS pressure boundary, is confirmed by stress analysis inaccordance with ASME Code, Section III (Reference 3.9-1), Subsection NB. Also, ahydrostatic test, at ambient temperature, is performed in accordance with ASME Code,Section III (Reference 3.9-1), Subsection NB. St... .a,,"lyi•. , ,ult a. .p,,v..cd in i MIC-03-03-Roforonc- 3. 59. 100058

The capability of the CRDM functions, including withdrawal, insertion, and trip delay areconfirmed by both lead unit tests and production unit tests to demonstrate that the designspecification requirements are met prior to shipment. The lead unit testing is described asfollows:

Lead unit test is performed on the first production unit of the applicable type of CRDM.

- Cold stepping test witih full design drive line load and maximum stepping speed(72 steps per minute).

Criteria: no mis-stepping

- Trip delay time test at cold condition.

Tier 2 3.9-66 Re~e~4Tier 2 3.9-66 I : 7 72" 1 1

3. DESIGN OF STRUCTURES, SYSTEMS,COMPONENTS, AND EQUIPMENT


3.9-50A L /'•

;Anloncn c eto 6.o, crcic.. A6. .aru, ;Acr~ccn "..crcic inemutC..gReauirements for Nuclear Safety-Related Concrete Structures (ACI 349-06')MIC-03-03-00066

and Commentary. American Concrete Institute, 2006.

3.9-51 Anchoring Components and Structural Suoports in Concrete. RegulatoryGuide 1.199, Rev. 0, U.S. Nuclear Regulatory Commission, Washington, DC,November 2003.

3.9-52 Quality Group Classifications and Standards for Water-, Steam-, andRadioactive-Waste-Containing Components of Nuclear Power Plants.Regulatory Guide 1.26 Rev. 4, U.S. Nuclear Regulatory Commission,Washington, DC, March 2007.

3.9-53 Guidance on Developing Acceptable Inservice Testing Program, GL 89-04,

U.S. Nuclear Regulatory Commission, Washington, DC, April, 1989.

3.9-54 Periodic Verification of Desiqn Basis Capability of Safety-Related Motor-Operated Valves, GL 96-05, U.S. Nuclear Regulatory Commission,Washington, DC, September, 1996.

3.9-55 Safety-Related Motor-Operated Valves, Testing and Surveillance, GL 89-10,U.S. Nuclear Regulatory Commission, Washington, DC, June 1989.

3.9-56 Primary Reactor Containment Leakage Testing for Water-Cooled PowerReactors, Domestic Licensing of Production and Utilization Facilities, Energy.Title 10, Code of Federal Regulations, Part 50, Appendix J, U.S. NuclearRegulatory Commission, Washington, DC.

3.9-57 Summary of Design Transient, MUAP-09001, Rev. 0 (Proprietary) and MUAP-09001, Rev. 0 (Non-Proprietary), February 2009.

3.9-58 Summary of Seismic and Accident Load Conditions for Primary Componentsand Piping, MUAP-09002, Rev. 1 (Proprietary) and MUAP-09002, Rev. 1(Non-Proprietary), March 2009.

3.9-59P A

00001 Ro'.' 0 (Proprietairy) an~d MUAP 00001 Rev. 0 (Non Proprita0) M.a~rch

Re'v-. 0 (Pro.prietarY) and MUAP 00005 Re'I . 0 (Non Pr t March 20,•-

0 (P•r2.pieta.y) and MUAP 00006 Re. 0 (Non Proprieta.Y), March 2000,Summa..' of StFrse A.aly-se Results fo•rPrFe6•-- rF-r,- MUAP 0r0007 Rev. 0(Proprfietary) and MUAP 090007 Rce. 0 (Non Proprietar~y), March 2000,

Summar; ef Slress .A.a•;;s• Results fe; ReatrWF Ge~Rt Pump, MWAR.0900,Rev. 0 (Proprietary) an~d MUAP 00008 Re%. 0 (Non Proprietary), March 2000r-

MUJAP 090000 Rev-' 0 (Proepiietarcy) anid MUAP 090000 Res. 0 (Non Proprietor;),Pig 200, 001m0 c..... ef StPrpri ,.taly' Res) tnd fUP R0a001 Ccc.a t (NonB=,.•=a-, IWAI 0IA 1 Rev.•~t/' I^, (PePi tay aB;• d, MW ki AP 09010/•£ Re, I "'

MIC-03-03-00058

Tier 2 3.9-109


Coolant Leoo Bra6ch Piing (1), MUAP 09011 Rcv. 0 (Pr..rietar;) andMUAP 000144 Re. 0 (Non Poprio.t..;), March14 2000, SUmmR:. . o•f toAnal..i. Results for Accumulato , MUAP 00012 Rev. 0 (Prcprictar) andMUAP 00012 Rey. 0 (Non Proprietar;), March 2000, Summa; f Strcc

Rey. 0 (Proepretary) and MUAP 0901 3 Rev-. 0 (Non PrOPrietar;), M4arch2-0.9Deleted.

3.9-60 Guidelines for Inservice Testing at Nuclear Power Plants, NUREG-1482, U.S.Nuclear Regulatory Commission, Washington, DC, April 1995.

3.9-61 MOV Periodic Verification (PV) Study, MPR 2524-aA, Joint Owners Group(JOG), November 2006.

3.9-62 Joint Owners Grouo Air Operated Valve Proaram Document, Revision 1,December 13, 2000.

3.9-63 Comments on Joint Owners' GrouD Air Operated Valve Program Document,USNRC Letter from Eugene V. Imbro to Mr. David J. Modeen, Nuclear EnergyInstitute, October 8, 1999.

3.9-64 Resolution of Generic Safety Issue 158: Performance of Safety-RelatedPower-Operated Valves Under Design Basis Conditions, Regulatory IssueSummary RIS 2000-03, U.S. Nuclear Regulatory Commission, Washington,DC, March 15, 2000.

3.9-65 PIC~EP: Pipe Crack Eivalwation Program. NP 3596 ISR, Rey.4, Electric PowerRo..... I" ';"itute, 10987.GOTHIC. Containment Analysis Package UserManual. Version 7.2a (QA&. NAI 8907-02, Rev. 17, Numerical ApplicationsInc.. Richmond. WA. January 2006.

3.9-66 Qualification of Active Mechanical Equipment Used in Nuclear Power Plants.American Society of Mechanical Engineers (ASME) QME-1-2007.

3.9-67 Alternative Rules for Preservice and Inservice Testing of Certain Motor-Operated Valve Assemblies in Light-Water Reactor Power Plants, AmericanSociety of Mechanical Engineers (ASME) Code Case OMN-1, Rev. 0. 1999.

3.9-68 Final Safety Evaluation on Joint Owner's Group Program on Motor-OperatedValve Periodic Verification. U.S. Nuclear Regulatory CommissionWashington,DC. September 25, 2006, with its supplement dated September 18,2008.

3.9-69 US-APWR RCCA Insertion Limit Load Test Report, MUAP-11012 Rev. 0(Proprietary) and MUAP-11012 Rev. 0 (Non-Proprietary), Mitsubishi HeavyIndustries, Ltd., March 2011.

MIC-03-03-00058

eDCD03.09.06-58

DCD_03.09.01-6

DCD_03.09.06-49

DCD_03.09.06-58

DCD_03.09.05-34

Tier 2 3.9-110 Tie 2 .9-10ReyieiR

3. DESIGN OF STRUCTURES, SYSTEMS, US-APWR Design Control DocumentCOMPONENTS, AND EQUIPMENT Appendix 3B

3B.3.2.4 Assessment of LBB Concept

To compare the piping stresses to the BAC curves, the normal operating stress and themaximum stresses must be determined for all analyzed locations (including all welds) inthe piping system. Examples of LBB evaluations are provided in Technical RePorts MIC-03-03-MUAP-09010 (Reference 3B-15), MUAP-09013 (Reference 3B-16) and MUAP-11003 00058

(Reference 3B-17).

1. Assessment points (unonlumaxl) are plotted on the BAC diagram for all locations in

a piping system.

2. If the assessing point falls below the BAC, the detection of leakage before ruptureis possible and then the LBB concept can be applied.

When specifically conducting LBB evaluation on the US-APWR utilizing the BACdiagrams, the following attention should be noted.

a. The wall thickness used in determining the BAC is different than that used inthe piping assessment. The BAC is based on the nominal wall thickness andnot the minimum wall thickness. The nominal wall thickness providesconservative results, since for the greater wall thickness, there will be lessleakage. However, in the LBB assessment, the evaluation should beconducted utilizing the minimum wall thickness, since the thinner the wall willresult in higher calculated stresses under the maximum applied load.

b. In the case of SMAWwelding for stainless steel, the maximum load must bemultiplied by the Z-factor for the specific pipe size because the fracturetoughness value is lower compared to the base metal or TIG welding.

3B.4 BAC Setting for LBB Evaluation

Table 3B-2 lists the piping system selected for development of BACs. The detailed BACsare shown in Figures 3B-6 through Figure 3B-18.

3B.5 References

3B-1 General Design Criteria for Nuclear Power Plants, Domestic Licensing ofProduction and Utilization Facilities, Energy. Title 10, Code of FederalRegulation, Part 50, Appendix A, U.S. Nuclear Regulatory Commission,Washington, D.C.

3B-2 Leak-Before-Break Evaluation Procedures, Design of Structures,Components, Equipment, and Systems, Standard Review Plan for the Reviewof Safety Analysis Reports for Nuclear Power Plants. NUREG-0800, SRP3.6.3, Rev. 1, U.S. Nuclear Regulatory Commission, Washington, DC, March2007.

Tier 2 313-18 Tier2 3-8Revison 3


3B-3 Evaluation of Potential for Pipe Breaks, Report of U.S. NRC Piping ReviewCommittee. NUREG-1 061, Vol. 3, U.S. Nuclear Regulatory Commission,Washington, DC, 1984.

3B-4 PICEP: Pipe Crack Evaluation Program. NP-3596-SR, Rev. 1, Electric PowerResearch Institute, 1987.

3B-5 Advances in Elastic-Plastic Fracture Mechanics. NP-3607, Final Report,Electric Power Research Institute, 1984.

3B-6 Henry, R.E., The Two-Phase Critical Discharge of Initially Saturated orSubcooled Liquid. Nuclear Science and Engineering, Vol. 41, pp.336-342,1970.

38-7 Calculation of Leak Rates Through Cracks in Pipes and Tubes. NP-3395,Electric Power Research Institute, 1983.

3B-8 Part D Properties, American Society of Mechanical Engineers, ASME Boilerand Pressure Vessel Code, Section II, 2001 (Addenda 2003).

3B-9 Elastic-Plastic Fracture Mechanics Analysis of Through-Wall and SurfaceFlaws in Cylinders. NP-4496, Final Report, Electric Power Research Institute,1988.

3B-10 Cofie, N.G., Miessi, G.A., and Deardorff, A.F., Stress-Strain Parameters inElastic-Plastic Fracture Mechanics, Transactions of the 1 0 th InternationalConference on Structural Mechanics in Reactor Technology, Volume L -Inelastic Behavior of Metals and Constitutive Laws of Materials, pp 91-96,1989.

38-11 State of the Art Report on Piping Fracture Mecolmhanics, NUREG/CR-6540,U.S. Nuclear Regulatory Commission, November 1997.

38-12 Assessment of Short Through-Wall Circumferential Cracks in Pipe, NUREG/CR-6235, U.S. Nuclear Regulatory Commission, April 1995.

3B-13 Probabilistic Pipe Fracture Evaluations for Leak-Rate Detection Applications.NUREG/CR-6004, U.S. Nuclear Regulatory Commission, April 1995.

38-14 Pipe Fracture Encyclopedia, Test Data - Volume 3, U.S. Nuclear RegulatoryCommission, December 1997.

38-15 Summary of Stress Analysis Results for the US-APWR Reactor Coolant Loop MIC-03-03-Pipinaq. MUAP-09010. P and NP, Rev. 3, Mitsubishi Heavy Industries, Ltd., 00058

March 2011.

38-16 Summary of Stress Analysis Results for the US-APWR Main Steam PipinqInside Containment Vessel, MUAP-09013, P and NP, Rev. 2, Mitsubishi HeavyIndustries, Ltd., March 2011.

Tier 2 3B-19 Re~o~4

Tier 2 313-19 Ravoigoan 2


3B-17 Summary of Stress Analysis Results for the US-APWR Pressurizer Surae MIC-03-03-Line, MUAP-11003, P and NP. Rev. 1. Mitsubishi Heavy Industries, Ltd.. March 000582011.

Tier 2 3B-20 Ro~o~4Tier 2 313-20 Rawos*an 2

5. REACTOR COOLANT AND US-APWR Design Control DocumentCONNECTING SYSTEMS

5.3-19 Materials, ASME Boiler and Pressure Vessel Code, Section II, AmericanSociety of Mechanical Engineers, 2001 Edition with 2003 Addenda.

5.3-20 Rules for Construction of Nuclear Power Plant Comoonents, ASME Boiler andPressure Vessel Code, Section III, American Society of MechanicalEngineers, 2001 Edition with 2003 Addenda.

5.3-21 Nondestructive Examination, ASME Boiler and Pressure Vessel Code,Section V, American Society of Mechanical Engineers, 2001 Edition with 2003Addenda.

5.3-22 Welding and Blazing Qualifications, ASME Boiler and Pressure Vessel CodeSection IX, American Society of Mechanical Engineers, Latest Edition andAddenda.

5.3-23 Rules for In-service Insoection of Nuclear Power Plant Comoonents, ASMEBoiler and Pressure Vessel Code Section Xl, American Society of MechanicalEngineers, 2001 Edition with 2003 Addenda.

5.3-24 Standard Practice for Conducting Surveillance Tests for Light-Water CooledNuclear Power Reactor Vessels, ASTM E-1 85-82.

5.3-25 Standard Test Method for Linear-Elastic Plane-Strain Fracture Touahness K-of Metallic Materials, ASTM E-399.

5.3-26 Standard Test Method for Measurement of Fracture Toughness, ASTM E-1820.

5.3-27 Timoshenko, S. P. and Goodier, J. N., Theorv of Elasticity, Third Edition,McGraw-Hill Book Co., New York, 1970.

5.3-28 US-APWR Reactor Vessel Pressure and Temoerature Limits Report, MUAP-09016 Rev.43, j..a..y. 201 Februarv 2013. MIC-03-03-

00058

Tier 2 5.3-29

ATTACHMENT - 1

Change List

Change ID No. Section Page I Reason for change Change SummaryMIC-03-03-00058 Tier 2

Table1.6-2(Sheet)

1.6-6, to1.6-8

Tier 2

3.6.3.4.13

3.6.3.5

3.6.3.5

3.6.5

3.9.3

3.9.4.4

3.9.10

Revised DesignCompletion Plan forUS-APWR Piping Systemsand Components,UAP-HF-12208, datedDecember 7, 2012

Revised DesignCompletion Plan forUS-APWR Piping Systemsand Components,UAP-HF-12208, datedDecember 7, 2012

3.6-36

3.6-36

3.6-36

3.6-39

3.9-36

3.9-66

3.9-109 to3.9-110

Update thedescriptionwithdrawing theTechnical Reportsassociated withstress analysis.

Tier 23B.3.2.4 3B-18 Revised Design Update the

Completion Plan for description adding3B.5 3B-19 US-APWR Piping Systems the Technical

and Components, Reports associatedUAP-HF-12208, dated with stress analysis.December 7, 2012

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

Tier 25.3.5 5.3-29 Revised Design

Completion Plan forUS-APWR Piping Systemsand Components,UAP-HF-12208, datedDecember 7, 2012

Update thedescription revisingthe TechnicalReports associatedwith stress analysis.

dcd markups associated with revised design completion …mua 0011 p coolant lazp branh piping,...

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