MHI's Response to US-APWR DCD RAI No. 540-4176 Revision 0. · Subject: MHI's Response to US-APWR DCD RAI No. 540-4176 REVISION 0 Reference: 1) "Request forAdditional Information No

AtMITSUBISHI HEAVY INDUSTRIES, LTD.

16-5, KONAN 2-CHOME, MINATO-KU

TOKYO, JAPAN

June 4, 2010

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

Attention: Mr. Jeffrey A. Ciocco

Docket No. 52-021MHI Ref: UAP-HF-10159

Subject: MHI's Response to US-APWR DCD RAI No. 540-4176 REVISION 0

Reference: 1) "Request forAdditional Information No. 540-4176 Revision 0, SRP Section:05.02.03 - Reactor Coolant Pressure Boundary Materials, ApplicationSection: DCD Tier 2, Section 5.2.3" dated March 2, 2010.

With this letter, Mitsubishi Heavy Industries, Ltd. ("MHI") transmits to the U.S. NuclearRegulatory Commission ("NRC") a document entitled "Response to Request forAdditionalInformation No. 540-4176 Revision 0."

Enclosed is the response to the RAI contained within Reference 1.

In the responses to Question No. 05.02.03-19, No.05.02.03-20, and No.05.02.03-25, thesupplemental answers are provided in Enclosure 3, 5, and 7 accordingly. As indicated inEnclosure 3, 5, and 7, this submittal contains information that MHI considers proprietary, andtherefore should be withheld from public disclosure pursuant to 10 C.F.R. § 2.390 (a)(4) astrade secrets and commercial or financial information which is privileged or confidential. Theproprietary information is bracketed by the designation "[ ]".

This letter includes a copy of the proprietary supplemental answers (Enclosure 3, 5, and 7)and the Affidavit of Yoshiki Ogata (Enclosure 1) which identifies the reasons MHI respectfullyrequests that all materials designated as "Proprietary" in Enclosure 3, 5, and 7 be withheldfrom public disclosure pursuant to 10 C.F.R. § 2.390 (a)(4).

Please contact Dr. C. Keith Paulson, Senior Technical Manager, Mitsubishi Nuclear EnergySystems, Inc. if the NRC has questions concerning any aspect of the submittals. His contactinformation is below.

Sincerely,

Yoshiki OgataGeneral Manager-APWR Promoting DepartmentMitsubishi Heavy Industries, LTD.

Enclosures:

1. Affidavit of Yoshiki Ogata

2. Response to Request for Additional Information No. 540-4176 Revision 0

3. Supplemental Answers to QUESTION 05.02.03-19 of RAI No. 540-4176 Revision 0(Proprietary Version)

4. Supplemental Answers to QUESTION 05.02.03-19 of RAI No. 540-4176 Revision 0(Non Proprietary Version)





CC: J. A. CioccoC. K. Paulson

Contact InformationC. Keith Paulson, Senior Technical ManagerMitsubishi Nuclear Energy Systems, Inc.300 Oxford Drive, Suite 301Monroeville, PA 15146E-mail: [email protected]: (412) 373-6466

ENCLOSURE 1Docket No.52-021

MHI Ref: UAP-HF-10159

MITSUBISHI HEAVY INDUSTRIES, LTD.

AFFIDAVIT

I, Yoshiki Ogata, state as follows:

1. I am General Manager, APWR Promoting Department, of Mitsubishi Heavy Industries,LTD ("MHI"), and have been delegated the function of reviewing MHI's US-APWRdocumentation to determine whether it contains information that should be withheld frompublic disclosure pursuant to 10 C.F.R. § 2.390 (a)(4) as trade secrets and commercial orfinancial information which is privileged or confidential.

2. In accordance with my responsibilities, I have reviewed the enclosed document entitled"Supplemental Answers to RAI No. 540-4176 Revision 0", and have determined thatportions of the document contain proprietary information that should be withheld frompublic disclosure. Those pages containing proprietary information are identified with thelabel "Proprietary" on the top of the page and the proprietary information has beenbracketed with an open and closed bracket as shown here "[ ]". The first page of thedocument indicates that all information identified as "Proprietary" should be withheld frompublic disclosure pursuant to 10 C.FR. § 2.390 (a)(4).

3. The information identified as proprietary in the enclosed document has in the past been,and will continue to be, held in confidence by MHI and its disclosure outside the companyis limited to regulatory bodies, customers and potential customers, and their agents,suppliers, and licensees, and others with a legitimate need for the information, and isalways subject to suitable measures to protect it from unauthorized use or disclosure.

4. The basis for holding the referenced information confidential is that it describes thewelding material procurement specifications developed by MHI and material vender forthe Reactor Coolant Pressure Boundary of the US-APWR, and test results conducted byMHI and are not issued publicly.

5. The referenced information is being furnished to the Nuclear Regulatory Commission("NRC") in confidence and solely for the purpose of information to the NRC staff.

6. The referenced information is not available in public sources and could not be gatheredreadily from other publicly available information. Other than through the provisions inparagraph 3 above, MHI knows of no way the information could be lawfully acquired byorganizations or individuals outside of MHI.

7. Public disclosure of the referenced information would assist competitors of MHI in theirdesign of new nuclear power plants without incurring the costs or risks associated with thedesign of the subject systems. Therefore, disclosure of the information contained in thereferenced document would have the following negative impacts on the competitiveposition of MHI in the U.S. nuclear plant market::

A. Loss of competitive advantage due to the costs associated with development of thewelding material procurement specifications for the Reactor Coolant PressureBoundary of the US-APWR. Providing public access to such information permitscompetitors to duplicate or mimic the methodology without incurring the associatedcosts.

B. Loss of competitive advantage of the US-APWR created by benefits of enhancedplant safety, and reduced operation and maintenance costs associated with theReactor Coolant Pressure Boundary.

I declare under penalty of perjury that the foregoing affidavit and the matters stated thereinare true and correct to the best of my knowledge, information and belief.

Executed on this 4th day of June, 2010.

Yoshiki Ogata,General Manager-APWR Promoting DepartmentMitsubishi Heavy Industries, LTD.


Enclosure 2

UAP-HF-10159Docket Number 52-021

Response to Request for Additional InformationNo. 540-4176 Revision 0

June 2010

RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION

6/4/2010

US-APWR Design Certification

Mitsubishi Heavy Industries

Docket No. 52-021

RAI NO.: NO. 540-4176 REVISION 0SRP SECTION: 05.02.03 - REACTOR COOLANT PRESSURE BOUNDARY

MATERIALS

APPLICATION SECTION: 05.02.03

DATE OF RAI ISSUE: 3/02/2010

QUESTION NO. : 05.02.03-19

Table 5.2.3-1 lists weld filler materials using the "G" classification. In RAI 05.02.03-4 (b), the staffrequested that the applicant modify FSAR Table 5.2.3-1 to list its weld filler metal requirementsfor weld filler metals using the "G" classification. In the applicant's response, it provided, asproprietary information in Enclosure 3, some of the chemical compositions of "G" class fillermaterials used in the RCPB. In order for the staff to continue its review, the applicant needs toaddress the following:

(a) The applicant did not provide chemical compositions of all "G" classification filler materialslisted in Table 5.2.3-1, in Enclosure 3 of its RAI response. The staff requests that the applicantprovide a list of the chemical compositions of all "G" classification filler materials listed in Table5.2.3-1. The applicant's response should include chemistry requirements for solid electrodes aswell as chemical composition requirements for as deposited weld material.

(b) The title for Table 3 in Enclosure 3 of the RAI response lists bare wire filler material but theAWS Classification listed above the table title lists shielded metal arc welding filler material. Thestaff requests that the applicant review the table title and revise as necessary.

(c) Shielded metal arc welding (SMAVV) and submerged arc welding (SAW) classifications listedin Table 5.2.3-1 do not include supplemental requirements for limits on diffusible hydrogen. Thestaff requests that the applicant specify a limit for diffusible hydrogen [e.g., a maximum of 4 ml ofhydrogen per 1 OOg of deposited weld metal (H4 designator) for all low-alloy-steel, weld-fillermaterials] and modify Table 5.2.3-1, accordingly.

(d) DCD Table 5.2.3-1 lists E9P4-EGN-GN and P10P2-EG-G. These filler material classificationsare not consistent with ASME Code, Section II, SFA-5.23. The staff requests that the applicantreview these classifications and revise DCD Table 5.2.3-1 accordingly.

5.2.3-1

ANSWER:

(a) Chemical compositions of all Welding Material with "G" classification filler material listed inTable.5.2.3-1 are proprietary and shown in the proprietary answer, Enclosure 3.Chemistry requirement for solid electrode are also included.

(b) The title ER90S-G is a typo and E10016-G is correct. Supplemental Answer to Question05.02.03-4 Table-3 is revised to the Table-2 in Enclosure 3 of this RAI response and modified toE10016-G.

(c) MHI will not include the supplemental requirements for limits on diffusible hydrogen because. welding material supplier can not assure the value of diffusible hydrogen.

(d) E9P4-EGN-GN and P10P2-EG-G are typo. The DCD Table 5.2.3-1 will be modified fromE9P4-EGN-GN to F9P4-EG-G, and P1OP2-EG-G to F10P2-EG-G.

Impact on DCD

Filler material classifications E9P4-EGN-GN and P1OP2-EG-G of the DCD Table 5.2.3-1 will bechanged as shown in the attached mark-up 19.

Impact on COLA

There is no impact on the COLA.

Impact on PRA

There is no impact on the PRA.

5.2.3-2

Attachment 19

Table 5.2.3-1 Reactor Coolant Pressure Boundary Material Specifications(Sheet I of 4)

Component Material Class, Grade, or TypeReactor Vessel Parts

Pressure boundary welds SFA-5.5 E9016-G(Low alloy steel) SFA-5.23 F9P4-EG-G,

________________ 9124 EGN ONSFA-5.28 ER80S-G

Table 5.2.3-1 Reactor Coolant Pressure Boundary Material Specifications(Sheet 2 of 4)

Component I Material J Class, Grade, or TypeSteam Generator Parts

Pressure boundary welds SFA-5.5 E9016-G(Low alloy steel) E10016-G

SFA-5.23 F9P4-EG-GE9P4 EGN GNP-10P2- EG G F1OP2-EG-G

SFA-5.28 ER80S-GER90S-G

Table 5.2.3-1 Reactor Coolant Pressure Boundary Material Specifications(Sheet 3 of 4)

Component Material Class, Grade, or TypePressurizer Parts

Pressure boundary welds SFA-5.5 E9016-G(Low alloy steel) E10016-G

SFA-5.23 F9P4-EG-GE912EO4 OP!OP2- EG G F1OP2-EG-G_

SFA-5.28 ER80S-GER90S-G

5.2.3-3


614/2010



Docket No. 52-021

RAI NO.: NO. 540-4176 REVISION 0

SRP SECTION: 05.02.03 - REACTOR COOLANT PRESSURE BOUNDARYMATERIALS



QUESTION NO. : 05.02.03-20

In response to RAI 05.02.03-5(b), the applicant modified DCD Section 5.2.3.3.2 to state thattaking into consideration the weldability and quality of the product, preheating above 122°F isapplied to the first and second passes of circumferential joint welding. The applicant further statedthat preheating of more than 250'F is applied to subsequent passes in accordance with AppendixD of the ASME Code, Section III. Given the significant difference between the Appendix-D-recommended, minimum-preheat temperature and the minimum-preheat temperature that theapplicant will use on the first and second weld passes, the staff requests that the applicantprovide justification and the technical basis for using a reduced minimum-preheat temperature of122°F.

ANSWER

Minimum preheat temperature of 122°F is applied to the first and second passes ofcircumferential joint welding.The soundness of weld joint is verified by procedure qualification test according to Sec. IX.Welding Procedure Specification (WPS) and Procedure Qualification Record (PQR), whichshow the reliability of the 122°F preheat temperature, are proprietary and attached in theproprietary answer, Enclosure 5, because these documents are MHI proprietary documents.Note that the Appendix-D of the ASME Sec.l1l is not mandatory.

Impact on DCD

There is no impact on the DCD.

Impact on COLA


5.2.3-4

Impact on PRA


5.2.3-5


6/4/2010



Docket No. 52-021





QUESTION NO.: 05.02.03-21

The applicant's use of post-weld baking, as described in DCD Section 5.2.3.3.2, is an alternativeto the guidance provided in RG 1.50 that recommends the preheat temperature applied to low-alloy steels be maintained until post-weld heat treatment (PWHT) begins. The applicant'sdiscussion related to post-weld baking was added to DCD Revision 2 in response to RAI05.02.03-5(c). The staff considers the applicant's alternative to RG 1.50 acceptable because itprovides reasonable assurance that delayed hydrogen cracking will not occur in the time that aweld is completed through completion of PWHT. However, Table 1.9-1 indicates that theapplicant conforms to RG 1.50 with no exceptions identified. This appears to be incorrectbecause the applicant provides the above-described alternative to RG 1.50. The staff requeststhat the applicant modify Table 1.9-1 to identify its alternative to the guidance provided in RG 1.50.

ANSWER:

Table 1.9-1 of the DCD will be changed to indicate that the alternative to the guidance provided inRG 1.50 will be used.

Impact on DCD

The DCD Table 1.9-1 RG 1.50 line will be changed as shown in the attached mark-up 21.

Impact on COLA


Impact on PRA


5.2.3-6

Attachment 21

Table 1.9.1-1 US-APWR Conformance with Division 1 Regulatory Guides (sheet 4 of 15)

Reg Title Status CorrespondingGuide Chapter/Section

Number /Subsection

1.50 Control of Preheat Temperature for Welding of Low-Alloy Conformance with nROexceptions4dentified. 5.3.1.2, 5.3.1.4,Steel (Rev. 0, May 1973) The alternative to the RG 1.50, for example 5.2.3.3.2, 6.1.1

post weld bakinq based on the ASME SectionIII NB-4622.9 item 7, may be applied.

5.2.3-7


6/4/2010



Docket No. 52-021





QUESTION NO. : 05.02.03-22

In the applicant's response to RAI 05.02.03-10, the applicant stated that it may use ASTM A-800to calculate the ferrite content of cast-austenitic-stainless-steel (CASS) materials used in the US-APWR design. The applicant contends that the methods used to calculate ferrite content in A-800is essentially the same as the Hull's equivalent factor method. In addition, the applicant did notspecify the maximum ferrite content limit for various CASS materials. The use of ASTM A-800 isinconsistent with the staffs position that ferrite content is calculated using Hull's equivalent factorsas indicated in NUREG/CR-4513, Revision 1, "Estimation of Fracture Toughness of CastStainless Steels During Thermal Aging in LWR Systems," issued May 1994. For ferrite contentabove 12 percent, ASTM A-800 may produce non-conservative ferrite levels lower than thosecalculated using Hull's equivalent factors. The staff, therefore, requests that the applicant modifythe DCD to require the use of Hull's equivalent factors to calculate the ferrite content of CASSRCPB components and CASS reactor vessel internals.

Regarding ferrite limits for CASS components, DCD Table 5.2.3-1 identifies CF3A, CF3M, CF8and CF8M materials for use in the RCPB. Table 4.5-2 identifies the use of the use of CF8 forguide funnel of the CRDM thermal sleeve. To be consistent with staff guidance, CF3A and CF8materials, with service conditions above 4820F, should have a ferrite content of -20% to beconsidered not susceptible to thermal aging embrittlement. Materials with higher levels ofmolybdenum such as CF3M and CF8M should have a ferrite content of -<14% to be considerednot susceptible to thermal aging embrittlement.

The staff requests that the applicant modify the DCD to limit the ferrite content of RCPB andreactor internal CASS components as discussed above.

In addition, address ferrite content for CASS components and the method used to calculate ferrite,as discussed above, for any ESF components that operate above 4820 F, such as ESFcomponents that interface with the RCPB. Also, modify DCD Section 6.1.1 as necessary.

5.2.3-8

ANSWER:

MHI will require to the material venders of RCPB of Class 1 components and Reactor Internals tobe less than 14% ferrite content for 316 CASS (higher level molybdenum materials) and be lessthan 20% ferrite content for 304 CASS (lower level molybdenum materials) by calculating theferrite content using Hull's equivalent factors as indicated in NUREG/CR-4513, Revision 1. DCDwill be modified to add the controlled ferrite content of the CASSs for the materials to be used 482deg F or higher.

Impact on DCD

Following DCD subsections will be revised as follows.- Fourth paragraph of the DCD Subsection 5.2.3.4 will be revised as shown in the attached

mark-up 22-1.- DCD Subsection 5.2.7 item 5.2-40 will be deleted as shown in the attached mark-up 22-2.

Impact on COLA


Impact on PRA


5.2.3-9

Attachment 22-1

5.2.3.4 Fabrication and Processing of Austenitic Stainless Steels

Unstabilized stainless steel can be subject to SCC when the steels are sensitized andcertain contaminants are adhered to the surface of the materials, and the materials areexposed to a stressed condition. In PWR plants, dissolved oxygen and chloride contentsof the reactor coolant are well controlled to prevent SCC of austenitic stainless steel.Field experience has shown that de-oxygenated, hydrogenated PWR primary water doesnot cause SCC in sensitized materials, unlike BWRs with oxygenated water (Ref. 5.2-31).

Although the water chemistry condition is well controlled, good metallurgical and

environmental practices will be followed to further assurance of avoiding SCC, such as:

- Process control of cleaning and protection against contamination

- Use of materials in the final heat treated condition

- Control of welding processes and procedures to avoid heat-affected zone (HAZ)sensitization

Subsections 5.2.3.4.1 and 5.2.3.4.2 address Regulatory Guide 1.44, "Control of the Useof Sensitized Stainless Steel," and present the methods and controls to avoidsensitization and to prevent intergranular attack (IGA) of austenitic stainless steelcomponents.

Cast austenitic stainless steel components used in light-water reactors can besusceptible to thermal aging embrittlement due to the formation of a Cr-rich phase fromthe decomposition of the ferrite phase by exposing the material to elevated temperatures.Cast austenitic stainless steels that are used in US-APWR RCPB components arecategorized based on molybdenum content, casting method, and 5-ferrite level, andexamined based on ASME Section Xl requirements. (Ref. 5.2-39) Ferrite content of theCast austenitic stainless steel will be controlled to be less than or equivalent to 14% for316 materials and 20% for 304 materials calculationq the 6 -ferrite level using Hull'sequivalent factor method or Schoefer Diagram in ASTM AA00 "Standard Practice forSteel GCSting, Austenitic Alloy, Estimating Ferrite Content Thereof' Is used t calculate-6ferite level. (Ref. 5.2 40)

5.2.3-10

Attachment 22-2

5.2.7 References

5.2-40 Standard PrSctice for Stee- MCtinq A,'JteR4tG Alley, Est'm~t*R-q Ferrte 8o0te0tTh4erefASTM 4 20014 800M 0!, 2006. Deleted

5.2.3-11


6/4/2010



Docket No. 52-021





QUESTION NO. : 05.02.03-23

In RAI 05.02.03-12, the staff requested that the applicant discuss its use of standard grades ofstainless steel (non-low-carbon steel). The applicant responded and stated that the basic policyfor use of austenitic stainless steel in the RCPB of the US-APWR is to select low-carbon gradematerials permitting market conditions and availability at the time of procurement. The staff notesthat the use of standard grades is acceptable per RG 1.44 as long as the reactor coolant has acontrolled oxygen content of 0.10 ppm dissolved oxygen. However, elevated dissolved oxygencontent in stagnant or dead-end primary coolant environments has contributed to stress-corrosioncracking of austenitic stainless steel components in operating PWRs. The staff requests that theapplicant verify, and state in the DCD, that a stagnant or dead-end reactor coolant environmentdoes not exist in any portion of the RCBP that could result in elevated dissolved oxygen above0.10 ppm at all operating temperatures above 200OF during normal operation.

ANSWER:

Basic policy of carbon content for use of austenitic stainless steel in the RCPB has alreadydescribed in the MHI letter UAP-HF-10058 "Amended MHI's Response to US-APWR DCD RAINo.289" dated March 1, 2010. RAI Response described that the progress of detail design, if thesignificant stagnant areas where the dissolved oxygen is evaluated to be elevated to above0.10ppm at all operating temperatures above 200 OF, MHI will evaluate the possibility of SCCcaused. Then, if the evaluated portions have possibilities to cause SCC, the low carbon stainlesssteel will be applied for the portions around that area.

Impact on DCD

DCD Table 5.2.3-1 will be revised as shown in the Enclosure 1 of the above MHI letter UAP-HF-10058.

5.2.3-12

Impact on COLA


Impact on PRA


5.2.3-13


6/4/2010



Docket No. 52-021





QUESTION NO. : 05.02.03-24

In RAI 05.02.03-14, the staff requested that the applicant describe the fabrication processrequirements employed to limit the effects of cold work and residual stress, caused by grinding,repair or other fabrication processes on surfaces that come into contact with RCS fluids. Theapplicant responded and stated that, regarding dissimilar metal welds with nickel-based alloywelding material, Alloy 690 filler materials are used to avoid stress-corrosion cracking. The staffnotes that welding repairs, including those made during initial fabrication, have contributed toseveral instances of stress-corrosion cracking in currently operating PWRs. While Alloy 690 andits matching weld filler materials have been shown to be highly resistant to stress-corrosioncracking (SCC), these materials are not immune to SCC.

The staffs expectation is that the grinding of weld surfaces, in contact with reactor coolant, will becontrolled to limit residual stress, thus making these areas less susceptible to SCC. The staff alsoexpects that the applicant will employ process controls that limit the size and number of repaircycles. Therefore, the staff requests that the applicant describe special fabrication processrequirements employed to limit the effects of cold work and residual stress, caused by grinding,repair or other fabrication processes on surfaces that come into contact with RCS fluids in orderto minimize the susceptibility of components to stress-corrosion cracking for the design life of theplant. The applicant's response should cover stainless steel to stainless steel similar metal weldsusing matching stainless steel filler material and dissimilar metal welds between nickel basedalloys and ferritic materials and austenitic stainless steel materials.

5.2.3-14

ANSWER:

Operating experience of the steam generator tubing and various research results show thatsusceptibility to the stress corrosion cracking of alloy 690, including alloy 690 welding materials,are extremely low comparing to alloy 600. MHI knows that an experimental result that show thepossibility of crack propagation of alloy 690 material has published recently, but the test conditionof this test is far beyond the actual plant condition because the test specimen consist of lowforging ratio material and have employed extremely high cold working, that the materials of theactual plants never experienced. In addition there is no experimental result in the world thatshows the possibility of crack initiation of stress corrosion cracking in the PWR condition.

Additionally, MHI conduct a suitable heat treatment process (Thermal Treatment (TT)) to avoidstress corrosion cracking and this TT process is established to be highly resistant to SCC bymany experimental data all over the world. We agree, generally, that alloy 690 materials are notimmune to SCC but we think the possibility to initiate stress corrosion cracking is very low even ifthin layer of the alloy 690 materials have effected by cold working. MHI does not plan to employspecial stress relief fabrication process as a process control except unusual preventive actionmight be employed.

Regarding austenitic stainless steels, cold bend of small diameter piping will be employed by lowresidual stress process with plug-less bending process following the heat treatment to avoidincrease of skin stress. Bending of large diameter piping is employed under hot condition. Buffingover the welding location, which is employed in BWRs, will be employed to the safe-ends ofreactor vessel nozzles and steam generator nozzles. Stress corrosion cracking induced by coldworking has been an outstanding issue appeared recently in BWR condition, however in PWRcondition operating experience shows that the possibility to initiate stress corrosion cracking isvery low comparing to BWR condition because the dissolved oxygen content is controlled under 5ppb in a operating condition, so additional preventive action is not planned as a process control.

Impact on DCD


Impact on COLA


Impact on PRA


5.2.3-15


6/4/2010



Docket No. 52-021





QUESTION NO.: 05.02.03-25

In RAls 05.02.03-15 and 05.02.03-16, the staff requested, in part, that the applicant discusswelding process controls employed to reduce weld-metal dilution to retain the maximumpercentage of chromium possible in order to decrease the susceptibility of components to stress-corrosion cracking. In the applicant's response, it stated that heat input is controlled to preventhot cracking and minimize weld-metal dilution. The staff finds this acceptable, however, theapplicant did not discuss how it verifies that its welding procedures and process controls willresult in dissimilar-metal welds with an acceptable level of chromium to resist stress-corrosioncracking. Therefore, the staff requests that the applicant discuss testing that has been or will beperformed to verify the minimum chromium content in dissimilar-metal welds and that theminimum chromium content is sufficient to resist stress-corrosion cracking.

ANSWER

MHI tested and verified using the residual material of a steam generator safe end of the actualplant. Test result shows that the minimum chromium content at stainless steel in dissimilar-metalwelds is more than 16% by the linear analysis of the weld test coupon of dissimilar-metal welds.It is shown that the minimum chromium content is sufficient to resist stress-corrosion cracking.

Verification test result is proprietary information and attached in the proprietary answer, Enclosure7.

Impact on DCD


Impact on COLA


5.2.3-16

Impact on PRA


5.2.3-17


Enclosure 4


Supplemental Answers to QUESTION 05.02.03-19 ofRAI No.540-4176 Revision 0

Chemical Composition of G-classification Filler Materials(Non Proprietary)

June 2010

5.2.3 Enclosure 4 - 1





Enclosure 6


Supplemental Answers to Question 05.02.03-20 ofRAI No.540-4176 Revision 0

WPS and PQR about preheating temperature ofcircumferential Welding

(Non Proprietary)

June 2010


Supplemental Answer to QUESTION 05.02.03-20



Enclosure 8


Supplemental Answers to QUESTION 05.02.03-25 ofRAI No.540-4167 Revision 0

Verification Test Report of Dilution at Dissimilar Weld(Non Proprietary)

June 2010

5.2.3 Enclosure 8 - I

Verification Test Report of Dilution at Dissimilar Weld


523 Enclosure 8 - 3

523 Enclosure 8 - 4


Documents

MHI's Response to US-APWR DCD RAI No. 540-4176 Revision 0. · Subject: MHI's Response to US-APWR DCD RAI No. 540-4176 REVISION 0 Reference: 1) "Request forAdditional Information No