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Category:CORRESPONDENCE-LETTERS
MONTHYEARML20216J0681999-09-29029 September 1999 Forwards Rev 3 of AP600 Design Control Document, Incorporating Documentation Changes Resulting from Final Review Performed to Check Consistency of Implementation of Approved Design Change Proposals.With Summary of Changes ML20207G5411999-06-0808 June 1999 Discusses Request Made by Westinghouse on 981109 That Proprietary WCAP-14252,Rev 1,be Withheld from Public Disclosure.Determined Info to Be Proprietary & Will Be Withheld from Public Disclosure ML20195D5201999-06-0404 June 1999 Discusses Westinghouse Request That Change Pages Submitted on 980921 to WCAP-14292 Be Withheld from Public Disclosure. Determined Info to Be Proprietary & Will Be Withheld from Public Disclosure ML20196G2431999-05-21021 May 1999 Informs That USNRC Has Published in Fr Encl Notice of Proposed Rulemaking Re AP600 Design Certification Rule. Rulemaking Allows Applicants or Licensees to Construct AP600 Std Plant Design by Referencing Design Certification Rule ML20206G5411999-05-0505 May 1999 Forwards Draft Environ Assessment Re Proposed Certification of AP600 Std Plant Design.Environ Assessment Will Be Used as Basis for NRC Finding of No Significant Environ Impact Resulting from Certification of AP600 Design ML20205J3351999-04-0707 April 1999 Informs That USNRC Staff Has Completed Review of Rev 2 of AP600 Design Control Document,Verified That All of Changes in Rev 2 Are Acceptable & Determined That AP600 Dcd,Rev 2, Can Now Be Used in Proposed Design Certification Rule NSD-NRC-99-5827, Forwards Rev 2 to AP600 Design Control Document.Attachment 1 Provides Summary of Changes Made as Part of Mar 1999 Rev to AP600 Design Control Document1999-03-31031 March 1999 Forwards Rev 2 to AP600 Design Control Document.Attachment 1 Provides Summary of Changes Made as Part of Mar 1999 Rev to AP600 Design Control Document ML20205D3051999-03-25025 March 1999 Requests Amend to 920626 Application for Design Certification of AP600,including AP600 Ssar & AP600 Dcd,To Reflect Sale of CBS Commercial Nuclear Business to W ML20203C5481999-02-10010 February 1999 Forwards Rev 1 to AP600 Design Control Document (DCD) for Docket File.Dcd Provides Reference Basis for AP600 Design Certification ML20199H8671999-01-20020 January 1999 Forwards Comments on AP600 Design Control Document,Submitted by Westinghouse ML20199B1121999-01-0707 January 1999 Advises That Info Contained in NSRA-APSL-92-0268 & Containing Presentation Matl Used in 921209-10 Meeting,Marked Proprietary,Will Be Withheld from Public Disclosure,Per 10CFR2.790(b)(5) & Section 103(b) of AEA ML20198B4601998-12-14014 December 1998 Discusses W Ltr NTD-NRC-95-4556,dtd 950918,provided as Status Rept on Proprietary Matl Submitted to NRC to Support AP600 Design Review Effort.Proprietary Info in Encl Have Been Removed from AP600 Docket File & Being Returned ML20197G3501998-11-30030 November 1998 Forwards AP600 Design Control Document, Vols 1-12 for Docket File.Submittal Closes Confirmatory Items 1.5-1 & 1.5-2 from Sept 1998 Final SER Re Certification of AP600 Std Design ML20195E9331998-11-0909 November 1998 Requests That Rev 1 to WCAP-14252, AP600 Low-Pressure Integral Sys Test at or State Univ Final Data Rept, Vols I- Iv,Be Withheld (Ref 10CFR2.790) NSD-NRC-98-5806, Forwards non-proprietary Rev 1 to WCAP-14253 & Proprietary Rev 1 to WCAP-14252, AP600 Low-Pressure Integral Sys Test at or State Univ Final Data Rept. Proprietary Rept Includes Vols 1-4.Proprietary Info Withheld,Per 10CFR2.7901998-11-0909 November 1998 Forwards non-proprietary Rev 1 to WCAP-14253 & Proprietary Rev 1 to WCAP-14252, AP600 Low-Pressure Integral Sys Test at or State Univ Final Data Rept. Proprietary Rept Includes Vols 1-4.Proprietary Info Withheld,Per 10CFR2.790 ML20155G9021998-11-0303 November 1998 Advises That Info Contained in 981012 ltr,NSD-NRC-98-5795, Will Be Withheld from Public Disclosure,Per 10CFR2.790(b) (5) & Section 103(b) of AEA of 1954,as Amended ML20155G9271998-11-0202 November 1998 Forwards Copy of NRC Staff FSER for Westinghouse AP600 Design.Staff Completed Review of Design & Issued Final Design Approval & FSER on 980903.Without Encl ML20155G7591998-11-0202 November 1998 Forwards Copy of NRC Staff FSER for Westinghouse AP600 Design,Per Discussion at Sept 1997 Meeting.Staff Completed & Issued Final Design Approval of FSER on 980903.Without Encl NSD-NRC-98-5795, Informs That Figures from Rev 0 & Rev 3 of PRA Which Indicate Location of H Igniters Should Be Considered Proprietary.Nonproprietary Versions of AP600 General Arrangement Drawings Were Provided in Rev 7 of Ssar1998-10-12012 October 1998 Informs That Figures from Rev 0 & Rev 3 of PRA Which Indicate Location of H Igniters Should Be Considered Proprietary.Nonproprietary Versions of AP600 General Arrangement Drawings Were Provided in Rev 7 of Ssar ML20154A4091998-09-29029 September 1998 Advises That Certain Info Contained in Westinghouse Ltr NTD-NRC-95-4506,dtd 950713,submitting WCAP-14425,will Be Withheld from Public Disclosure,Per 10CFR2.790(b)(5).Figure 3-1 Will Be Placed Into Public Record ML20153G2521998-09-25025 September 1998 Advises That Proprietary Matl Discovered by NRC in NSD-NRC-97-4966 & Proprietary Matl Noted by W in NSD-NRC-98-5772 Will Be Withhheld from Public Disclosure Per 10CFR2.790(b)(5) & Section 103(b) of AEA of 1954 ML20153C1001998-09-15015 September 1998 Advises That Matls Re AP600 Notrump Final Validation Rept, WCAP-14807,marked Proprietary Will Be Withheld from Public Disclosure,Per 10CFR2.790(b)(5) & Section 103(b) of AEA of 1954,as Amended NSD-NRC-98-5788, Forwards Formal Transmittal of Correspondence Previously Sent Informally Over Period of 980702-0826.Index of Encl Matl Provided1998-09-15015 September 1998 Forwards Formal Transmittal of Correspondence Previously Sent Informally Over Period of 980702-0826.Index of Encl Matl Provided ML20151X4381998-09-15015 September 1998 Informs That Staff Has Decided to Accept Claim That Info in WCAP-14135,Rev 1 Is Proprietary & Will Be Withheld from Public Disclosure,Per W 980821 & 26 Ltrs ML20151X4041998-09-11011 September 1998 Discusses Revised Tier 2 Info for AP600 Design.Staff Revised Decision on Whether Fire Protection Should Expire at First Full Power Encl ML20151V1711998-09-0808 September 1998 Advises That Info Re Westinghouse AP600 Std Safety Analysis Rept Through Rev 4 & PRA Through Rev 5 Will Be Withheld from Public Disclosure,Per 10CFR2.790(b)(5).Disposition of Ssar & PRA Proprietary Info Encl ML20151V2231998-09-0808 September 1998 Informs That NRC Determined That WCAP-14132 Encl in Westinghouse Ltr NTD-NRC-94-4244,dtd 940729 & Marked as Proprietary,Will Be Withheld from Public Disclosure,Per to 10CFR2.790(b)(5) & Section 103(b) of AEA of 1954 ML20151S9181998-09-0303 September 1998 Advises That Info Marked as Proprietary Re Westinghouse AP600 Design Ltrs Concerning Pxs Scaling & Pirt Closure Rept WCAP-14727,will Be Withheld from Public Disclosure,Per 10CFR2.790(b)(5) ML20239A1481998-09-0303 September 1998 Forwards Notice of Issuance of Final Design Approval & Final SER for AP600.FDA Allows AP600 Design to Be Referenced in Application for Construction Permit or Operating License Under 10CFR50 or Application for Combined License ML20239A3111998-09-0303 September 1998 Forwards Final SER Which Summarizes Staff Safety Review of AP600 Design Against Requirements of Subpart B of 10CF5R52 & Delineates Scope of Technical Details Considered in Evaluating Proposed Design ML20151V8521998-09-0101 September 1998 Extends Invitation to Attend Ceremony on 980911,where NRC Will Present Final Design Approval for AP600 Std Nuclear Reactor Design to Westinghouse NSD-NRC-98-5781, Informs That W Determined That AP600 FSER Contains No Proprietary Info1998-09-0101 September 1998 Informs That W Determined That AP600 FSER Contains No Proprietary Info ML20151V2201998-08-31031 August 1998 Informs That EPRI Documents, GOTHIC Containment Analysis Package Qualification Rept, GOTHIC Containment Analysis Qualification Manual, & Listed Documents Dtd Sept 1993 Will Be Withheld from Public Disclosure ML20151V8431998-08-31031 August 1998 Extends Invitation to Attend Ceremony on 980911,where NRC Will Present Final Design Approval for AP600 Std Nuclear Reactor Design to Westinghouse ML20238F3241998-08-31031 August 1998 Advises That AP600 RAI Responses Encl in Westinghouse Ltrs NTD-NRC-95-4598,dtd 951117,as Modified by NSD-NRC-98-5776, Dtd 980826,marked Proprietary Will Be Withheld from Public Disclosure,Per 10CFR2.790(b)(5) & Section 103(b) ML20238F5991998-08-31031 August 1998 Refers to W 980821 Revised Response to Insp Rept 99900404/97-01 That Contained All Substantive Info Provided w/DCP/NRC-1074 Ltr.Nrc Will Destroy DCP/NRC-1074,as Requested NSD-NRC-98-5783, Forwards non-proprietary Rev 2 to WCAP-14989, Accident Specification & Phenomena Evaluation for AP600 Passive Containment Cooling Sys1998-08-28028 August 1998 Forwards non-proprietary Rev 2 to WCAP-14989, Accident Specification & Phenomena Evaluation for AP600 Passive Containment Cooling Sys NSD-NRC-98-5782, Forwards Corrected Pages 4-28 Through 4-32 to Be Inserted in Rev 2 to non-proprietary WCAP-14953, AP600 Pxs Scaling & Pirt Closure Rept. Pages Were Originally Submitted W/Incorrect Header Which Stated 'W Proprietary Class 2.'1998-08-28028 August 1998 Forwards Corrected Pages 4-28 Through 4-32 to Be Inserted in Rev 2 to non-proprietary WCAP-14953, AP600 Pxs Scaling & Pirt Closure Rept. Pages Were Originally Submitted W/Incorrect Header Which Stated 'W Proprietary Class 2.' NSD-NRC-98-5778, Provides Commitment to Include Requested Changes Into AP600 Design Control Document.Discussion W/Vijuk on 980825 Re Disposition of Ltrs DCP/NRC1074,dtd 971016 & DCP/NRC1324,dtd 980403,formalized1998-08-27027 August 1998 Provides Commitment to Include Requested Changes Into AP600 Design Control Document.Discussion W/Vijuk on 980825 Re Disposition of Ltrs DCP/NRC1074,dtd 971016 & DCP/NRC1324,dtd 980403,formalized ML20237E4421998-08-27027 August 1998 Advises That Info Contained in Ltr DCP/NRC-0985,dtd 970821, Sought to Be Withheld,Will Be Withheld from Public Disclosure,Per 10CFR2.790(b)(5) & Section 103(b) of AEA of 1954,as Amended NSD-NRC-98-5777, Provides Revised Response to NRC Ltr Re W Requests for Withholding Info Re AP600 PCS Final Data rept,WCAP-14135 & WCAP-141381998-08-26026 August 1998 Provides Revised Response to NRC Ltr Re W Requests for Withholding Info Re AP600 PCS Final Data rept,WCAP-14135 & WCAP-14138 NSD-NRC-98-5776, Forwards Proprietary Revised Response to Ref NRC Ltr Re W Requests for Withholding Info Re AP600 Design Certification Test Program,Notrump Computer Code,Wcobra/Trac Computer Code & Loftran Computer Code.Proprietary Info Withheld1998-08-26026 August 1998 Forwards Proprietary Revised Response to Ref NRC Ltr Re W Requests for Withholding Info Re AP600 Design Certification Test Program,Notrump Computer Code,Wcobra/Trac Computer Code & Loftran Computer Code.Proprietary Info Withheld ML20237E0731998-08-26026 August 1998 Advises That Info in WCAP-14812,revs 1 & 2, Accident Spec & Phenomena Evaluation for AP600 PCS, Will Be Withheld from Public Disclosure ML20238F8031998-08-26026 August 1998 Requests That Proprietary W Revised Response to NRC Ltrs Re Requests for Withholding Info,Be Withheld from Public Disclosure IAW 10CFR2.790 NSD-NRC-98-5774, Informs That W Faxed Last Response Re Afser Open Item 1.1-1 to NRC on 980821.Last of Attachments Supporting Several of Responses Faxed to NRC Over Past Several Days Were Express Mailed to NRC on 980821.Responses Close Subject Open It1998-08-24024 August 1998 Informs That W Faxed Last Response Re Afser Open Item 1.1-1 to NRC on 980821.Last of Attachments Supporting Several of Responses Faxed to NRC Over Past Several Days Were Express Mailed to NRC on 980821.Responses Close Subject Open Item NSD-NRC-98-5773, Responds to Questions Raised in NRC 971022 & s Re Proprietary Info Contained in 970930 Summary of Meeting Held on 970804-15 Concerning Structural Design of AP600.Figure 12-24 of WCAP-14407 Still Considered Proprietary by W1998-08-21021 August 1998 Responds to Questions Raised in NRC 971022 & s Re Proprietary Info Contained in 970930 Summary of Meeting Held on 970804-15 Concerning Structural Design of AP600.Figure 12-24 of WCAP-14407 Still Considered Proprietary by W ML20237E7991998-08-21021 August 1998 Forwards non-proprietary Results of AP600 Design Assurance Review (Dar) That Was Commitment in W Response to NRC Insp Rept 99900404/97-02.Info Should Assist in Closing Nonconformances & Unresolved Item Identified in Insp Rept NSD-NRC-98-5771, Forwards non-proprietary Rev 2 to WCAP-14138, Final Data Rept for PCS Large-Scale Tests,Phase 2 & Phase 3. Submittal Satisfies Verbal Commitment to Revise non-proprietary Version of PCS Final Data Rept Made by W1998-08-21021 August 1998 Forwards non-proprietary Rev 2 to WCAP-14138, Final Data Rept for PCS Large-Scale Tests,Phase 2 & Phase 3. Submittal Satisfies Verbal Commitment to Revise non-proprietary Version of PCS Final Data Rept Made by W NSD-NRC-98-5769, Responds to NRC Ref Ltrs Re W Requests for Withholding Info. Info Provided as Proprietary in Ref 1 Has Either Been Moved from Ssar to Ref 3,which Includes non-proprietary Version of Rept or Made non-proprietary in Current Version of1998-08-21021 August 1998 Responds to NRC Ref Ltrs Re W Requests for Withholding Info. Info Provided as Proprietary in Ref 1 Has Either Been Moved from Ssar to Ref 3,which Includes non-proprietary Version of Rept or Made non-proprietary in Current Version of Ssar NSD-NRC-98-5772, Forwards non-proprietary & Proprietary Info in Response to NRC Ltrs Re W Requests for Withholding Info.Separate Ltr & Affidavit Justifying Proprietary Nature of Info,Encl. Proprietary Info Withheld,Per 10CFR2.7901998-08-21021 August 1998 Forwards non-proprietary & Proprietary Info in Response to NRC Ltrs Re W Requests for Withholding Info.Separate Ltr & Affidavit Justifying Proprietary Nature of Info,Encl. Proprietary Info Withheld,Per 10CFR2.790 1999-09-29
[Table view] Category:INCOMING CORRESPONDENCE
MONTHYEARML20216J0681999-09-29029 September 1999 Forwards Rev 3 of AP600 Design Control Document, Incorporating Documentation Changes Resulting from Final Review Performed to Check Consistency of Implementation of Approved Design Change Proposals.With Summary of Changes NSD-NRC-99-5827, Forwards Rev 2 to AP600 Design Control Document.Attachment 1 Provides Summary of Changes Made as Part of Mar 1999 Rev to AP600 Design Control Document1999-03-31031 March 1999 Forwards Rev 2 to AP600 Design Control Document.Attachment 1 Provides Summary of Changes Made as Part of Mar 1999 Rev to AP600 Design Control Document ML20205D3051999-03-25025 March 1999 Requests Amend to 920626 Application for Design Certification of AP600,including AP600 Ssar & AP600 Dcd,To Reflect Sale of CBS Commercial Nuclear Business to W ML20203C5481999-02-10010 February 1999 Forwards Rev 1 to AP600 Design Control Document (DCD) for Docket File.Dcd Provides Reference Basis for AP600 Design Certification ML20197G3501998-11-30030 November 1998 Forwards AP600 Design Control Document, Vols 1-12 for Docket File.Submittal Closes Confirmatory Items 1.5-1 & 1.5-2 from Sept 1998 Final SER Re Certification of AP600 Std Design NSD-NRC-98-5806, Forwards non-proprietary Rev 1 to WCAP-14253 & Proprietary Rev 1 to WCAP-14252, AP600 Low-Pressure Integral Sys Test at or State Univ Final Data Rept. Proprietary Rept Includes Vols 1-4.Proprietary Info Withheld,Per 10CFR2.7901998-11-0909 November 1998 Forwards non-proprietary Rev 1 to WCAP-14253 & Proprietary Rev 1 to WCAP-14252, AP600 Low-Pressure Integral Sys Test at or State Univ Final Data Rept. Proprietary Rept Includes Vols 1-4.Proprietary Info Withheld,Per 10CFR2.790 ML20195E9331998-11-0909 November 1998 Requests That Rev 1 to WCAP-14252, AP600 Low-Pressure Integral Sys Test at or State Univ Final Data Rept, Vols I- Iv,Be Withheld (Ref 10CFR2.790) NSD-NRC-98-5795, Informs That Figures from Rev 0 & Rev 3 of PRA Which Indicate Location of H Igniters Should Be Considered Proprietary.Nonproprietary Versions of AP600 General Arrangement Drawings Were Provided in Rev 7 of Ssar1998-10-12012 October 1998 Informs That Figures from Rev 0 & Rev 3 of PRA Which Indicate Location of H Igniters Should Be Considered Proprietary.Nonproprietary Versions of AP600 General Arrangement Drawings Were Provided in Rev 7 of Ssar NSD-NRC-98-5788, Forwards Formal Transmittal of Correspondence Previously Sent Informally Over Period of 980702-0826.Index of Encl Matl Provided1998-09-15015 September 1998 Forwards Formal Transmittal of Correspondence Previously Sent Informally Over Period of 980702-0826.Index of Encl Matl Provided NSD-NRC-98-5781, Informs That W Determined That AP600 FSER Contains No Proprietary Info1998-09-0101 September 1998 Informs That W Determined That AP600 FSER Contains No Proprietary Info NSD-NRC-98-5782, Forwards Corrected Pages 4-28 Through 4-32 to Be Inserted in Rev 2 to non-proprietary WCAP-14953, AP600 Pxs Scaling & Pirt Closure Rept. Pages Were Originally Submitted W/Incorrect Header Which Stated 'W Proprietary Class 2.'1998-08-28028 August 1998 Forwards Corrected Pages 4-28 Through 4-32 to Be Inserted in Rev 2 to non-proprietary WCAP-14953, AP600 Pxs Scaling & Pirt Closure Rept. Pages Were Originally Submitted W/Incorrect Header Which Stated 'W Proprietary Class 2.' NSD-NRC-98-5783, Forwards non-proprietary Rev 2 to WCAP-14989, Accident Specification & Phenomena Evaluation for AP600 Passive Containment Cooling Sys1998-08-28028 August 1998 Forwards non-proprietary Rev 2 to WCAP-14989, Accident Specification & Phenomena Evaluation for AP600 Passive Containment Cooling Sys NSD-NRC-98-5778, Provides Commitment to Include Requested Changes Into AP600 Design Control Document.Discussion W/Vijuk on 980825 Re Disposition of Ltrs DCP/NRC1074,dtd 971016 & DCP/NRC1324,dtd 980403,formalized1998-08-27027 August 1998 Provides Commitment to Include Requested Changes Into AP600 Design Control Document.Discussion W/Vijuk on 980825 Re Disposition of Ltrs DCP/NRC1074,dtd 971016 & DCP/NRC1324,dtd 980403,formalized ML20238F8031998-08-26026 August 1998 Requests That Proprietary W Revised Response to NRC Ltrs Re Requests for Withholding Info,Be Withheld from Public Disclosure IAW 10CFR2.790 NSD-NRC-98-5776, Forwards Proprietary Revised Response to Ref NRC Ltr Re W Requests for Withholding Info Re AP600 Design Certification Test Program,Notrump Computer Code,Wcobra/Trac Computer Code & Loftran Computer Code.Proprietary Info Withheld1998-08-26026 August 1998 Forwards Proprietary Revised Response to Ref NRC Ltr Re W Requests for Withholding Info Re AP600 Design Certification Test Program,Notrump Computer Code,Wcobra/Trac Computer Code & Loftran Computer Code.Proprietary Info Withheld NSD-NRC-98-5777, Provides Revised Response to NRC Ltr Re W Requests for Withholding Info Re AP600 PCS Final Data rept,WCAP-14135 & WCAP-141381998-08-26026 August 1998 Provides Revised Response to NRC Ltr Re W Requests for Withholding Info Re AP600 PCS Final Data rept,WCAP-14135 & WCAP-14138 NSD-NRC-98-5774, Informs That W Faxed Last Response Re Afser Open Item 1.1-1 to NRC on 980821.Last of Attachments Supporting Several of Responses Faxed to NRC Over Past Several Days Were Express Mailed to NRC on 980821.Responses Close Subject Open It1998-08-24024 August 1998 Informs That W Faxed Last Response Re Afser Open Item 1.1-1 to NRC on 980821.Last of Attachments Supporting Several of Responses Faxed to NRC Over Past Several Days Were Express Mailed to NRC on 980821.Responses Close Subject Open Item ML20237E7991998-08-21021 August 1998 Forwards non-proprietary Results of AP600 Design Assurance Review (Dar) That Was Commitment in W Response to NRC Insp Rept 99900404/97-02.Info Should Assist in Closing Nonconformances & Unresolved Item Identified in Insp Rept ML20237D3161998-08-21021 August 1998 Requests That W Response to NRC Ltrs Re Requests for Withholding Info Be Withheld from Public Disclosure,Per 10CFR2.790 NSD-NRC-98-5773, Responds to Questions Raised in NRC 971022 & s Re Proprietary Info Contained in 970930 Summary of Meeting Held on 970804-15 Concerning Structural Design of AP600.Figure 12-24 of WCAP-14407 Still Considered Proprietary by W1998-08-21021 August 1998 Responds to Questions Raised in NRC 971022 & s Re Proprietary Info Contained in 970930 Summary of Meeting Held on 970804-15 Concerning Structural Design of AP600.Figure 12-24 of WCAP-14407 Still Considered Proprietary by W NSD-NRC-97-5370, Transmits Results of AP600 Foake Design Assurance Review as Requested in NRC .Info Is Being Provided to Close Unresolved Item Identified in Insp Rept 99900404/97-011998-08-21021 August 1998 Transmits Results of AP600 Foake Design Assurance Review as Requested in NRC .Info Is Being Provided to Close Unresolved Item Identified in Insp Rept 99900404/97-01 NSD-NRC-98-5763, Responds to NRC Ltrs Re W Claim for Treatment of Proprietary Info Submitted in .As Discussed W/Nrc on 980821, Info Contained in W Is self-critical Analysis of W QA Program & Therefore Falls Under Items of Affidavit1998-08-21021 August 1998 Responds to NRC Ltrs Re W Claim for Treatment of Proprietary Info Submitted in .As Discussed W/Nrc on 980821, Info Contained in W Is self-critical Analysis of W QA Program & Therefore Falls Under Items of Affidavit NSD-NRC-98-5772, Forwards non-proprietary & Proprietary Info in Response to NRC Ltrs Re W Requests for Withholding Info.Separate Ltr & Affidavit Justifying Proprietary Nature of Info,Encl. Proprietary Info Withheld,Per 10CFR2.7901998-08-21021 August 1998 Forwards non-proprietary & Proprietary Info in Response to NRC Ltrs Re W Requests for Withholding Info.Separate Ltr & Affidavit Justifying Proprietary Nature of Info,Encl. Proprietary Info Withheld,Per 10CFR2.790 NSD-NRC-98-5769, Responds to NRC Ref Ltrs Re W Requests for Withholding Info. Info Provided as Proprietary in Ref 1 Has Either Been Moved from Ssar to Ref 3,which Includes non-proprietary Version of Rept or Made non-proprietary in Current Version of1998-08-21021 August 1998 Responds to NRC Ref Ltrs Re W Requests for Withholding Info. Info Provided as Proprietary in Ref 1 Has Either Been Moved from Ssar to Ref 3,which Includes non-proprietary Version of Rept or Made non-proprietary in Current Version of Ssar NSD-NRC-98-5771, Forwards non-proprietary Rev 2 to WCAP-14138, Final Data Rept for PCS Large-Scale Tests,Phase 2 & Phase 3. Submittal Satisfies Verbal Commitment to Revise non-proprietary Version of PCS Final Data Rept Made by W1998-08-21021 August 1998 Forwards non-proprietary Rev 2 to WCAP-14138, Final Data Rept for PCS Large-Scale Tests,Phase 2 & Phase 3. Submittal Satisfies Verbal Commitment to Revise non-proprietary Version of PCS Final Data Rept Made by W ML20237D3551998-08-20020 August 1998 Requests That Proprietary Info Re W Response to NRC Ltrs Re Requests for Withholding Info Concerning AP600 Pxs Scaling & Pirt Closure Rept, WCAP-14727 Be Withheld from Public Disclosure IAW 10CFR2.790 NSD-NRC-98-5762, Forwards Proprietary Change Pages for Chapters 7 & 8,for Inclusion as Rev 2 of TR WCAP-14727,in Response to NRC Ltr Dtd 980714.Non-proprietary Rev 2 to TR WCAP-14953 Re AP600 Scaling & Pirt Closure Encl.Proprietary Info Withheld1998-08-20020 August 1998 Forwards Proprietary Change Pages for Chapters 7 & 8,for Inclusion as Rev 2 of TR WCAP-14727,in Response to NRC Ltr Dtd 980714.Non-proprietary Rev 2 to TR WCAP-14953 Re AP600 Scaling & Pirt Closure Encl.Proprietary Info Withheld NSD-NRC-98-5764, Responds to NRC Ltrs Re Requests for Withholding Info from Public Disclosure for W AP600 Design Ltr of 951117,dtd 960229.W Considers Matl Proprietary Since Matl Discusses Approach Used by W to Develop Analysis Models of Cctf Test1998-08-20020 August 1998 Responds to NRC Ltrs Re Requests for Withholding Info from Public Disclosure for W AP600 Design Ltr of 951117,dtd 960229.W Considers Matl Proprietary Since Matl Discusses Approach Used by W to Develop Analysis Models of Cctf Test NSD-NRC-98-5768, Responds to NRC Ref Ltrs Re W Requests for Withholding Info. Specific Info Considered non-proprietary or Proprietary Discussed1998-08-20020 August 1998 Responds to NRC Ref Ltrs Re W Requests for Withholding Info. Specific Info Considered non-proprietary or Proprietary Discussed NSD-NRC-98-5766, Provides Summary of 980813 Telcon W/Nrc Re W Explanation of Rationale Which Was Used to Determine What Info Was Considered to Be Proprietary in TR WCAP-148121998-08-20020 August 1998 Provides Summary of 980813 Telcon W/Nrc Re W Explanation of Rationale Which Was Used to Determine What Info Was Considered to Be Proprietary in TR WCAP-14812 NSD-NRC-98-5770, Responds to NRC Ltrs Re W Requests for Withholding Info.W Withdraws AP600 Rept Entitled, Wcobra/Trac Core Makeup Tank Preliminary Validation Rept, Since Rept Was Not Required by NRC to Make Safety Determination of AP6001998-08-20020 August 1998 Responds to NRC Ltrs Re W Requests for Withholding Info.W Withdraws AP600 Rept Entitled, Wcobra/Trac Core Makeup Tank Preliminary Validation Rept, Since Rept Was Not Required by NRC to Make Safety Determination of AP600 NSD-NRC-98-5760, Forwards Rev 25 to GW-GL-021, AP600 Ssar, Vols 1-11. Revised Tables of Contents,Change Page Instructions,List of Effective Pages,Document Cover Sheet & Change Roadmap Outlining Changes in Each Section Also Encl1998-08-19019 August 1998 Forwards Rev 25 to GW-GL-021, AP600 Ssar, Vols 1-11. Revised Tables of Contents,Change Page Instructions,List of Effective Pages,Document Cover Sheet & Change Roadmap Outlining Changes in Each Section Also Encl NSD-NRC-98-5761, Provides Written Confirmation That Identified Figure Is Not Considered Proprietary by W.Refs 1-5,discussed1998-08-19019 August 1998 Provides Written Confirmation That Identified Figure Is Not Considered Proprietary by W.Refs 1-5,discussed ML20237D2681998-08-18018 August 1998 Requests That Proprietary Informal Correspondence Be Withheld from Public Disclosure,Per 10CFR2.790 NSD-NRC-97-5046, Submits Rev 1 of W Ltr Originally Submitted 970917,which Transmitted Proprietary & non-proprietary Informal Correspondence.Rev Includes Notarized Affidavit as Part of Application for Withholding Info.Proprietary Info Withheld1998-08-18018 August 1998 Submits Rev 1 of W Ltr Originally Submitted 970917,which Transmitted Proprietary & non-proprietary Informal Correspondence.Rev Includes Notarized Affidavit as Part of Application for Withholding Info.Proprietary Info Withheld NSD-NRC-98-5759, Submits Response to NRC Ltrs Re Request for Withholding Info.Figure 8-19 Will No Longer Be Considered Proprietary by W1998-08-17017 August 1998 Submits Response to NRC Ltrs Re Request for Withholding Info.Figure 8-19 Will No Longer Be Considered Proprietary by W NSD-NRC-98-5756, Responds to NRC 980714 & 21 Ltrs Re Request for Withholding Proprietary Info That Was Not Clearly Identified Other than Being Marked W Proprietary Class 21998-08-14014 August 1998 Responds to NRC 980714 & 21 Ltrs Re Request for Withholding Proprietary Info That Was Not Clearly Identified Other than Being Marked W Proprietary Class 2 NSD-NRC-98-5757, Responds to Ref NRC Ltrs Re Request for Withholding Proprietary Info Re W AP600 Ltr.Per 980708 Telcon,W Has Reviewed TRs WCAP-13288 & WCAP-13289 & Considers None of Info to Be Proprietary1998-08-14014 August 1998 Responds to Ref NRC Ltrs Re Request for Withholding Proprietary Info Re W AP600 Ltr.Per 980708 Telcon,W Has Reviewed TRs WCAP-13288 & WCAP-13289 & Considers None of Info to Be Proprietary ML20151Z0971998-08-13013 August 1998 Requests That Proprietary Rev 5 to WCAP-14807, Notrump Final Validation Rept, Be Withheld from Public Disclosure, Per 10CFR2.790 NSD-NRC-98-5754, Forwards non-proprietary Versions of Revs 3-5 to WCAP-14808 & Proprietary Version of Rev 5 to WCAP-14807, Notrump Final Validation Rept for AP600, in Form of Replacement Pages. Proprietary Encl Withheld1998-08-13013 August 1998 Forwards non-proprietary Versions of Revs 3-5 to WCAP-14808 & Proprietary Version of Rev 5 to WCAP-14807, Notrump Final Validation Rept for AP600, in Form of Replacement Pages. Proprietary Encl Withheld NSD-NRC-98-5749, Forwards Rev 13 to AP600 PRA for Simplified Passive Advanced LWR Plant Program. Rev Represents Final Version of AP600 Pra.All NRC Comments Related to Rev 12 of AP600 PRA Have Been Resolved1998-08-13013 August 1998 Forwards Rev 13 to AP600 PRA for Simplified Passive Advanced LWR Plant Program. Rev Represents Final Version of AP600 Pra.All NRC Comments Related to Rev 12 of AP600 PRA Have Been Resolved NSD-NRC-98-5753, Responds to Ref NRC Ltrs & W Ltr Re Request for Withholding Proprietary Info Re AP600 Design Ltrs1998-08-13013 August 1998 Responds to Ref NRC Ltrs & W Ltr Re Request for Withholding Proprietary Info Re AP600 Design Ltrs NSD-NRC-98-5752, Responds to Open Item 1.1-2 Contained in 980501 Advance Final SER for AP600 Requesting Update of Comparison of AP600 to NRC Reviewed Version of Alwr Util Requirements Document. Changes Resulting from NRC Comments Re Plant Design,List1998-08-13013 August 1998 Responds to Open Item 1.1-2 Contained in 980501 Advance Final SER for AP600 Requesting Update of Comparison of AP600 to NRC Reviewed Version of Alwr Util Requirements Document. Changes Resulting from NRC Comments Re Plant Design,Listed NSD-NRC-98-5755, Forwards Rev 7 to Simplified Passive Advance Light Water Reactor Plant Program AP600 Tier 1 Matl, Incorporating Comments Received from NRC Technical Staff,As of 980812.Encl Closes Open Item 14.3-1 from Advanced Final SER1998-08-13013 August 1998 Forwards Rev 7 to Simplified Passive Advance Light Water Reactor Plant Program AP600 Tier 1 Matl, Incorporating Comments Received from NRC Technical Staff,As of 980812.Encl Closes Open Item 14.3-1 from Advanced Final SER ML20153D9491998-08-12012 August 1998 Submits Response to NRC Ltrs Re Requests for Withholding Info.Info Does Not Have Commercial Value & No Longer Considered to Be Proprietary by W NSD-NRC-98-5751, Submits Response to NRC Ltrs Re Requests for Withholding Info.Info Does Not Have Commercial Value & Is No Longer Considered Proprietary by W1998-08-12012 August 1998 Submits Response to NRC Ltrs Re Requests for Withholding Info.Info Does Not Have Commercial Value & Is No Longer Considered Proprietary by W NSD-NRC-98-5748, Forwards Rev 24 of AP600 Ssar.Revised Tables of Contents, Change Page Instructions,List of Effective Pages,Document Cover Sheet & Change Roadmap Outlining Changes in Each Section Also Encl.Rev Submitted Under Encl Oath1998-08-0707 August 1998 Forwards Rev 24 of AP600 Ssar.Revised Tables of Contents, Change Page Instructions,List of Effective Pages,Document Cover Sheet & Change Roadmap Outlining Changes in Each Section Also Encl.Rev Submitted Under Encl Oath NSD-NRC-98-5743, Forwards Table Which Compares Combined License (COL) Info Items Identified in AP600 Design Certification Application W/Col Action Items Identified in AP600 Advance Final SER Provided to W by NRC on 9805061998-07-31031 July 1998 Forwards Table Which Compares Combined License (COL) Info Items Identified in AP600 Design Certification Application W/Col Action Items Identified in AP600 Advance Final SER Provided to W by NRC on 980506 NSD-NRC-98-5733, Forwards Rev 6 to GW-GL-030, Simplified Passive Advance LWR Plant Program,AP600 Tier 1 Matl, Which Incorporates Comments Received from NRC Technical Staff as of 9807221998-07-27027 July 1998 Forwards Rev 6 to GW-GL-030, Simplified Passive Advance LWR Plant Program,AP600 Tier 1 Matl, Which Incorporates Comments Received from NRC Technical Staff as of 980722 NSD-NRC-98-5742, Forwards Rev 12 to AP600 PRA Rept. All NRC Open Items Related to AP600 PRA Have Been Resolved1998-07-24024 July 1998 Forwards Rev 12 to AP600 PRA Rept. All NRC Open Items Related to AP600 PRA Have Been Resolved 1999-09-29
[Table view] |
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/NEA idaho National Engineering laboratory Q 00 July 21,1992 Dr. L. M. Shotkin U. S. Nuclear Regulatory Commission 5640 Nicholson Lane MS NL/N-353 Washington, DC 20555 COMMENTS ON DR. ZUBER REPORT TO DR. CATTON ON ROSA-IV/AP600 MEETING, JUNE 3-4, FROM JUNE 23, 1992 - SMM-38-92
Dear Dr. Shotkin:
I have read Dr. Zuber's report to Dr. Catton from June 23, 1992. Upon l reading his comments, I have the impression he discusses application of the ROSA-IV test facility for demonstration type tests or AP600 simulation rather than for generation of data for code assessment. His argumentation .
of significance of distortions focuses on the capabilities of ROSA-IV to I duplicate the AP600 behavior. Irregardless of ROSA-IV's shortcomings, duplication of reference system behavior in scaled facilities under two-phase transient flow conditions is a priori impossible. On page 13 of his memorandum, Dr. Zuber observed critically that ROSA-IV will generate conservative as well as non-conservative results. I believe test results must be understood, but the aspect of conservatism is irrelevant for gathering test data for code assessment.
The work we have performed so far is aimed at evaluation of ROSA-IV as a facility that should provide data to assess capabilities of computer codes to model the overall system behavior during postulated transients. Our analyses showed that ROSA will exhibit most of the AP600 processes. The sequence and relative magnitude of events will be reproduced by ROSA-IV for most transients. However, ROSA will not be a AP600 simulator or a demonstration facility. Based upon our analyses, we believe that ROSA-IV can provide very useful data for assessment of ccm.nuter codes to simulate AP600 integral system behavior during high pressure and depressurization phases.
On page 11 of his memorandum, Dr. Zuber discusses the issue of asymmetrical behavior of AP600 and respective code capabilities:
...RELAP5, being a one-dimensional code, cannot properly model the effects of flow asymmetries." Because the ROSA shortcomings in simulation of the asymmetries was considered as the most important, I would like to discuss this aspect in more detail.
9408040306 940629 PDR COMMS NRCC CORRESPONDENCE PDR jfEGcG,..Inc. P.O. Box 1625 Idaho Falls, ID 83415
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Dr. L. M. Shotkin !
July 21, 1992 ;
SMM-38-92 i Page 2 Potential asymmetry in AP600 response is associated with flows in such components as the cold legs, pressure balancing lines, direct vessel l injection (DVI) lines and downcomer. The asymmetry is a result of i interactions between these components for a set of transients. Most of ,
4 the DBA transient will be of a symmetrical nature in AP600 with respect to l safety systems response. Only transients with breaks in DVI lines or pressure balancing lines will exhibit asymmetrical behavior not typical for current generation reactors.
Flows in all the components of interest, except in the downcomer, can be treated one-dimensionally as it is practiced for present generation reactors (all current system codes such as TRAC or RELAP treat piping flow as one-dimensional). Because of potential multi-dimensional effects in the downcomer we have nodalized it applying interconnected mesh of cells.
However, to increase the fidelity of simulation, we suggested a rigorous two-dimensional modeling of the downcomer. Furthermore, we are currently analyzing the downcomer behavior of both systems with a 3-D CFD code I (FIDAP), to obtain an independent evaluation of ROSA /AP600 comparison.
1 Current analyses indicate that ROSA-IV can exhibit the local phenomena '
that control AP600 response. With the currently proposed configuration .
. ROSA-IV will be able to provide data on symmetrical system response and on l asymetrical response. The asymmetrical transients will not duplicate AP600 system behavior but will provide, in an integral system environment, interactions and phenomena that govern the asymmetric AP600 behavior.
Codes validated using these data should be able to simulate AP600 symmetric and asymmetric system response.
Si r i
i S. M. Modro l
NRC Thermal Hydraulic Analysis Programs dap cc: D. Bessette, US/NRC W. H. Rettig, DOE Field Office, Idaho, MS 1134 J. C. Okeson, EG&G Idaho, MS 3600
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"** miwf( f Nuclear Safety Research Review Committee
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.' : 1cly 1992
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i Mr. Eric S. Beckjord Director Office of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission l Washington,DC 20555
Dear Mr. Beckjord:
Enclosed please find a copy of a letter report of NSRRCs ALWR Subcommittee on AP600 thermal hydraulic t: sting. This letter report has been received and reviewed l
l by the NSRRC and is accepted as a statement of the Committee's cunent position on AP600 thermal hydraulic testing.
l If you have any questions on this NSRRC report, please contact Dr. Neil Todreas i
or me.
l Sincerely,
. / n.
David L. Morrison l
Chairman Nuclear Safety Research Review Committee l
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e Nuclear Safety Research Review Committee S, # Vvashington. O.C. 2o555 July 20,1992 Dr. David Morrison
! The MITRE Corporanon 7525 Cc! shire Drive. MC W766 M: Lean, VA 12102 I
Dear Dr. Momson.
The NSRRC Advancec Reactors Subcommittee (Messrs T Boulette. S. Burstem.
H. Ishin and N. Todreas (Chairman) in attendance) met on July 1 and 2.1992, and reviewed Office of Nuclear Regulatory Research (RES) programs pertammg prmespally to the AP 600 l program. Among these programs the RES prt>posaj to conduct integral systems tests at the l ROSA facility of the Japanese Atomic Energy Research Insutute (JA2RI) was examined m i detail. Because of the timeliness of this NSRRC review regartiing the forthcoming ;
Commission decision whether to prcceed with this program, this letter has been prepared to ,
I set forth the relevant conclusions of our review. A supplementary report of the full scope of the July 1 and 2 meettng will follow which will contam detailed obsenanons and suggesuons '
relevant to the RES programs examined.
THE RES PROPOSAL The RES proposal examined was to conduct USNRC sponsored confirmatory integral systems tests on AP-600 using a full. pressure, full. height factilty. The purpose of these tests is to develop a sufficient data base with which to enhance the assessment of an analytical tool
! that could then be used with confidence to assess full size plant responses to initiating accident sequences. The selection of a facility is constrained by the: (a) desire to obtain test results prior to the currently scheduled preparation (Summer,1994) of the Draft Safety Evalus: ion Report (DSER) and the issuance (November,1994) of the Final Design Assessment (FDA),
and (b) need to obtain these results within the currently anticipated budget for this work of approximately $10 million. The selected facility is ROSA modified as pro psed by RES and agreed to by JAERI to a configurat;on (ROSA V) representing a 1/30 by vo: ume scaled model of AP 600 with the mm.ior rnodel deviation being the use of a single versus the actual two cold legs per loop.
l The NSRRC Subcommittee examined this proposal by posing and resolving a series of -
questions, starting with the need for this testing and culminating in the examination of the efficacy and adequacy of the proposed solution. These questions, restated specifically for AP 600 integral systems testing, will be sequentially reviewed next by summamias the NRC position and then stating the Subcommittee's conclusion.
"What are the NRC's needs for confirmatory systems research r eP.600"?
Integral systems tests in a full pressurs; full height facility areerewi beeense the response of the iystems to mitiating' events canno', be. analytically .with ggf 94 conddenos:by the ass of existing anatydcaLapons (computerW His is to bedrtbs-q of interacticos.betwoon systems.and cc= rend tbs low driving.beeds-
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( nec to quality an anatvit:al tool to assess plant resconse Tnis approsca is taken because r:o
- aied fac:hty can se:Ye as a cemonstration of full s:ze piant response to imtiating events.
I system behavior uncer three acc; cent sequences is of pamcular interest because the passive l Mety systems are called upon to operate at hign pressure.
, Small break loss-of-coolant c:ident. l Steam generator tube rupture, and
i Independent NRC testing at low pressure is not considered essential since the planned l vendor test program is deemed to y1cid sufficient data. However,it is anticipated that the tugh i
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pressure ROSA fac1hty to be used by the NRC as discussed later can be run to yield !
t supplementary lower pressure cata.
I 45RRC Subcommittee concurs with the NRC's need for indepencent confi. . story l sys ; search to insure that its analytic tools are qualified to assess full plant response.
! The availability of integral systems low pressure cata to insure performance of the gnvity I dram / core cooling system behavior is recognized as equally important as high pTessure test results.
"What integral systems testing program has the vendor proposed for A P-600" ?
l high pressure test program will be conducted in the full pressure, full height,1S95 by vt- e, scaled SPES 2 facility in Italy. A low pressure (400 psi maximuro) test program will be conducted in the 1/200 by volume scaled Oregon State University (OSU) facility. The extension of SPES tests below 400 psi so as to initialize OSU tests is being explored.
De NSRRC Subcommittee took note of this planned vendor test program.
l "Why should the NRC conduct confirmatory integral systems tests on AP 600 using a full pressure, full-height facility when the vendor will conduct a similar test program"?
! The NRC stated that they had a need to extend the expected vendor test matrix beyond the design ba. sis to develop confidence that the design basis is a satisfactory limit. His wuld be l achieved by experiments at or slightly beyond design basis conditions to ensure that no l unanticipated phenomena or major effects occurred in this operating band, and thereby l confirm the adequacy of the design basis limit.
The NSRRC Subcommittee concurs with the NRC need to develop confidence in the
! design basis in this manner. However, it is emphasized that we do believe that vendor l demonstration of satisfactory plant performance within the design basis should clearly remam
! the required standard for design approval.
l l "Why did the NRC select ROSA as the test facility rather than use the Italian
! SPES. faelifty in which the vendor will conduct tests or construct a newv dotnestic fae,llty"7 ,
The NRC couki have chosee so cocmetseparately with tlw SPES cv.ie= fecondoct.
of warnoan !@t w7g.-. 4 imo n==e r-gy prescribed-NRC test matrix; thereby-av
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- ;n caon m T-wc_ spa Esc 1e e :=ce =.02 Morrison,7/20/92 V Page 3 Vendor access to the facility will take precedence over NRC access. Delay in conducting the vendor program or extension of the vendor program is possible and would severely upset the NRC schedule for acquisition of NRC independendy-produced test data, ne value of the test results will be muimiwd if they can be used in the assessment of codes requued for NRC safety analyses.
The vendor has not presented analysis to the NRC to firmly establish that the data from SPES is valid by itself to qualify an analytic tool for use on a full scale plant.
Scale effects probably need to be assessed and confirmed, as they have in past NRC l
thermal / hydraulic test programs, by tests at different scales.
The NSRRC Subcommittee concurs that plans to conduct NRC tests in SPES would not be prudent because of the cited schedule and test scale concems.
"What is the NRC doing to ensure that the ROSA facility will be configured correctly and will simulate the performance of the AP 600 passive safety features with acceptable fidelity"?
The NRC has performed an extensive cmuj-Etive assessment, using the RELAP 5.
MOD 2.5 analytic tool, of the behavior of the ROSA facility and the AP-600 plant to the same set of initiating events. From thsee analyses, desired L.y....- sis in the ability of ROSA to simulate the phenamena appearingin the plant weeidentified. Costs for these Levi w specifically changes in the facdhy config-iaa. were estimated and suh===tly negotiated with the ROSA owner. The final negotiated configuradon has been analyzed and is --;- Ed to satisfactorily spresent all full plant phenomena including many, but not all, aspects of asyns esl loop behavior. The cost for ROSA modifications and the schedule for their
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implementation and the conduct of the test program meet NRC criteria. Further, the NRC
' stated to us that no domestic facility could come close to n==nns the NRC cost and schedule criteria in that a cost of $40 50 million and a tiene of appr =ie='dy three (3) years would be required to construct a domestic facility meeting or i..y.s.hg on the ROSA V facility criteria.
ne NSRRC Subcommines reviewed the technical basis for the proposed modifications to ROSA and its consequent suitability as the NRCs selected high presses test facility. The Subcommittee concludes that the following key facers need to be balanced in reaching a decision:
. The i. w of obtaining '=4, - %t NRC data to confirm the adequacy of the ,
design limit. l i
. The of obtaining these data in a timely meaner to allow their use in i assassing usedin safety analysgs. .
. Ihs need to avoid tha possibility of'=+=aMa r 204iy into the assessmaat < zooess from experimental data taken on a test facil.ty which may not represent ful plant phenmennin allaspects.
After reviewing the data pressoned1 and weighing these facers, the Subcomsmitess concurs with the RES swamraandanan to M with the ROSA V program for integral systems testing of the ANiOO plant design. Tids e.24ri needs no be part of a wellinesgrated 1
1 Dming the of this repost, the SatKh reesived and soviewed the commeno e Acas consehenes sensaming the SPES sad 105A hassel test $scilkka-
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- rogram involving careful coce enhancement ans assessment. anc possloly well-selec:cd separate-effects tests for phenomena that cannot be tully explorce in t .ese Intepal facdi:ll J Such a program is r.eeced since the purpose of the integral testing is not a demonst acon:
AP-600 performance, but rather it is to gatner data for code assessment. Rese aspects v ciscussed more fully in our supplemental report 1
Sincerely,f l'
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dlsi Neil E. Todreas :
l Chairman NSRRC Advancec Reacters Subcommir.ee .
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Appendix III
Subject:
Reply to Memorandum by V. K. Dhir on the June 3-4 Meeting in Idaho Falls The following responds point-by-point to comments provided by Dr. Dhir on the subject meeting. It is organized according to Dr. Dhir's memo.
Meeting Summary No comment. l l
Observations j
- 1. The ROSA tests are planned to cover the full pressure range from full system pressure to IRWST injection. They will, thus, overlap both the SPES high pressure and the OSU low pressure testing. We have stated we are interested in both high pressure and low pressure confirmatory testing. We explained to the ACRS that at the time when the staff was proposing to do separate low pressure l confirmatory testing, the facility we envisaged was identical to l OSU. At the time we were precluded from interacting with OSU due to conflict of interest considerations. Once these were resolved, we could no longer justify pursuing a duplicate facility.
I
( In contrast, based on our interest in high pressure systems interactions phenomena and processes, by June,1991 we identified ROSA as the best candidate for performing high pressure confirmatory testing. We have been working since that time towards formulating ,
a technically sound program to modify the facility and conduct !
testing.
l i In terms of experimental programs in scaled facilities intended to l model full scale power reactors, it is a well-established principal l that facilities of difftirent scales and scaling approaches should be used to ensure that the effects of scale are well-understood. By definition, scaling introduces distortions in all scaled facilities.
Testing programs must be - formulated accordingly. The integral system test program carried out to study small break LOCAs in Babcock and Wilcox reactors was an example of such a program.
Research Information Letter 164, describing the results of this program, is attached. We would like to refer Dr. Dhir to Commissioner Rogers' memorandum (attached) on thir, program. In addition, to quote from MIT Professor Peter Griffith's review of the ROSA program:
"This is a well structured program which clearly benefitted from our experience with the LOCA work done on LWRs during the 70's and 80's. Because the experimental program consists of l three integral tests being run on three quite different rigs l i l
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2 APPENDIX III each scaled according to three different rationales, I can't imagine that there will be many questions outstanding about the system performance when these experiments are completed.
We have, in addition, got an operating, documented computer
! code, RELAP-5, which can be used for the prediction and analysis of the ROSA-IV experiments. The pieces of the program will come together in a timely way so that the results of this program can be used to design out-any problems which might arise in the course of this research."
- 2. The costs and schedule would be effectively prohibitive. We have already given serious consideration to a domestic facility. We have stated that, ideally, this is oui preference, however, such a l facility could not be built within the FDA schedule, and would be considerably more expensive than the ROSA program.
In its letter of March 10, 1992, the ACRS stated " Inasmuch as FHFP integral system testing will require at least three to four years to complete, there is a risk that the present certification schedule will be affected unless the test program is begun now. We believe the likelihood of such an impact is great. If the present l certification schedule is to be adhered to, we recommend that a FHFP j testing program be initiated now." RES agrees with the ACRS' l
assessment.
- 3. The use of RELAP to perform comparative calculations of ROSA and AP600 is not circular and certainly is logical. We completed a i RELAP code applicability review in early 1991. RELAP models and correlations were reviewed from the perspective of new AP600
, phenomena and features that could be important to reactor safety.
l The purpose was to identify those areas in which new mathematical f
models of physical phenomena would be required to be added to RELAP5. In most cases, the AP600 design and its systems and the planned and off-normal operations were found to be similar enough to current PWRs that RELAP safety analysis applicability was unchanged.
There were basically no new phenomena involved in the AP600, however the physical parameter ranges and applications of the phenomena may be different than those in present generation reactors. Therefore, a validation program was laid out accordingly.
Review of ROSA with respect to our separate scaling study for a FHFP facility showed that ROSA meets all requirements. If Dr. Dhir is aware of a better method to demonstrate facility similitude and scaling adequacy, we should like to know of it.
{
l 3 APPENDIX III
- 4. Distortions will be present in ROSA, as they will be in each and every scaled facility. The true question is whether the major phenomena and processes are preserved. Our analyses have shown that they are.
In addition, JAERI has agreed that the facility will remain available beyond the initial set of 10 experiments.
- 5. The experiments must meet the objectives and specifications.
l determined prior to the experiments. Otherwise, they will be repeated.
l 6. RES has assessed the cost-benefit ratio for the ROSA program, along l with schedule requirements of the FDA, and found that this facility meets all requirements identified by NRC staff and contractors for confirmatory integral system testing.
- 7. We never planned to stop the experiments at the actuation of 3rd L stage of ADS. Rather, we plan to run the ROSA experiments to full i
depressurization, including the initiation of IRWST injection.
j 8. Aside from the problem of schedule, we could not achieve the same
! data base for the same cost with a new U.S. facility.
1
- 9. In an ideal world, we wodid also prefer a dedicated U.S. facility, j however, this is not possible within the constraints of schedule and budget. Pragmatic alternatives must be sought that will successfully meet the same needs within the given constraints.
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- UMTED STATES t i NUCLE AR REGULATORY COMMISSION
. * - -f ;l g .% A SmNG T O N O C '0555
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"E*0RANDUM FOR: Thomas E. Murley. Director !
! Office of Nuclear Reactor Regulation i
- CM: Eric S. Beckjord. Director Office of Nuclear Regulatory Research
SUBJECT:
RESEARCH INFORMATION LETTER 164, " THERMAL HYDRAULIC DATA BASE RELEVANT TO PLANTS OF THTBABC0CK AND WILCOX LOWERED-LOOP DESIGN" eferences: 1. ALAB-708, 16 NRC1770 December 29. 1982.
2.
Clarification of TM1 Action Plan Requirements. NUREG- !
0737, November 1980.
- 3. Letter from H. R. Denton to R. B. Minogue. " Request for the Conceptual Design of a facility for the Study of B&W and CE Integral System Characteristics," December 30, 1981.
4 J. Gloudemans and D. P. Birmingham, " MIST Program:
Summary of Key Results," NUREG/CP-0097, Vol. 4, March 1989.
! 5. K. Almenas, et al., " Scaling of Integral Facilities l
at Reduced Pressures " MDNE/061589, June 15, 1989.
- 6. K. Almenas, et al., " Evaluation of four MIST
' Atypicalities," MDNE/041089, April 10, 1989.
- 7. Letter from H. R. Denton to R. B. Minogue, " Request for Follow-on Program in the B&W Integral System Test l Facility (MIST)," October 31, 1984 This memorandum transmits results from research conducted in the Integral System Test (IST) program. IST includes the Multi-loop Integral System Test (MIST) facility at the Alliance Research Center in Alliance Ohio, and the University of haryland at College Park (UMCP) 2x4 Loop facility. This l research provided thermal-hydraulic experimental data relevant to plants of i
the Babcock and Wilcox (B&W) lowered-loop design. MIST was jointly sponsored by the U.S. Nuclear Regulatory Commission, the Electric Power Research Institute (EPRI), B&W Owners Group (B&WOG) and B&W. The UMCP 2x4 Loop, a reduced-pressure and small-scale facility, was designed to address scaling atypicalities of the MIST facility and to provide data for code assessment.
I Contacts: ,
R. Y. Lee, RES/DSR, 49-23560 H. Scott, RES/DSR, 49-23563
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- =culatorv !ssue:
r ailowing the Three Mile islano Unit 2 (TMI-2) accident, a numoer of reculatory issues concerning the design of the B&W reactors -ere raised. The effectiveness of feed ana bleed ano the boiler condenser mooe (SCM) of natural
- rculation was cnallengea during the THI-l Restart Hearing conoucted by the
- mic Safety ana Li:ensing Appeal Board [1]. In BCM, heat is removed from u.e or1 mary system througn vapor concensation in the steam generator and the accompanying primary-to-secondary heat transfer. Clarification of THI Action Plan Reouirements (NUREG-0737) Item II.K.3.30 (2) required that small-break i loss-of-coolant accident (LOCA) calculational models be compared to applicable data.
In response to NRR's request for integral system characteristics for B&W reactors (3), the NRC and industry formed a Test Advisory Group to make recommendations regarding the type of data base required to validate small-treak LOCA models. The IST program was formed in 1983 to acoutre the desired
- ata. The primary experimental facility in the IST program is the MIST facility.
lthough MIST is designed as a full-height and full-pressure integral experiment facility, it is still a scaled model of a B&W plant. Thus, it
~
entails various design compromises such as an atypical downtomer. These design compromises are a potential source of distortion of some of the physical phenomena (e.g., variation of flow regimes in the hot legs) which in turn could lead to atypical transient behavior (e.g., premature interruption of natural circulation). The UMCP 2x4 Loop, a reduced-height and reduced-pressure, integral experiment facility employed an alternate design approach (e.g., a more typical downcomer) to assess the impact, of some of the MIST design compromises on transient behavior.
Conclusion:
IST produced an integral experiment data base for natural circulation, small-break LOCA, feed and bleed, steam generator tube rupture, effects of non-condensible gases, and pump operations on small-break LOCA behavior [4, 5].
Key observations (1 to 6 for MIST, I and 7 for UMCP) are summarized below.
, (1) Natural circulation was studied under varying degrees of loss of primary l inventory. A key question about natural circulation was whether the BCM l would remove decay heat effectively and depressurize the reactor coolant j system. This mode of heat transfer was consistently observed.
(2) During small-break LOCA, heat removal from the primary system was
! further augmented by the staan yenting from the upper plenum to the i downconer through the reactor vessel vent valves (RVVV). Adequate heat j
removal was. observed in MIST tests for a wide range of pripry boundary conditions (i.e., variation of break sizes from 5 to 50 cm , variation of break locations, and both full- and half-capacity HPI flows). In all cases tested, the MIST system depressurized and attained primary circuit mass equilibrium without uncovering the core.
(3) MIST results show that the feed and bleed technique can be utilized to cool the core and 4tpressurize the primary system.
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- 4) for multiple simulated steam generator tube ruptures. the primary system ;
cepressurized rapidly due to tube rupture discharge. SCM like activity was observed between primary system break flow to the seconaary sloe of the ruptured steam generator. For a smaller numoer of tube ruptures.
- he primary system depressur12ed by single-loop cooldown of the intact ;
000. )
- 5) The presence of non-condensible gases reduced BCM cooling, but did not )
prevent primary system cooldown and depressurization.
! (6) Reactor coolant pump operation is a m tageous during small-break LOCA. ]
With forced flow, primary-to-secondar. neat transfer is maintained '
longer and more energy is removed from the break. Therefore, the primary system pressure decreases more rapidly and the primary system refills more quickly.
) The comparisons of the experimental result:; from the two facilities l indicated that the UMCP 2x4 Loop is able to simulate the thermal- l l
hyoraulic behavior observed in MIST. First. it reproduced the l cualitative aspects of the flow modes. That is. it exnibited similar local flow regimes, flow regime transitions, the presence of both steady state and boiler-condenser natural circulation flows, and loop
. asymmetries [5). Second, it reproduced the sequence at which these flow modes occur during an inventory depletion transient. Inventory scaling was used to estimate the quantitative aspects of the flow modes (e.g.,
duration, magnitude of pressure changes). A precise parameter to parameter mapping between the UMCP and MIST data is not implied and is, in fact, precluded by the stochastic nature of some flow mode
- transitions. However, key phenotaena of inventory transients can be simulated and the effect of pressure on the characteristics of these phenomena is understood. Despite the differences between the design of l the two facilities, similar thermal-hydraulic characteristics were l observed. The MIST design atypicalities do not affect the expected thermal-hydraulic behavior during a small-break LOCA [6).
Reaulatory Imolications:
i The MIST and the'UMCP 2x4 Loop experimental data provide a sufficient small-l break LOCA data base to satisfy the requirements of NUREG-0737. The integral system data is self-consistent, comprehensive, and suitable for benchmarking computer codes used to calculate B&W plant transients. Such benchmark calculations for MIST were performed and are in good agreement with experimental data. The data, as well as code calculations, show that various methods, such as feed and bleed and BCM, are effective modes of decay heat removal in a B&W plant during a small break LOCA.
Castricttons on OD olications: 0
'he scaling evaluation has shown that key thermal-hydraulic behavior (e.g.,
3CM) coserved in these facilities can be expected to occur in full scale B&W
- lants of the lowered-loop design. However, these test facilities are scaled coels of a B&W lowered-loop nuclear steam supply system. As such, various
', ;caling atypicalities in simulating a plant were required. Hence, these cata snould not be applied directly to a full scale plant. Rather, validated computer coces (TRAC-PWR, RELAP5) should be used to calculate plant small-break LOCA. The requisite code validation was performed as part of the IST program.
Further work:
At the request of NRR [7], additional testing was performed in the MIST facility to obtain data for: small-break LOCA without high pressure injection: station blackout; and examining scaling questions. Analysis of these tests is expected to be completed by the end of 1989. Additional experiments are being performed at UMCP to further test the scaling concepts ancer more complicated boundary conditions, i.e., with HP! flow. Any new
- ignificant results will be reported in a future RIL.
Eric S. Beckjord($0irector Office of Nuclear Regulatory Research
Enclosures:
(1) MIST Program: Summary of
- Key Results (2) Evaluation of Four MIST Atypicalities e
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