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Category:CORRESPONDENCE-LETTERS
MONTHYEARML20217M8851999-10-0808 October 1999 Informs of Staff Determination That Listed Calculations Should Be Withheld from Public Disclosure,Per 10CFR2.790, as Requested in 990909 Affidavit ML20211J7731999-08-31031 August 1999 Forwards Insp Rept 50-312/99-03 on 990802-06.No Violations Noted.Insp Included Decommissioning & Dismantlement Activities,Verification of Compliance with Selected TS & Review of Completed SEs ML20211H7481999-08-13013 August 1999 Forwards Amend 126 to License DPR-54 & Safety Evaluation. Amend Changes Permanently Defueled Technical Specification (PDTS) D3/4.1, Spent Fuel Pool Level, to Replace Specific Reference to SFP Level Alarm Switches with Generic Ref 3F0799-22, Provides Update & Rev to Submittal Made by Util Ltr with Regard to EAL Classification Methodology for Unit 3.Reponses to NRC Staff Questions Provided as Attachment D to Ltr & Reflects Discussions Held1999-07-13013 July 1999 Provides Update & Rev to Submittal Made by Util Ltr with Regard to EAL Classification Methodology for Unit 3.Reponses to NRC Staff Questions Provided as Attachment D to Ltr & Reflects Discussions Held ML20210H9541999-07-0707 July 1999 Informs NRC of Change to Rancho Seco Decommissioning Schedule,As Described in Licensee Post Shutdown Decommissioning Activities Rept ML20209D2501999-06-24024 June 1999 Informs That Util Has Revised All Sections of Rancho Seco Emergency Plan (Rsep),Change 4,effective 990624 ML20196G0431999-06-22022 June 1999 Forwards Certified Copies of Listed Nuclear Liability Endorsements for Smud Rancho Seco Nuclear Generating Station ML20195D1851999-05-27027 May 1999 Forwards Rancho Seco Annual Rept, IAW Plant Permanently Defueled TS D6.9.4 & D6.9.6b.Rept Contains Shutdown Statistics,Narrative Summary of Shutdown Experience,Er Info & Tabulation of Facility Changes,Tests & Experiments ML20195B8511999-05-27027 May 1999 Forwards Change 4 to Rancho Seco Emergency Plan, Incorporating Commitments Made to NRC as Outlined in NRC .Emergency Plan Includes Two Listed Supporting Documents ML20207E9181999-05-27027 May 1999 Informs That Effective 990328,NRR Underwent Reorganization. within Framework of Reorganization,Div of Licensing Project Mgt Created.Reorganization Chart Encl ML20206U7411999-05-18018 May 1999 Provides Summary of 990217-18 Visit to Rancho Seco Facility to Become Familar with Facility,Including Onsite ISFSI & Meeting with Representatives of Smud to Discuss Issues Re Revised Rancho Seco Ep,Submitted to NRC on 960429 ML20206M1611999-05-10010 May 1999 Forwards Listed Proprietary Calculations to Support Review of Rancho Seco ISFSI Sar.Proprietary Encls Withheld ML20206E8591999-04-12012 April 1999 Provides Info Re High Total Coliform Result in Plant Domestic Sewage Effluent Prior to Confluence with Combined Effluent.Cause of High Total Coliform Result Was Broken Flow Rate Instrument.Instrument Was Repaired on 990318 ML20204H6751999-03-19019 March 1999 Forwards Insp Rept 50-312/99-02 on 990309-11.No Violations Noted.Portions of Physical Security & Access Authorization Programs Were Inspected ML20204E4031999-03-16016 March 1999 Submits Rept of Status of Decommissioning Funding for Rancho Seco,As Required by 10CFR50.75(f)(1).Plant Is Currently in Safstor, with Operating License Scheduled to Expire in Oct 2008 ML20204E6661999-03-11011 March 1999 Forwards Rancho Seco Exposure Rept for Individuals That Received Greater than 100 Mrem During 1998,IAW TS D6.9.2.2 & NRC Regulatory Guide 1.16 ML20204E6441999-03-11011 March 1999 Forwards Individual Monitoring Repts for Personnel That Required Radiation Exposure Monitoring During 1998 ML20207L1711999-03-10010 March 1999 Informs of Staff Determination That Supporting Calculations & Drawings Contained in Rev 2 of Sar, Should Be Withheld from Public Disclosure,Per 10CFR2.790 NL-99-002, Submits Info About Property Insurance for Rancho Seco Nuclear Generating Station,Iaw 10CFR50.54(w)(3)1999-03-10010 March 1999 Submits Info About Property Insurance for Rancho Seco Nuclear Generating Station,Iaw 10CFR50.54(w)(3) ML20207D4431999-03-0101 March 1999 Forwards Annual Radioactive Effluent Release Rept, for Rancho Seco Nuclear Generating Station for 1998 ML20207H6181999-02-18018 February 1999 Provides Attached Metrix & Two Copies of Rancho Seco ISFSI Sar,Rev 2 on Compact Disc,As Requested in 990209 Meeting. First Rounds of RAIs Dealt Primarily with Use of Cask as Storage Cask.Without Compact Disc ML20203D0761999-02-10010 February 1999 Ltr Contract:Task Order 37 Entitled, Technical Assistance in Review of New Safety Analysis Rept for Rancho Seco Spent Fuel Storage Facility, Under Contract NRC-02-95-003 ML20155D4431998-10-27027 October 1998 Forwards Amend 3 to Rancho Seco Dsar,Representing Updated Licensing Basis for Operation of Permanently Shutdown & Defueled Rancho Seco Nuclear Facility During Permanently Defueled Mode.With Instructions & List of Effective Pages NL-98-032, Forwards Certified Copies of Listed Nuclear Liability Endorsements for Util1998-09-30030 September 1998 Forwards Certified Copies of Listed Nuclear Liability Endorsements for Util ML20237A6031998-08-0707 August 1998 Forwards Insp Rept 50-312/98-03 on 980706-09.No Violations Noted ML20237A9481998-08-0303 August 1998 Forwards Smud 1997 Annual Rept, IAW 10CFR50.71(b),which Includes Certified Financial Statements ML20236Q9461998-07-15015 July 1998 Ack Receipt of Informing NRC of Steps Taken to Correct Violations Noted in Insp Rept 50-312/98-02 ML20236J6331998-06-30030 June 1998 Forwards Response to Violations Noted in Insp Rept 50-312/98-02.Corrective Actions:Util Revised RSAP-1003 to Clarify District Security Staff Responsibilities Re Handling & Review of Criminal History Info ML20236E8211998-06-0303 June 1998 Forwards Insp Rept 50-312/98-02 on 980519-21 & NOV Re Failure to Review & Consider All Info Obtained During Background Investigation.Areas Examined During Insp Also Included Portions of Physical Security Program ML20217G8391998-04-20020 April 1998 Forwards Copy of Rancho Seco Monthly Discharger Self-Monitoring Rept for Mar 1998 IR 05000312/19980011998-03-25025 March 1998 Ack Receipt of Informing NRC of Steps Taken to Correct Violations Noted in Insp Rept 50-312/98-01 on 980205 ML20217F1891998-03-18018 March 1998 Forwards Signed Original & Amend 7 to Rancho Seco Long Term Defueled Condition Physical Security Plan & Rev 4 to Long Term Defueled Condition Training & Qualification Plan.Encls Withheld,Per 10CFR2.790 ML20217G6661998-03-18018 March 1998 Forwards Discharge Self Monitoring Rept for Feb 1998, Which Makes Note of One Wastewater Discharge Permit Violation ML20217H0451998-03-18018 March 1998 Submits Rancho Seco Exposure Rept for Individuals Receiving Greater than 100 Mrem During 1997,per TS D6.9.2.2 & Guidance Contained in Reg Guide 1.16.No One Exposed to Greater than 100 Mrem in 1997 ML20216K1091998-03-11011 March 1998 Forwards NRC Form 5 Individual Monitoring Repts for Personnel Who Required Radiation Exposure Monitoring,Per 10CFR20.1502 During 1997.W/o Encl ML20217N9531998-03-0505 March 1998 Responds to Violations Noted in Insp Rept 50-312/98-01. Corrective Actions:Radiation Protection Group Wrote Potential Deviation from Quality (Pdq) 97-0082 & Assigned Radiation Protection Action to Determine Cause & CAs ML20203H7001998-02-25025 February 1998 Forwards Annual Radioactive Effluent Release Rept for Jan- Dec 1997, IAW 10CFR50.36a(a)(2) & TS D6.9.3.Revs to Radiological Environ Monitoring Manual & off-site Dose Calculation Manual,Encl ML20202G0131998-02-12012 February 1998 Forwards Certified Copies of Listed Nuclear Liability Endorsements & Master Worker Policy Certificate of Insurace for Facility NL-98-006, Submits Annual Rept of Property Insurance Levels for Rsngs, IAW 10CFR50.54(w)(3)1998-02-12012 February 1998 Submits Annual Rept of Property Insurance Levels for Rsngs, IAW 10CFR50.54(w)(3) ML20202C4641998-02-0505 February 1998 Forwards Insp Rept 50-312/98-01 on 980105-08 & Notice of Violation.Insp Included Decommissioning & Dismantlement Work Underway,Verification of Compliance W/Selected TS & Main & Surveillance Activities Associated W/Sfp ML20199A5881997-11-10010 November 1997 Responds to NRC Re Violations Noted in Insp Rept 50-312/97-01.Corrective Actions:Reviewed SFP Water Temp & Instrument Calibr Records,Generated Otr 97-001 to Document out-of-tolerance Instrument & Generated Pdq 97-0064 ML20198R9501997-11-0505 November 1997 Requests Interpretation of or Rev to NUREG-1536, Std Review Plan for Dry Cask Storage Sys, Re Compliance W/ 10CFR72.236(e) & 10CFR72.122(h)(4) for Dry Fuel Storage Casks ML20198K5391997-10-21021 October 1997 Forwards Insp Rept 50-312/97-04 on 970922-25 & Notice of Violation.Response Required & Will Be Used to Determine If Further Action Will Be Necessary ML20217D3101997-09-25025 September 1997 Forwards Update of 1995 Decommissioning Evaluation, for Rancho Seco Nuclear Generation Station & Annual Review of Nuclear Decommissioning Trust Fund for Adequacy Re Assumptions for Inflation & Rate of Return ML20211F0991997-09-23023 September 1997 Forwards One Certified Copy of Mutual Atomic Energy Liability Underwriters Nuclear Energy Liability Insurance Endorsement 120 for Policy MF-0075 for Smud Rancho Seco Nuclear Facility ML20198G8141997-08-22022 August 1997 Forwards Amend 125 to License DPR-54 & Safety Evaluation. Amend Permits Smud to Change TS to Incorporate Revised 10CFR20.Amend Also Revises References from NRC Region V to NRC Region IV ML20151L0281997-07-29029 July 1997 Provides Response to NRC Request for Addl Info Re TS Change,Relocating Administrative Controls Related to QA to Ufsar,Per NUREG-0737 ML20149E5031997-07-10010 July 1997 Second Partial Response to FOIA Request for Documents. Forwards Records Listed in App C Being Made Available in Pdr.Records in App D Already Available in PDR ML20148P5161997-06-30030 June 1997 Second Partial Response to FOIA Request for Documents.App B Records Being Made Available in PDR ML20141A1721997-06-17017 June 1997 Forwards Insp Rept 50-312/97-03 on 970603-05.No Violations Noted.Areas Examined During Insp Included Portions of Physical Security Program 1999-08-31
[Table view] Category:INCOMING CORRESPONDENCE
MONTHYEAR3F0799-22, Provides Update & Rev to Submittal Made by Util Ltr with Regard to EAL Classification Methodology for Unit 3.Reponses to NRC Staff Questions Provided as Attachment D to Ltr & Reflects Discussions Held1999-07-13013 July 1999 Provides Update & Rev to Submittal Made by Util Ltr with Regard to EAL Classification Methodology for Unit 3.Reponses to NRC Staff Questions Provided as Attachment D to Ltr & Reflects Discussions Held ML20210H9541999-07-0707 July 1999 Informs NRC of Change to Rancho Seco Decommissioning Schedule,As Described in Licensee Post Shutdown Decommissioning Activities Rept ML20209D2501999-06-24024 June 1999 Informs That Util Has Revised All Sections of Rancho Seco Emergency Plan (Rsep),Change 4,effective 990624 ML20196G0431999-06-22022 June 1999 Forwards Certified Copies of Listed Nuclear Liability Endorsements for Smud Rancho Seco Nuclear Generating Station ML20195B8511999-05-27027 May 1999 Forwards Change 4 to Rancho Seco Emergency Plan, Incorporating Commitments Made to NRC as Outlined in NRC .Emergency Plan Includes Two Listed Supporting Documents ML20195D1851999-05-27027 May 1999 Forwards Rancho Seco Annual Rept, IAW Plant Permanently Defueled TS D6.9.4 & D6.9.6b.Rept Contains Shutdown Statistics,Narrative Summary of Shutdown Experience,Er Info & Tabulation of Facility Changes,Tests & Experiments ML20206M1611999-05-10010 May 1999 Forwards Listed Proprietary Calculations to Support Review of Rancho Seco ISFSI Sar.Proprietary Encls Withheld ML20206E8591999-04-12012 April 1999 Provides Info Re High Total Coliform Result in Plant Domestic Sewage Effluent Prior to Confluence with Combined Effluent.Cause of High Total Coliform Result Was Broken Flow Rate Instrument.Instrument Was Repaired on 990318 ML20204E4031999-03-16016 March 1999 Submits Rept of Status of Decommissioning Funding for Rancho Seco,As Required by 10CFR50.75(f)(1).Plant Is Currently in Safstor, with Operating License Scheduled to Expire in Oct 2008 ML20204E6441999-03-11011 March 1999 Forwards Individual Monitoring Repts for Personnel That Required Radiation Exposure Monitoring During 1998 ML20204E6661999-03-11011 March 1999 Forwards Rancho Seco Exposure Rept for Individuals That Received Greater than 100 Mrem During 1998,IAW TS D6.9.2.2 & NRC Regulatory Guide 1.16 NL-99-002, Submits Info About Property Insurance for Rancho Seco Nuclear Generating Station,Iaw 10CFR50.54(w)(3)1999-03-10010 March 1999 Submits Info About Property Insurance for Rancho Seco Nuclear Generating Station,Iaw 10CFR50.54(w)(3) ML20207D4431999-03-0101 March 1999 Forwards Annual Radioactive Effluent Release Rept, for Rancho Seco Nuclear Generating Station for 1998 ML20207H6181999-02-18018 February 1999 Provides Attached Metrix & Two Copies of Rancho Seco ISFSI Sar,Rev 2 on Compact Disc,As Requested in 990209 Meeting. First Rounds of RAIs Dealt Primarily with Use of Cask as Storage Cask.Without Compact Disc ML20155D4431998-10-27027 October 1998 Forwards Amend 3 to Rancho Seco Dsar,Representing Updated Licensing Basis for Operation of Permanently Shutdown & Defueled Rancho Seco Nuclear Facility During Permanently Defueled Mode.With Instructions & List of Effective Pages NL-98-032, Forwards Certified Copies of Listed Nuclear Liability Endorsements for Util1998-09-30030 September 1998 Forwards Certified Copies of Listed Nuclear Liability Endorsements for Util ML20237A9481998-08-0303 August 1998 Forwards Smud 1997 Annual Rept, IAW 10CFR50.71(b),which Includes Certified Financial Statements ML20236J6331998-06-30030 June 1998 Forwards Response to Violations Noted in Insp Rept 50-312/98-02.Corrective Actions:Util Revised RSAP-1003 to Clarify District Security Staff Responsibilities Re Handling & Review of Criminal History Info ML20217G8391998-04-20020 April 1998 Forwards Copy of Rancho Seco Monthly Discharger Self-Monitoring Rept for Mar 1998 ML20217H0451998-03-18018 March 1998 Submits Rancho Seco Exposure Rept for Individuals Receiving Greater than 100 Mrem During 1997,per TS D6.9.2.2 & Guidance Contained in Reg Guide 1.16.No One Exposed to Greater than 100 Mrem in 1997 ML20217F1891998-03-18018 March 1998 Forwards Signed Original & Amend 7 to Rancho Seco Long Term Defueled Condition Physical Security Plan & Rev 4 to Long Term Defueled Condition Training & Qualification Plan.Encls Withheld,Per 10CFR2.790 ML20217G6661998-03-18018 March 1998 Forwards Discharge Self Monitoring Rept for Feb 1998, Which Makes Note of One Wastewater Discharge Permit Violation ML20216K1091998-03-11011 March 1998 Forwards NRC Form 5 Individual Monitoring Repts for Personnel Who Required Radiation Exposure Monitoring,Per 10CFR20.1502 During 1997.W/o Encl ML20217N9531998-03-0505 March 1998 Responds to Violations Noted in Insp Rept 50-312/98-01. Corrective Actions:Radiation Protection Group Wrote Potential Deviation from Quality (Pdq) 97-0082 & Assigned Radiation Protection Action to Determine Cause & CAs ML20203H7001998-02-25025 February 1998 Forwards Annual Radioactive Effluent Release Rept for Jan- Dec 1997, IAW 10CFR50.36a(a)(2) & TS D6.9.3.Revs to Radiological Environ Monitoring Manual & off-site Dose Calculation Manual,Encl NL-98-006, Submits Annual Rept of Property Insurance Levels for Rsngs, IAW 10CFR50.54(w)(3)1998-02-12012 February 1998 Submits Annual Rept of Property Insurance Levels for Rsngs, IAW 10CFR50.54(w)(3) ML20202G0131998-02-12012 February 1998 Forwards Certified Copies of Listed Nuclear Liability Endorsements & Master Worker Policy Certificate of Insurace for Facility ML20199A5881997-11-10010 November 1997 Responds to NRC Re Violations Noted in Insp Rept 50-312/97-01.Corrective Actions:Reviewed SFP Water Temp & Instrument Calibr Records,Generated Otr 97-001 to Document out-of-tolerance Instrument & Generated Pdq 97-0064 ML20198R9501997-11-0505 November 1997 Requests Interpretation of or Rev to NUREG-1536, Std Review Plan for Dry Cask Storage Sys, Re Compliance W/ 10CFR72.236(e) & 10CFR72.122(h)(4) for Dry Fuel Storage Casks ML20217D3101997-09-25025 September 1997 Forwards Update of 1995 Decommissioning Evaluation, for Rancho Seco Nuclear Generation Station & Annual Review of Nuclear Decommissioning Trust Fund for Adequacy Re Assumptions for Inflation & Rate of Return ML20211F0991997-09-23023 September 1997 Forwards One Certified Copy of Mutual Atomic Energy Liability Underwriters Nuclear Energy Liability Insurance Endorsement 120 for Policy MF-0075 for Smud Rancho Seco Nuclear Facility ML20151L0281997-07-29029 July 1997 Provides Response to NRC Request for Addl Info Re TS Change,Relocating Administrative Controls Related to QA to Ufsar,Per NUREG-0737 NL-97-030, Forwards Endorsement 132 to Nelia Policy NF-0212 & Endorsement 118 to Maelu Policy MF-0075 for Smuds Rsngs1997-05-13013 May 1997 Forwards Endorsement 132 to Nelia Policy NF-0212 & Endorsement 118 to Maelu Policy MF-0075 for Smuds Rsngs ML20138F5321997-04-28028 April 1997 Forwards Response to RAI Re License Amend 192,updating Cask Drop Design Basis Analysis,Per NRC 960510 Request for Addl Info on 960318 Application NL-97-027, Forwards Certified Copies of Listed Nuclear Liability Endorsements for Utility1997-04-17017 April 1997 Forwards Certified Copies of Listed Nuclear Liability Endorsements for Utility ML20137W8091997-03-20020 March 1997 Forwards Biennial Update to Rancho Seco Post-Shutdown Decommissioning Activities Rept ML20137S3571997-03-19019 March 1997 Provides Notification of Use of Revised Quality Manual for Activities Re Rancho Seco ISFSI ML20137D0981997-03-18018 March 1997 Submits Rancho Seco Exposure Rept for Individuals Receiving Greater than 100 Mrem During 1996.Provided IAW TS D6.9.2.2 & Guidance Contained in NRC Reg Guide 1.16.No One Exposed to Greater than 100 Mrem in 1996 ML20137D1221997-03-18018 March 1997 Submits,Iaw 10CFR20.2206 & TS D6.9.2.1,1996 NRC Form 5 Individual Monitoring Repts for Personnel Requiring Radiation Exposure Monitoring Per 10CFR20.1502 During 1996. W/O Encl NL-97-012, Submits Rept of Listed Current Levels of Property Insurance for Plant,Iaw 10CFR50.54(w)(3)1997-02-11011 February 1997 Submits Rept of Listed Current Levels of Property Insurance for Plant,Iaw 10CFR50.54(w)(3) ML20138L1091997-01-29029 January 1997 Informs of Schedule Change Re Decommissioning of Rancho Seco.Incremental Decommissioning Action Plan,Encl NL-97-005, Forwards Certified Copies of Listed Nuclear Liability Endorsements for Utility1997-01-22022 January 1997 Forwards Certified Copies of Listed Nuclear Liability Endorsements for Utility NL-96-056, Forwards Certified Copies of Listed Nuclear Liability Endorsements for Util1996-12-16016 December 1996 Forwards Certified Copies of Listed Nuclear Liability Endorsements for Util ML20134E0041996-10-23023 October 1996 Forwards Response to NRC GL 96-04, Boraflex Degradation in Spent Fuel Pool Storage Racks ML18102B6871996-08-0606 August 1996 Informs That Util Will Revise Loading & Unloading Procedures & Operator Training as Necessary ML20149E4491994-05-16016 May 1994 Forwards 1993 Annual Rept of Sacramento Municipal Utility District,For Info ML20149E3971994-05-10010 May 1994 Forwards Re Updated Decommissioning Cost Estimate for Rancho Seco & Attached Rept by Tlg Engineering,Inc. W/Svc List ML20059H6731994-01-20020 January 1994 Forwards Revised Rancho Seco Quality Manual, Reflecting Current Rancho Seco Pol Phase Nuclear Organization Changes ML20059E1221994-01-0303 January 1994 Forwards Amend 7 to Long Term Defueled Condition Physical Security Plan.Encl Withheld (Ref 10CFR73) ML20059C1681993-12-22022 December 1993 Forwards Suppl Info to Support Review & Approval of 930514 Proposed License Amend 186 Re Nuclear Organization Changes, Per NRC Request 1999-07-07
[Table view] Category:UTILITY TO NRC
MONTHYEARML20059L5431990-09-20020 September 1990 Requests Exemptions from Certain Requirements of 10CFR50.47(b) & 50,App E & Proposes New Emergency Plan That Specifically Applies to Long Term Defueled Condition ML20059J9161990-09-13013 September 1990 Notification of Change in Operator/Senior Operator Status for R Groehler,Effective 900907 ML20059J9221990-09-13013 September 1990 Responds to Generic Ltr 90-03, Relaxation of Staff Position in Generic Ltr 83-28,Item 2.2,Part 2, 'Vendor Interface for Safety-Related Components.' No Vendor Interface Exists for Spent Fuel Pool Liner NL-90-442, Forwards Endorsements 13 to Nelia Certificate N-49 & Maelu Certificate M-49,Endorsements 91 & 92 to Maelu Policy MF-75 & Endorsements 103 & 104 to Nelia Policy NF-2121990-09-12012 September 1990 Forwards Endorsements 13 to Nelia Certificate N-49 & Maelu Certificate M-49,Endorsements 91 & 92 to Maelu Policy MF-75 & Endorsements 103 & 104 to Nelia Policy NF-212 ML20059G0791990-09-0606 September 1990 Forwards Supplemental Fitness for Duty Performance Data, Omitted from 900725 Rept Re Random Drug Testing Results ML20059E0031990-08-30030 August 1990 Forwards Semiannual Radioactive Effluent Release Rept,Jan- June 1990, Corrected Repts & Revs to ODCM ML20059C2491990-08-27027 August 1990 Advises That M Foster & B Rausch Leaving Util Effective on 900810 & 17,respectively & Will No Longer Require Active Operator Licenses ML20056B2591990-08-20020 August 1990 Forwards Long-Term Defueled Condition Security Training & Qualification Plan. Encl Withheld (Ref 10CFR2.790) ML20056B2961990-08-10010 August 1990 Discusses 900731 Meeting Re Future of Util & Closure & Decommissioning of Facility.Request for Possession Only License Pending Before Commission ML20058Q2811990-08-0909 August 1990 Forwards Updated Listing of Commitments & long-range Scope List Items Deferred or Closed by Commitment Mgt Review Group Since Last Update ML20058N0911990-08-0707 August 1990 Notifies of Minor Change in List of Tech Specs Applicable in Plant Defueled Condition.Determined That Surveillance Requirements Table 4.1-1,Item 63 Not Required to Be Included in List of Tech Specs Applicable in Defueled Condition ML20056A1131990-07-30030 July 1990 Apprises of Status of Plans to Use 3 of 4 Emergency Diesel Generators as Peaking Power Supplies & Responds to Questions in .Util Obtained Authorization for Operation of Diesel Generators for No More than 90 Days Per Yr ML20056A2041990-07-30030 July 1990 Provides Response to NRC Bulletin 90-001, Loss of Fill Oil in Transmitters Mfg by Rosemount. Pressure & Differential Pressure Transmitters 1153 & 1154 Do Not Perform Any safety-related Function in Current Plant Mode ML20055J0311990-07-25025 July 1990 Forwards fitness-for-duty Performance Data for Facility from 900103-0630 ML20055J0331990-07-25025 July 1990 Notifies of Change in Operator/Senior Operator Status. Operators Terminating Employment & No Longer Require License ML20055H8081990-07-24024 July 1990 Forwards Decommissioning Financial Plan for Plant,Per 10CFR50.33(k)(2) & Requests Interim Exemption Re Requirement to Have Full Decommissioning Funding at Time of Termination of Operation,Per 10CFR50.12 ML20055H7561990-07-24024 July 1990 Requests Exemption from Performing Annual Exercise of Emergency Plan,Activation of Alert & Notification Sys & Distribution of Public Info Brochures,Per 10CFR50.12 Requirements ML20055F8421990-07-13013 July 1990 Forwards Application for Proposed Decommissioning of Plant. Util Needs Relief from Equipment Maint,Surveillance,Staffing & Other Requirements Not Necessary to Protect Public Health & Safety During Defueled Condition ML20055G9821990-07-12012 July 1990 Advises That Environ Exposure Controls Action Plan Will Be Provided by Sept 1990,per Insp Rept 50-312/90-02 ML20055E5111990-07-0606 July 1990 Notifies of Change in Operator/Senior Operator Status for D Rosenbaum & M Cooper,Effective 900622 & 29,respectively ML20055C3541990-02-14014 February 1990 Forwards Updated Response to Insp Rept 50-312/88-30. Calculations for Liquid Effluent Monitors Completed & in Use & Rev to Reg Guide 4.15 in Procedure RSAP-1702 Scheduled to Be Completed & Implemented by Apr 1990 ML20055C3511990-02-14014 February 1990 Forwards Addl Info Re 900306 Response to NRC Bulletin 88-003, Inadequate Latch Engagement in Hfa Type Latching Relays Mfg by Ge. Util Will Replace Only Relays Found Not to Meet Insp Criteria ML20248H2571989-10-0606 October 1989 Responds to NRC Re Addendum to Safety Evaluation Supporting Amend 92 to License DPR-54 Re Reactor Vessel Vent Valve Testing.No Testing of Reactor Vessel Vent Valves Will Be Performed ML20248H2391989-10-0606 October 1989 Requests Exemption from Requirements of 10CFR26 Re Fitness for Duty Programs Based on Present & Future Operational Configuration ML20248A8271989-09-25025 September 1989 Requests Permission to Submit Next Amend to Updated FSAR W/Decommissioning Plan Submittal.Extension Will Allow District to Incorporate Plant Closure Status in SAR Update to Reflect Plant Conditions Accurately ML20248D4611989-09-13013 September 1989 Responds to 890906 Request for Assessment of Util Compliance W/Ol & Associated Programs & Commitments,Per 10CFR50.54(f). Staffing Requirements for Emergency Preparedness Will Not Be Violated & Future Shortfalls Will Be Remedied ML20247G1991989-09-11011 September 1989 Requests Extension for Time Period Equivalent to That of Current Shutdown.Extension Would Result in Revised Final Expiration Date of Not Earlier than 900318.Plant Would Not Be Brought Above Cold Shutdown W/O NRC Prior Concurrence ML20247H3551989-09-0707 September 1989 Informs That Util Stands by Commitments of 890621 & 0829 Re Implementation of Closure Plan in Safe & Deliberate Manner in Compliance W/License & W/All Applicable Laws & Regulations ML20247H5541989-09-0101 September 1989 Responds to Violations Noted in Insp Rept 50-312/89-14. Corrective Actions:Stop Order on Fuel Movement Issued & Action Plan Generated on 890908 to Address Broader Issues 05000312/LER-1988-010, Forwards Rev 1 to LER 88-010,due to Change in Commitment Date for re-evaluating Fire Zones.Date Changed to 901001. Zones re-evaluated in Conjunction W/Mods to Fire Detection Annunciator Sys1989-08-23023 August 1989 Forwards Rev 1 to LER 88-010,due to Change in Commitment Date for re-evaluating Fire Zones.Date Changed to 901001. Zones re-evaluated in Conjunction W/Mods to Fire Detection Annunciator Sys ML20246A4011989-08-16016 August 1989 Forwards Rev 5 to Inservice Testing Program Plan. Changes Identified Consistent W/Guidance Provided by Generic Ltr 89-04 NL-89-593, Forwards Plant Closure Organizational Charts & Administrative Procedure RSAP-0101,per 890802 Request1989-08-15015 August 1989 Forwards Plant Closure Organizational Charts & Administrative Procedure RSAP-0101,per 890802 Request ML20245H4781989-08-10010 August 1989 Requests Exemption from Generic Ltr 89-07, Power Reactor Safeguards Contingency Planning for Surface Vehicle Bombs Because on 890607,util Board of Directors Ordered That Plant Cease Operation ML20245H1781989-08-0909 August 1989 Notifies of Change in Operator/Senior Operator Status. J Dailey & J Reynolds Terminated Employment on 890721 & 890802,respectively ML20245L1831989-08-0808 August 1989 Informs That Official Correspondence Must Be Directed to Listed Individuals Due to Reorganization of Util Following 890606 Election ML20247L9221989-07-26026 July 1989 Provides Revised Response to NRC Re Violations Noted in Insp Rept 50-312/88-33.Corrective Action:Portable Shield Walls Inspected Every 6 Months to Ensure All Safety Factors Met & Area Surveys Conducted on Weekly Basis ML20247M4121989-07-24024 July 1989 Requests Exemption from 10CFR50,App E,Section IV.F.2 to Allow Util Not to Perform Annual Emergency Plan Exercise for 1989.Request Results from Transitional Mode of Plant from Operating Plant to Plant Preparing for Decommissioning NL-89-541, Requests That Completion Date for Addl Training of Personnel Involved in Performing Work on Environ Qualified Equipment Be Extended from 890616 to 8912151989-07-14014 July 1989 Requests That Completion Date for Addl Training of Personnel Involved in Performing Work on Environ Qualified Equipment Be Extended from 890616 to 891215 ML20246P4011989-07-14014 July 1989 Informs That Evaluation of Contracts & Agreements Identified No Restrictions on Employee Ability to Provide Info About Potential Safety Issues to NRC NL-89-547, Forwards Amend 110 to License DPR-54,issued on 890609, Identifying Discrepancy in Tech Spec Page X (Table of Contents) Which Does Not Reflect Changes Approved in Amend 1061989-07-0606 July 1989 Forwards Amend 110 to License DPR-54,issued on 890609, Identifying Discrepancy in Tech Spec Page X (Table of Contents) Which Does Not Reflect Changes Approved in Amend 106 ML20246A9751989-06-30030 June 1989 Advises That Concerns Addressed in Generic Ltr 89-08 Inapplicable,Since Util Intends to Defuel Reactor.Generic Ltr Will Be Reviewed Prior to Placing Facility in heatup-cooldown Operational Mode for Return to Power ML20246A5171989-06-30030 June 1989 Forwards Rancho Seco Closure Plan, Per 890621 Request for Addl Info Re Plan CEO-89-289, Notifies of Change in Operator/Senior Operator Status.Listed Operator/Senior Operator Terminated Employment on Listed Effective Date1989-06-27027 June 1989 Notifies of Change in Operator/Senior Operator Status.Listed Operator/Senior Operator Terminated Employment on Listed Effective Date NL-89-526, Lists Discrepancies Noted in Amend 109 to License DPR-54,per 890615 Discussion W/S Reynolds.Tech Specs Encl1989-06-22022 June 1989 Lists Discrepancies Noted in Amend 109 to License DPR-54,per 890615 Discussion W/S Reynolds.Tech Specs Encl ML20245H4181989-06-21021 June 1989 Discusses Util Plans Re Overall Closure of Plant,Per 890615 Meeting W/Nrc.Util Will Request Appropriate Changes to Tech Specs to Reflect Defueled Mode & Will Evaluate & Request Changes to Emergency Plan ML20245D9281989-06-21021 June 1989 Discusses Activities Underway Re Plan for Closure of Plant Discussed During 890615 Meeting W/Region V.Util Intends to Continue Use of Essential Programs,Such as Preventive Maint Program,For Sys within Scope of Closure Process ML20245A0981989-06-16016 June 1989 Responds to NRC Bulletin 89-001, Failure of Westinghouse Steam Generator Tube Mechanical Plugs. No Westinghouse Plugs Used at Plant ML20248B5751989-06-0202 June 1989 Advises That Util Anticipates That Final Analysis of Thermal Striping Will Conservatively Support Surge Line Lifetime Significantly Longer than June 1994 Date,Per NRC Bulletin 88-011, Pressurizer Surge Line Thermal Stratification NL-89-468, Submits Justification for Absence of Functional Testing Requirement in Proposed Tech Spec 4.14(f) Re Snubber Svc Life Monitoring,Per 890517 Request1989-05-30030 May 1989 Submits Justification for Absence of Functional Testing Requirement in Proposed Tech Spec 4.14(f) Re Snubber Svc Life Monitoring,Per 890517 Request ML20247N2601989-05-25025 May 1989 Requests Guidance Re Whether NRC Concurs W/Arbitrator Order Concerning Employee Access to Plant 1990-09-06
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$'SMUD SACRAMENTO MUNICtPAL UTIUTY DISTRICT O P. O. Box 15830, Sacramento CA 95852-1830,(916) 452-3211 AN ELECTRIC SYSTEM SERVING THE HEART OF CALIFORNIA JEW 87-172 March 3, 1987 Director of Nuclear Reactor Regulation Attention: Frank J. Miraglia, Jr.
Division of PWR Licensing-B U S Nuclear Regulatory Commission Washington D C 20555 Docket 50-312 Rancho Seco Nuclear Generating Station Unit #1 NUREG 0737, ITEM II.D.1 REQUEST FOR INFORMATION
Dear Mr. Miraglia:
Your letter dated December 16, 1986, requested additional information to complete your review of NUREG 0737, Item II.D.1. Attached is the District's response to this request. As noted in the response to question number 7 the results of the as-built reanalysis is scheduled for April 1987.
If you have any questions pertaining to this submittal contact John Atwell of my staff at extension 3906.
Sincerely,
/
Deputy General Manager, Nuclear 1
Attachments cc: Syd Miner, NRC - Bethesda A. D'Angelo, NRC - Rancho Seco A oMo g3oRDoEkfsNh e P
DISTRICT HEADQUARTERS O 6201 S Street, Sacramento CA 95817-1899
ATTACIDENT 1 NUREG 0737, ITEM II.D.1 QUESTION 1 Provide the torque setting for the PORY block valve operator at Rancho Seco 1 and the torque produced at this setting. If the torque is less than 82 ft-lbs (the mini === torque tested by EPRI), it is the staff's position that it is not adequate to conclude proper operation solely on manufacturer's calculations. The problems encountered with Westinghouse gate valves on closing, which were traced to the calculations used to size the valve operator torque requirements, indicate the need to experimentally verify the adequacy of the block valve / operator combination. SMUD should provide test data to demonstrate the SMB-00-10 operator at Rancho Seco 1 is capable of providing adequate torque to close the block valve.
RESPONSE: The torque setting for the PORV valve has been determined to be 87.6 ft-lbs. This value has been determined as a result of the District's MOVATS testing program (IEB 85-03). The important parameter for gate valve operability is the thrust applied to the stem. MOVATS testing of the PORV block valve has confirmed that the required thrust value (7186 lbs.) will be generated by the valve notor at a torque setting of 87.6 ft-lbs. As a result of this testing, it is the District's position that the PORV block valve motor will provide the required torque to close the valve under the expected operating conditions.
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NUREG 0737, ITEM II.D.1 (CONT.)
QUESTION 2: Discuss how REIAP-FORCE calculates piping forces from REIAPS output and describe how the code was verified. Provide comparisons of RRLAP-FORCE calculations and EPRI/CE data.
RESPONSE: The thermal-hydraulic analyses were performed with the RELAPS-FORCE code, version 14, developed by Gilbert Associates.
This code was developed by revising REIAPS/ MOD 1 to include the hydraulic force equation. Verification of this modification was performed by running a sample problem without the force option and comparing it to RELAPS/ MOD 1 results to ensure that the modifications did not alter the basic RELAP5/ MOD 1 calculations.
The adequacy of the hydrodynamic force option was subsequently verified by comparing analytical results to test results from several tests including the EPRI/CE SRV test program. The report indicated reasonable agreement with test data and that RELAP5-FORCE is an acceptable program for thermal-hydraulic analyses of the S/RV discharge piping. A copy of the verification report is enclosed as Attachment 2 to this letter.
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NUREG 0737, ITEM II.D.1 (CONT.)
QUESTION 3: The submittal stated intensification factors were used to compensate for a lack of sufficient detail in the REIAP5 thermal-hydraulic analysis. How were the intensification factors developed? This was not clear from the submittal. How was the approach, i.e., applying intensification factors to results from a model with less than sufficient detail, verified? Supply the results of verification analyses of appropriate EPRI/CE tests using this approach for our review to assure this method conservatively bounds the discharge piping loads.
RESPONSE: In the nodalization study performed in Appendix C of reference 1, the minimum control volume size was 0.5 ft. in length. Using this length control volume for the Rancho Seco model would have greatly exceeded the RELAP5 program capacity. In order to overcome this problem, intensification factors were developed to use as multipliers for the forcing functions in the piping structural analyses. The forcing function for each straight pipe segment was multiplied by an intensification factor yielding values which are representative of the values expected from a finer model.
The intensification factor was determined by performing a nodalization sensitivity study similar to that performed in App.
C of reference 1.
The discharge line for valve PSV 21507 was chosen for this study as being representative of the three Rancho Seco lines. It was modeled with an average control volume length of 1.96 ft. Force time histories for both the steam discharge and subcooled water cases were developed and applied to the model. The line was
! then reanalyzed using a model with double the number of control volumes.
l The study showed that when more than 5 control volumes where used for a straight pipe segment, doubling the number of volumes increased the peak force by less than 10%. In two volumes the I force actually decreased. If fewer than 5 volumes were used in a segment, then doubling the volumes increased the forces from 10% to 30% in the steam blow-down case and 10% to 68% in the subcooled water case.
With this information, the intensification factors were applied as multipliers to the forcing functions as described below:
For pipe segments in which both the fine and coarse models had fewer than 5 control volumes an intensification factor of 2 was used, i
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c NUREG 0737, ITEM II.D.1 (CONT.)
RESPONSE 3: (Cont.)
For the remaining segments, which contained more than 5 control volumes in the coarse model, an intensification factor equal-to the ratio of peak forces in the fine model volume to those in the coarse model volume (with round-up margin) was used.
A more detailed discussion on intensification factors was provided in our April 12, 1985, submittal in response to question 8b. The overall thermal-hydraulic model as well as the coarse and fine nodalization models are shown as Attachments 6, 7, and 8, respectively, to that submittal.
REFERENCE:
- 1. EPRI Report, " Application of RELAP5/ MOD 1 for Calculation of Safety and Relief Valve Discharge Piping Hydrodynamic Loads" EPRI-2479, December,1982.
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NUREG 0737, ITEM II.D.1 (CONT.)
QUESTION 4:- The maximum time step used in the thermal-hydraulic analysis for subcooled water was 2 X 10-4 sec, based on Reference 1.
However, this reference did not consider subecoled water discharge. Analyses by EG&G Idaho, Inc. have shown that with subcooled water discharge a A x/ At ratio of about 20,000 f t/see is needed to optimize calculation of acoustic wave propagation.
With the coarse nodalization used at Rancho Seco the mini = =
node size was 1.25 ft. This would require a maximum At of 6.25 X 10-5 see rather than the 2 X 10-4 seco used. For the fine nodalization study the maximum A t should be one-half that calculated above. Provide assurance the maximum time step used in the RELAP5 analysis was small enough so that the forces calculated from the RRLAP5 output are not underestimated due to numerical smearing.
RESPONSE: The time step used in a finite difference solution is chosen according to the Courant criteria, as follows:
dt(dx/C Where:
C= acoustic wave velocity dx= node length dt= time step The saturated steam C is approximately 1500 ft/sec. The corresponding time step for a minimum node length of 1.25 ft. is therefore 0.0008 sec.
For subcooled water C is approximately 4000 ft/sec. The corresponding maximum time step for a node length of 1.25 ft. in this case is therefore 0.0003 sec. It should be noted that the acoustic velocity of 20,000 ft/see suggested in the question is on the order of the speed of sound in steel. This velocity is Considered to be overly Conservative for use in subcooled water analyses.
l l The maximum time step used in the thermal-hydraulic analysis was j 0.0002 sec. We feel that this was well within the criteria j limits to guarantee the RELAP output was not underestimated by l numerical smearing. This is consistent with the recommendations of section C4 of reference 1 (Question f3).
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NUREG 0737, ITEM II.D.1 (CONT.) i QUESTION 5: The RELAP5 thermal-hydraulic analysfa used a PORY opening time of 0.50 see for the Dresser PORY installed at the plant. A review of the test report showed opening times for the Dresser PORY were less than 0.23 sec. Confirm the PORY opening time
-used in the analysis is appropriate and results in the calculation of conservative piping loads.
RESPONSE: The original thermal-hydraulic analyses employed a PORV opening time of 0.050 sec. , as correctly stated in our April 12, 1985, responses to questions 8a and 8b. However, the opening time was mistyped as 0.50 sec. in our response to question 8c.
Therefore, the PORV opening time of 0.05 seconds was used in the analysis and is less than the 0.23 seconds cited.
NUREG 0737, ITEM II.D.1 (CONT.)
QUESTION 6: The submittal stated a cutoff frequency of 50 Hz was used in the piping analysis. The cutoff frequency of 50 Hz appears to be too low for piping r.nalysis based on EG&G Idaho, Inc.
experience. Most piping analysis uses a cutoff frequency of 100 Hz. The submittal mentions that 50 Hz is more conservative than the NRC guideline of 33 Hz. This guideline, however, is only for earthquake analysis. Provide assurance the use of a cutoff frequency of 50 Hz does not invalidate the analysis performed.
This also affects the maximum time step used in the analysis since the maximum time step is the reciprocal of 5'X (highest natural frequency).
RESPONSE: In the structural analyses of the pressurizer S/RV discharge piping, a cutoff frequency of 50 Hz was used in the seismic portion of the analyses. In the thermal-hydraulic dynamic analyses, however, a cutoff frequency of 300 Hz was used. The results of these two analyses were then combined by the SRSS method to obtain the total response of the system to dynamic loading.
The time step used in the analyses of thermal-hydraulic loads was 0.0005 see for the saturated steam case which compares favorably with the maximum recommended time step of the reciprocal of 5 times the highest forcing function frequency or 0.00067.
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4 NUREG'0737, ITEM II.D.1 (CONT.)
QUESTION 7: The submittal' indicated the limiting transient for piping stresses was the steam discharge transient and not subcooled water discharge. This result is different from the results of a similar B&W plant (TMI-1) which found, with similar initial conditions, that subcooled water discharge was the limiting case. .In addition, EPRI/CE testing found water seal discharge resulted in greater piping loads than steam discharge without water seals. Therefore, provide more details on the differences in the stress analysis between the steam discharge case and the subcooled water discharge case. Include reasons why the steam discharge case was the limiting transient. Also include a table comparing the calculated stress with the allouable stress for the most highly loaded pipes and supports for both the steam and water discharge cases.
RESPONSES: As discussed in our April 12, 1985 submittal (Question 8) the Rancho Seco thermal-hydraulic analyses considered transients involving both steam and subcooled water discharge through the SRV's and.PORV. Under these transients, the saturated steam discharge' cases generally produced higher thermal-hydraulic piping reactions. It was judged that these would be the more credible transients especially when combined uith seismic forces. The stress analyses for the subcoolel cases were therefore not completed.
The concern of greater piping loads from water seal discharge is not applicable to Rancho Seco since there are no loop seals at the safety and relief valves. Thus the extremely high dynamic forces due to a slug of cold water being forced down the discharge piping by high pressure steam were not analyzed at Rancho Seco.
Based on these analyses, the piping system was upgraded in 1986 to meet the anticipated thermal-hydraulic and other loads.
During the performance of confirmatory analyses to verify piping stresses under the as-built support locations it was determined that the system modeling required additional refinements to accurately predict the system response to dynamic loads. A reanalysis effort is currently underway to improve the model and fully verify system adequacy under all the design basis loadings including the subcooled cases. Upon completion of this analysis the results will be provided to the staff. It is anticipated that the reanalysis will be completed by April, 1987.
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NUREG 0737, ITEM II.D.1 (CONT.)
QUESTION 8: SMUD in its. response to our request for information stated that ,
the qualification requirement of NUREG-0737, Item II.D.1, '
pertains to qualification of the block valve and does not address qualification of the PORY control circuitry.. However, environmental qualification of the PORY control circuitry is exactly what was required by Item II.D.1. It is the staff position that in order to demonstrate the Rancho Seco 1 control circuitry is qualified to the requirements of NUREG-0737, the design qualifications must be compared to the environment the control circuits will be exposed to. Provide documentation to show the PORY control circuitry has been qualified under 10 CFR 50.49, or to allow a complete review of the qualification of the control circuitry for the PORY under NUREG-0737, provide the following:
A. Provide a list of all PORY control circuitry needed to mitigate NUREG-0737 transients.
_ _ . . _ B. For each item of equipment identified in A, provide the following:
- 1. Type (functional designation)
- 2. Manufacturer
- 3. Manufacturer's type number and model number
- 4. Plant ID/ tag number and location C. For each item of equipment listed above, provide the environmental envelope, as a function of time, that includes all extreme parameters, both maximum and minimum values, expected to occur during NUREG-0737 transients, including post accident conditions.
D. For each item of equipment identified above, state the actual qualification envelope simulated during testing (defining the duration of'the environment and the margin in excess of the design requirements). If any method other than type testing was used for qualification, identify the method and define the equivalent " qualification envelope" so derived.
E. Provide a summary of test results that demonstrated the adequacy of the qualification program. If any analysis is used for qualification, justification of all analysis assumptions must be provided.
7 NUREG 0737, ITEN II.D.1 (CONT.)
~ QUESTION 8 -(Cont.)
F. Identify the qualification. documents that contain detailed
. supporting information, including test data, for items ~D and E.
RESPONSEt' NUREG'0737, Item II.D 1 requested licensees to review relief and safety valve operating conditions expected.due to accidents and anticipated operational occurrences referenced in' Regulatory Guide 1.70, Revision 2. Resulting qualification of the safety and relief valves was to. include qualification of associated .
. control circuitry, piping and supports in addition to the valves themselves.
The District has_ developed a. list of equipment required for accident mitigation based on the accident and transient analyses contained in Chapter 14 of the Rancho Seco Updated Safety Analyses Report (USAR). This list was developed in response to
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10CFR50.49 and-has been documented by a District calculation.
- The results of this review reveal that the PORY is not required to provide primary system overpressure protection as a result of any USAR design- basis transient. In all Chapter 14 events where overpressure occurs in the primary system the safety relief valves are assumed to lift, thus providing overpressure protection. This is consistent with the description / intent of, the EMOV (PORV) and Code Safety Valves found in the USAR (Section 4.2.4). The EMOV is required to remain closed to maintain the pressure integrity of the RCS while exposed to post accident environmental conditions. As a result, the solenoid has been passively qualified to ensure that the valve remains in the closed position. Thus failure of the solenoid will not result' in a spurious actuation of the PORV as a result of short circuit, open circuit or loss of power in accident conditions.
Additional information on the qualification status of the PORY control circuitry is provided in Table 1. As noted in the Table all devices associated with the PORV circuitry have been qualified to 10CFR50.49 for the environment in which each is expected to operate.
The only time Rancho Seco requires use of the EMOV for overpressure protection is during LTOP conditions. Under the conditions that would exist in this event the current EMOV and associated circuitry has been found to be adequate to fulfill the overpressure protection function.
Based on this, it is the District's position that the PORV control circuitry does not require further upgrade of its qualification beyond the current qualification as presented in Table 1 and described above. All qualification documentation is available at Rancho Seco for your inspection.
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- TABLE 1 QUALIFICATION STATUS OF PORY CONTROL CIRCUITRY ENVIROIRENT&LLY QUALIFIED DEVICE TAC MANUFACTURER'S TYPE NO. PLANT LOCATION IBISER ,
NO. TYPE (PtBICTION DESIGNATION) MANUFACTURER AND MODEL NO. (BQ ENVIRONMDff) 10CPR50.49 S2A1 Vital 480V AC MCC Ceneral Electric Co. 325x915 West 480V Swar. Rm. Not Required Aux. Bids. (Mild)
PSV-21511 Solenoid General Electric co. S/N BM09686 Containment (Harsh) Yes*
Connector BNC to Mirco Dot Adaptor TEC Mone Containment (Harsh) Yee FT-21038 RCS Pressure Transaitter Rosemount 1153CD9 Containment (Harsh) Yes PT-21040 RCS Pressure Transmitter Rosemount 1153CD9 Containment (Harsh) Yes PT-21092 RCS Hot Leg "B" Control DHS Could PC3200-0$M Containment (Harsh) Yes Pres. Permissive Interlock PSH-21511 Pressuriser Relief Bailey Control 6623819-1 [H4 PIA] Control Room .Not Required (Mild)
PSH-21092 EM07 Low Setpoint Bailey control 6623819C1 [H4 PIA] Control Room Not Required (Mild) ,
IE-21524 Position Indicator on PORY TEC 2273A Containment (Harsh) Yes.
PSV-21511 XE-21525 Position Indicator on PORY TEC. 2273A Containment (Harsh) Yes PSV-21511 XY-21524 Charger Amplifier for TEC 504A Containment (Hersh) Yes XE-21524 XY-21525 Charger Amplifier for TEC 504A Containment (Harsh) Yes XE-21525 H3CRR Aux. Control Relay Panel General Atomic None West 480V Swgr. Rm. Not Required Aux. Bldg. (Ra. 217)
(Mild)
H1RC Reactor Coolant Console Babcock & Wilcox None Control Room (Mild) Not Required H4PIO1 NNI Cabinet Not Found None Control Room (Mild)'- Not Required H4EFA Nuclear Interface Automation Industries None NSEB Elect. Equip. Not Required Instrumentation Rack Room (Mild)
H4EFB Nuclear Interface Automation Industries None NSEB Elect. Equip. Not Required Instrumentation Rack Room (Mild)
H7RP50 Containment Penetration Conas 2325 Containment (Hersh) Yes H7RP63 Containment Penetration Conax 7073 Containment (Hersh) Yes H7RP21 Containment Penetration Conax 7073 containment (Harsh) Yes Splices Electrical Connection Raychem- Various Containment (Harsh) Yes Terminal Termination (H7RP50, Kulka 7TB8/7T512 Containment (Hersh) Yes Blocks H7RP63 H7RP21)
N/A 600 Volt Cable Brand Rex XLPE Insulation / Containment (Harsh) Yes Hypelon Jacket N/A 600 Volt Cable Cerro XLPE Insulation / Containment (Harsh) Yes Neoprene Jacket N/A 600 Volt cable Eatom XLPE Insulation / Containment (Harsh) Yes Hypelon Jacket-N/A 600 Volt cable Rockbestos XLPE Insulation / Containment (Harsh) Yes Neoprene Jacket
- Passive qualification meets 10CPR50.49, not qualified as an active component.
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ATTACHMENT 2 VERIFICATION OF THE RELAP5-FORCE HYDRAULIC FORCE CALCUIATION CODE
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