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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 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 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 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 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 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 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 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 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 ML20248H2391989-10-0606 October 1989 Requests Exemption from Requirements of 10CFR26 Re Fitness for Duty Programs Based on Present & Future Operational Configuration 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 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 1990-09-06
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$k u SACRAMENTO MUNICIPAL UTILITY DISTRICT O 6201 s street, Box 15830, sat.rartiento, California 95813; (916) 452-3211 n
't 347{ . -No vembe r 15, 1978 .
1
! , 3.d Director of Nuclear Reactor Regulation; ii fp
/
Attention: Mr. Robert W. Reid, ChiefW .i Operating Reactors, Branch a M 3 U. S. Nuclear Regulatory Commission '., -
Washington, D. C. 20555 .:
Docket No. 50-312! ,'"
Proposed Angndment No. 60 .
Rancho Seco Nire Station,UnitNog-Q,,.r Generating 7p. . *[* ['
Dear Mr. Reid:
Your letter of November 7,1978 requested additional information in support of Proposed Amendment No. 60 to Operating License DPR-54 for Rancho Seco Nuclear Generating Station, Unit No.1. This proposed amendment changes the Technical Specifications to permit operation with the fuel loading planned in Cycle 3. The requested information is attached to this letter.
Sincerely yours, l
- , TL[ vt
'/J.Mb.
Mattimoe Assistant General Manager and Chief Engineer l
Attachment l
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, AN ELLCTRIC sYSTE M S E R VIN G f.10 R L THAN C 0 0,0 0 0 th 1HE H l: ART Of calif 0PNIA
4
= ANSWERS T0~ QUESTIONS'0N PROPOSED' AMENDMENT'NO. 60 :
- 1. 1 question: Provide the maximum ejected ~ rod worths at the 10-3 .
1 power level and 100% power level for B0C and EOC conditions and state whether these are bounded by the values used in the FSAR. If these values- are not bounded by previous analysis provide the peak enthalpy for the hottest fuel rod.considering rod ejection at each of the above mentioned' power levels.' ;
Response: The. zero power- ejected rod worths are provided in Table 5-1 of BAW-1499. The maximum full power ejected rod worth (a BOC. value)-_
is provided in Table 7 .1 of BAW-1499. The values listed for the maxi-mum ejected rods at the various conditions are well below FSAR values as stated in BAW-1499. For your information, the HFP,'E0C value is 0.27% Ap.
- 2. Question: Provide the post-ejection maximum peaking factors and compare these to what was used in the FSAR.
Respo_nse:
os The methods and techniques for rod ejection accidents, which have been approved by the NRC, are described in the FSAR and do not' depend on ' post ejection peaking factors. The approved techniques have ber benchmarked against space-time-kinetics methods (See the FSAR).
- 3. _ Question:iState the reason -for withdrawal of the APSR during the '
20 days of Cycle 3.
_ Resp,on se : The APSR's are being withdrawn during the last 20 day 1 cycle 3. to gain treactivi.ty which will increase cycle length, e . ovide
- a more' efficient utilization of. the fuel.
~
- 4. Question: Define the' axial stability index and describe how the numerical values presented on Page' 5-2 were obtained.
Response: Sufficient axial stability was demonstrated'by perturbing the xenon
- andfiodine con'centrations in several different ways and then calculating ~ the stability'index. . The stability index is the variable (b)'in the following equation: (which can be derived from Reference 1).
J l
2 OS.=:OSo+ebt[A] SINH +ACOS$]
2 whe re' OS = offset at time t i OSo= equilibrium offset T = period of oscillation In using this equation .to describe the perturbed offset behavior of the core, it is assumed that the offset is a sinusoidal function that decays exponen.ti ally. If b is positive, the oscillation- is divergent. A value of zero means an indefinitely sustained oscillation and a negative wlue ,
describes, a convergent one. ]
The xenon and iodine were perturbed in .three ways:
'l) The APSR withdrawal procedure (Figure 1)
- 2) A 100-50-100 ' design transient.with return to full power ;
at peak xenon (Figure 2)-
- 3) A return .to full power with 50% power equilibrium xenon ]
and iodine concentrations.
. m . -_ _
The _ stability index was not calculated for. the first case,-but inspection of the offset behavior showed it to be convergent. The stability index for the second case was -0.051 hr-1, and -0.061 hr-1 for the third case.
This amply demonstrates' tfe axial stability of the core. In addition, the maximum _imb'alance encountered after the return to full power was
-10%-and +10%. These are within the LOCA imbalance limits which are
-32.8 and +17.1 percent-imbalance.
The damping of the oscillation can be enhanced by insert Se control rods several percent as the imbalance goes through .its pos.:.i.e peak,
~
or withdrawing the rods as the imbalance goes through its negative peak.
The effectiveness of this can be seen in figure 1. This figure shows the imbalance behavior with no rod movement after the return to full power from the APSR withdrawal (Case 2). The maximum positive imbalance encountered in case 2 is 4.4 percent less than in case 1. The subsequent negative peak is also less for case 2 by 1.1 percent.
- 5. Ques tio_n: State whether the axial stability index was calculated with !
NRC approved models.
Response: The axial stability index was calculated as described in the i response to Question 4. The FLAME computer code (2,3) was used to i supply input to the calculation. References 1, 2, and 3 have been approved by the NRC.
- 6. Question: What is the maximum value of FAH predicted for this fuel cycle?
R_esjonse: The maximum theoretical value of FAh is 1.54 (no xenon peak) at BOC. This provides an 11% margin to the design value of 1.71 which l is more than adequate to cover calculational uncertainties. As indicated l in the response to Question 12 though, the actual maximum value of FAh at operating conditions (with xenon) is 1.51. Thus margin in excess of 13% exists. The max. predicted Fah decreases. to 1.35 at E0C for a 27% margin to 1.71.
- 7. Question: What is the basis for the selection of 1.71 for FAH?
Response: The 1.71 value W Fah was chosen for SMUD Cycle 3 because experience has she-
- nFAh of 1.7) is more realistic than the previously used 1.. 'l provides conservative margins to steady state and maneuverins limi ts . This value has been used for licensing Davis-Besse I Cycle 1, t, I Cycle 5, Oconee II Cycle 4, and Oconee III Cycle 4.
- 8. Question: In determining the flux / flow trip setpoint it is stated that an assumed one pump coast down was analyzed. Justify that this-is the worst event of this class when compared with a total loss of flow.
Response: The flux / flow trip setpoint is derived based on reaching the design minimum DNBR plus . applicable rod bow penalty during the one
; pump coastdown. The minimum DNBR used for the cycle 3 flux / flow set-point was 1.43 (B&W2) based on' the correlation limit plus 11.2% rod bow penal ty. Therefore, since the limiting.DNBR is used in the flux / flow
- analysu, it is; by definition the limiting flow coastdown. All other possible pump coastdown combinations will cause a reactor trip to be initiated by the pump' monitors. 0f these, the four pump coastdown is most limiting. The minimum DN3R during the four pump coastdown is 1
/
.a 4
- 1.660for(16% margin.to the design minimum.
- 9. Question: a. ilt:is noted that in the Technit al Specifications reactor -
coolant system flows previously stated in ibm /hr have 'been changed to gpm. _ Explain the need for'this change and verify that the new flows were calculated using the. reactor coolant density corresponding to the identified power levels,
- b. Explain why the title of' Figure 2.3-2 was changed from -
that:specified by the reactor vendor or restore it;to -
its original wording. If the intent of this change in-
-wording is to allow maximum trip, setpoints_- above Curves --
1, 2 and 3 for. the. indicated pump configurations , provide analyses to' justify such higher setpoints.
R_esponse: a. . Reactor coolant' system flows can be expressed equally well. in lbm/hr or gpm for various power / pump combinations. Flow rates in gpm are independent of rower level whereas lbm/hr calculations necessitate densi.ty corrections'. .Since ' reactor coolant pumps .are constant '
volume pumps gpm is more representative.of their operation. The values in gpm in the Tech Specs were calculated using the' reactor- coolant density corresponding' to the identified power levels.
b._ The title of Figure 2.3-2 will be restored to its -
original wording of " Protective System Maximum Allowable Setpoints, Reactor Power-Imbalance. "
10._ Question: Figure 2.1-2 ' Identify the specific credits taken in percent to allow the reactor power imbalance tent to change as' indicated. For example, what credits were taken to allow a 10% increase in power at
+50% and -50% axial imbalance .
Response: The credits taken in modifying the power imbalance tent-in Cycle 3 from that used in Cycle 2 were simply the differences in the three dimensional power distributions ' calculated for the two cycles.
- Table 1 below'shows the differences in FLAME total peaks (adjusted for radial-local, i.e. ,' hot pin to assembly average power) for the nominal fuel cycle depleti,on throu. gh- 100, EFPD.
c;
-TABLE l' Comparison of Cycles.2 and 3 Total
~
'- Peaking Factors r
TTime'(EFPDJ . Cycle 2 Cycle 3
- _ .t s0._ 7 1.93 1.85
~4' l.82' 'l.73 a25, ,
- 1. 81' 1.67-150: .
- 11. 8 0 / l'65.
, '100 1._76 1.63' L
[
I S. ,
- / w
4 s
- Further, comparison of similar extrene cases which led to the power imbalance tent limits .is shown .in Table 2.
TABLE 2 Comparison of Extreme Condition Cases for Cycles 2_ and 3 Cycle Negative Offset / Peak Positive Offset / Peak
'2 -35.3 2.75 16.9 2.17 3 -42.6 2.68 16.3 1.94 The imbalance limits derived from 1300 FLAME cases, including variations in burnup, xenon concentration, and control rod and APSR positions are given in Table.3.
TABLE 3 Comparison of Actual RPS i Imbalance Limits for Cycles 2. and 3 Power l L evel , Cycle 2 Cycle 3
%FP Neg. Imb. Pos. Imb. Neg. Imb. Pos. Imb.
. 112 -28.3 +16.1 -32.5 +35.8 100 -40.5 +44.1 -45.0 +60.0 80 -54.3 +70.1 -56.8 +83.2
- 11. Question: Quandrant Power Tilt - Justify the use of the values of ,
4.92%,11.07% and 20% for quadrant tilt on this reload. These values '
have been used in the Standard T/S as typical.
Response: As stated in BAW-1499, page 8-1, the value of 4.92% actual quadrant power tilt as a steady state limit was accounted for by the use of a peaking factor of 1.0736 in the development of limits on control rod positions and imbalance. All values of the quadrant tilt ;
limits proposed for use in cycle 3 have been reviewed and approved for i use in Davis Besse I, Cycle 1 and Crystal River III, Cycle 1. The same or larger values were approved for ANO-1, Cycle 3 and TMI-1, Cycle 4.
- 12. Question: Verify that the consequences of all the accidents and transients analyzed in the SAR are acceptable considering: (a)4.92% i tilt as an initial condition; and.(b) 20% tilt at 60% rated power as !
an initial condition.
Response: (a) A 4.92% actual tilt would cause a_ maximum peaking increase of 7.36%,f as stated above. The maximum hot pin radial peak k
4 observed in Cycle :3 at operating conditions is 1.51 (assembly radial of 1.39 from Figure 5-1 of BAW-1499 times radial-local factor of 1.087). i The design radial peak used as .the initial condition for all accidents l and transients -(1.71) is 13% larger than the maximum hot pin for this- l cycl e. Thus, theiconsequences of all accidents are acceptable consider- 1 ing a steady state quadrant tilt limit of 4.92%.
(b) When the reactor is operating at. initial power. levels less than rated power the initial steady state DNBR isL significantly larger than that at rated power operation. This increased initial ,
DNBR provides increased margin to the MDNBR limit for transients 1 initiated at less than rated power. . Four pump operation at rated -
power represents the gieatest power to flow combination for initial l conditions:for DNBR evaluation following transients hnalyzed in the l SAR.
Allowable power tilts of the magnitude presently considered for Cycle-3 have 'been found licensable for other 177FA B&W plants as enumerated in response to Question. ll. i
- 13. Question: Justify the changs to the core imbalance vs power level trip tents (Figures 8-11, 8-12 and 8-13 of BAW-1499) and explain why l the permissible operating region falls outside the maximum setpoint j values for reactor power imbalance shown in Figure 8-2. ;
i Response: The imbalance versur power level tents'shown in Figures l T391 5Fough 8-13 are derived cased on LOCA Kw/ft criteria and are "
the actual limit values - not adjusted for measurement errors. The changes from Cycle 2 to Cycle 3 are the result of changes in the core i peaking characteristics, as discussed in the answer to Question 10. I Figure 8-2.shows the measurement error adjusted trip setpoints of the Reactor Protection System. In deriving these setpoints from the actual limits shown. in Table 3, many conservatisms are introduced. If, after the limits of Figures 8-11 through 8-13 are also error adjusted for alarm setpoints', they still fall outside of the trip setpoints, no problem arises. The result would simply be a trip signal prior to reaching a condition less severe than those for which the RPS provides protection. . "In' actual practice, the alarm setpoints would be moved I inside'of the' trip 'setpoints .to provide a warning to the operator that l a trip condition ~ Was being approached. l
REFERENCES:
l 1
- 1. Stability Margin' for Xenon Oscillations - Modal Analysis, BAW-10010, !
P_ art :1, Balkock and' Wilcox', August 1969.
1
- 2. ; FLAME ;Three-Dimensional Nodal Code for Calculating Reactivity !
and Power Distributions, BAW-10124A, Babcock &.Wilcox, August 1976.
- 3. Verification of Three-Dimensional FLAME Code, BAW-10125, Babcock &
Wilcox, August 1976..
- 4. ,0perational Parameters for B&W Rodded. Plants, BAW-10078, Babcock &
- Wilcox, September 1973.
i I
1
14; Question: ' Vour description of the ejected control rod reactivity worth test in BAW-1499 does not state that four symmetric control rods will be measured. As stated in BAW-1477."0conee 1 Cycle 4 Quadrant Flux. Tilt" page 12, this test "has proven to be an indicator of core
- symme try" . Please indicate if the measurement of ejected rod worth of four symmetric: location is part of your test program for the Cycle 3 core.
Response: Measurement of the ejected rod worth at four symmetric .
locations is not part of the standard restart test program. Howeve r, the detailed zero power physics test procedure incorporates the requested measurement of symmetric ejected rod worths at four core locations.
- 15. Question: In Section 9.4 you describe actions to be taken if the -
Acceptance Criteria are not met. This description is too general for the.measumment of rod worth. The usual action to be taken if the sum of the worth of groups' 5, 6 and 7 differs from predicted by more than 110%, is. to measure group 4 by dilution and take additional measurements and make an evaluation of the. discrepancies. Indicate ^your commitment to thes'e corrective actions -if the acceptance criteria for measurement
~
of control bank reactivity worth are not met.
Response: The : action to be ta' ken in the event the total measured worth of Groups 5-7, differed from the predicted value by'more than i10% is-to perform an; evaluation consisting of. one or more of the following
' items as appropriate to the situation:
- 1. Review of measurement data anf data analysis.
- 2. Verifjcation that the avilable shutdown margin based on the measured data satisfies the minimum shutdown margin require-ment.
- 3. Review of the. results of other physics test.
- 4. Review of calculations used to obtain the predicted value.
Evaluation of the impact of the discrepancy on safety of
- 5. -
operation and on Technical Specifications limits, if any.
- 6. Determination as to whether retest of one or more of the regulating groups would be required.
- 7. Determination as to whether measurement of one or more of the safety. groups would be required based on considerations of the. extent and nature of the discrepancy and of-item 5 above.
If it 'is determined' that measurement of one or more of the' safety groups would be required to resolve the discrepancy, such measurements will be performed.
1
.. :l
. j l
.16. Question: Indicate your commitment to submit a physics startup l test report within 45 days of completion of the tests.
Response: The results of the startup. physics program for Cycle 3 will be compiled and submitt'ed with one of the monthly. reports filed within 90 days after completi'on of the physics test program. Startup. j test data is availab'le for review at the station by NRC personnel prior to the fonnal submittal.
i 17 Question: With regard to your request to modify T/S 3.3.3, the statement i that the proposed provision is like. the Standard Technical Specifications (STS) is not' sufficient justification for the change. The reason this is insufficient is because the STS are baseion a number of interrelated p rovi sions . Thus the adoption of a single provision without including j all of the related provisions may not be technically justified. Accord- )
ingly, if you desire to-make this change, you s1ould revise your sub-mittal as necessary to incorporate all STS operebility and surveillance requirements for engineered safety features in the Rancho Seco T/S, or propose a lesser revision supported by a detailed evaluation which demonstrates the safety of the new' provisions relative to the safety of the present provisions.
Respons,e: The request to nodify . Technical Specifications Section 3.3.3 has been a subject of concern during Plant Review Committee discussions and NRC inspections. Both the Plant Review Committee and NRC inspectors !
agree that the proposed change in this' Technical Specifications will increase the safety of- operations.
We believe that this provision of the Standard Technical Specifications contains a value judgment with which we concur; i.e., it is far safer to accept as a risk the very low probability that a safety features system, subject to a reasonable set of surveillance requirenents, wculd fail to start on demand rather than deliberately create a situation in which there is a certainty that both SFAS systens will be out'of i service' for a period of time.
If after re-evaluation you do not concur with this judgnent, we will withdraw our request for a Technical Specification change.
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