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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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( )SMU-SACRAMENTO MUNICIPAL UTILITY DISTRICT O 6201 S Street, P.O. Box 15830, Sacramento CA 95852-1830.(916) 452-3211 l AN ELECTRIC SYSTEM SERVING THE HEART OF CALIFORNIA 1
l AGM/NPP 89-040 March 24, 1989 x
U. S. Nuclear Regulatory Commission Attn: Document Contrid Sask Hashington, DC 20555 Docket No. 50-312 Rancho Seco Nuclear Generating Station License No. DPR-54 PROPOSED AMENDMENT 127. REQUEST FOR ADDITIONAL INFORMATION Attention: George Knighton Pursuant to your letter dated February 9,1987, attached is the District's response to the request for additional information regarding Proposed Amendment 127.
Members of your staff with questions requiring additional information or clarification may contact Mr. Robert Roehler at (209) 333-2935, extension 4918.
Sincerely, h(b d b.tuv+
4 Dan R. Keuter Assistant General Manager Nuclear Plant Manager Attachment cc w/atch: J. B. Martin, NRC, Halnut Creek A. D'Angelo NRC, Rancho Seco gg32;ggggggggggp y
P RANCHO SECO NUCLEAR GENERATING STATION O 1444o Twin Cities Road, Herald, CA 95638 9799;(209) 333-2935
Attachment
- AGM/NPP 89-040 Page 1 of 10
- 1. NRC REQUEST Sacramento Municipal Utility District (SMUD) has indicated in its letter dated November 14.-1985 that the assumptions used in B&W Document 86-1153322-00, " Rancho Seco: Main Steam Safety Valve (MSSV) Analysis," is consistent with the assumptions used in the B&W Topical Report BAH 10043, " Overpressure Protection for B&W Pressurized Hater Reactors." The former report concluded that Rancho Seco could meet the requirement of overpressure protection of the secondary system when the plant is operated at 112 percent of rated power with three out of nine MSSVs per steam generator inoperable. However, the latter report concluded that the capacity margin for the main steam safety valves is only 6 percent. Provide discussion on the above discrepancy in capacity margin for the MSSVs and justify that the proposed change to the technical specification will provide reasonable assurance of adequate overpressure protection of the secondary system.
DISTRICT
RESPONSE
B&W Topical Report BAH-10043, " Overpressure Protection for B&H Pressurized Hater Reactors" was issued in 1972. This topical report is based on analytical results from the POWERTRAIN computer code and shows that the B&W plant design is adequately protected against overpressure events on both the primary and secondary sides. The POWERTRAIN code simulated the NSSS, i 9
including several key secondary plant features such as the main feedwater system from the feedwater pump suction to the Once Through Steam Generators (OTSGs), and the steam lines from the OTSGs to the main turbine. Major control and relief valving was also included in the model. It was used extensively to .
simulate many upset (anticipated) events and power maneuvers because it included in its simulation a model of the Integrated Control System. Since the version of POWERTRAIN that generated the results cited in BAH-10043 preceded the operation of the first B&H commercial power plant, there was no opportunity to benchmark the code to actual operating plant data prior to its initial application. However, pruder.t selection of conservative inputs was considered adequate to ensure conservative results from the code.
More recent computer codes and models (and later versions of POWERTRAIN) have been refined to provide more realistic estimates of actual plant transient performance. RELAP5/ MODI is one such code that provides best estimate NSSS responses for a wide spectrum of events. Through the application of the code to actual plant analyses and its benchmarking to the Multiple-Loop Integral System Test Program and to operating plant data, the code has been validated quite extensively. The analyses performed to suoport the Rancho Seco Technical l
' Specification revision reflect a more accurate assessment of !
plant response than did the original P0HERTRAIN work reported in BAH-10043. By continued application of the original acceptance criteria and conservative assumptions, the design basis for overpressu e protection remains unchanged.
1
Attachment AGM/NPP 89-040
[
Page 2 of 10 B&W Document 85-1153322-00 uses the same conservative assumptions as BAH-10043 and shows that the maximum pressure reached on the secondary side is about 1165 psia, which is 5 psi below the acceptance criteria with the assumption of the three lowest setpoint MSSVs out of service. As noted in the previous submittal, however, the MSSV model assumed that all functional MSSVs on each OTSG header lift at 1.M above their nominal setpoints and do not reach full release capability until 3% accumulation. No credit is taken for increases in steam flow rate that would occur as steam pressure increases above the pressure associated with the full. lift point.
Consistent with all other safety analyses, the MSSV flow rates are taken to be vendor-stamped rated, not those expected at full lift.
Even though the analysis assumed the three lowest setpoint MSSVs were out of service for the limiting case that is reported, this valve configuration would not be allowed by the ASME Code. The Code requires at least one safety valve be set at the design pressure for the protected system. Here both of the MSSVs set at this pressure on any one OTSG declared inoperable, another valve would have to be reset to this value or the inoperable valves _ restored to operable condition.
Hence, there is additional margin in the reported results since lower set valves in service have a larger impact on peak pressures than those set at high pressures.
Similar analyses (4), performed using the TRAP 2 code (5) for another B&W 2772 MHt plant with similar installed MSSV capacity, also demonstrate large margins in the installed MSSV capacity. This analysis used the same assumptions as BAH-10043. With the three lowest MSSVs assumed out of service and all remaining MSSVs set at 1100 psig, the turbine trip event from 112% power yielded a peak secondary system pressure of 1181 psia. This result is for a situation more conservative than was analyzed for Rancho Seco because of the assumption of ,
high setpoints for the operable MSSVs; however, the peak secondary pressure is within 16 psi of the RELAP result. The TRAP 2 model was a complete NSSS simulation and showed that for this case the peak RCS pressure was 2681 psia. Plots showing these results are included in Figures 1 and 2. The same case wasrerunwitg81pnly at 1100 psig. two The MSSVs results out ofaservice showed and all others peak pressure in the set OTSG of 1170 psia and peak pressure in the RCS of 2678 psia.
These results, shown in Figures 3 and 4, demonstrate the relative insensitivity of the RCS peak pressure to the MSSV capacity in service.
The operating history of B&W plants since the first unit's startup has confirmed that the original overpressurization analytical results were conservative. A survey of all D&H plant reactor trips from January 1980 through December 1987 showed that only four of 253 plant events resulted in RCS pressure above 2400 psig. W The allowable design overpressure is 2750 psig. This survey is based on B&W Owners
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Attachment
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,. .AGM/NPP 89-040 Page 3~of-10 Group (B&WOG) Transient. Assessment Reports'. During this period the plants had a high. pressure trip setpoint of 2300 psig and m anticipatory reactor trips on loss of both main feedwater pumps and on trip of the main turbine. However, the PORV setpoint was 2450 psig. The four events that did exceed 2400 psig involved post-trip events in which RCS pressure was increasing-relatively slowly due to slow coolant overheating or because of HPI operation that increased the pressurizer level.. In such cases, the pressure relief capacity requirement is a small, fraction of that required to protect against=the design basis.
event (a continuous rod withdrawal event from zero power).
Operating experience is not well_ documented prior to 1980.
During this period the high pressure trip setpoint was-2355 psig, the PORV was 2255 psig.. No anticipatory trips were-in use. One lift of an RCS safety valve is thought to have occurred at one of the Oconee plants when _the pressurizer spray and PORV were both isolated for a turbine trip test from 100% power. Peak RCS pressure was approximately 2500.psig.
It is not known whether one or both RCS safety valves lifted.
This event and other operating experience demonstrates that.
the B&H operating plants are well protected against RCS over-pressurization. Although none of-these events involved as limiting assumptions 'as were used in the topical report, the wide margin to the overpressure limits exhibited by actual plant experience suggests that the analyses were significantly conservative.
The same observations are true with regard to secondary side peak pressures. It has been a rare occurrence at a B&W plant for steam pressure to exceed 1100 to 1110 psig as measured on the steam line pressure instruments. Most plant trips involve peak pressures in the range of 1060 to 1100 psig. This compares very favorably to maximum design overpressure of 1155 psig.
In summary, the computer codes in current use, i.e., RELAP5 and TRAP 2, have been demonstrated to provide reasonably accurate predictions for plant upsets. As a result, excess margin that was provided for in the original equipment design can be '
quantified and used to provide additional operational flexibility in the plant without jeopardizing equipment safety or the health and safety of the public. The cumulative operating record for B&H designed plants supports the conclusion that approximately 30 to 35% excess MSSV capacity exists in these plants. Forced removal of up to three MSSVs per OTSG, in itself a highly unlikely occurrence, would not jeopardize either primary or secondary side overpressure protection. This conclusion is based en the simulation results which show that three valves can be taken out of service l' without exceeding the acceptance criteria for overpressure protection even when extremely conservative assumptions for several systems or equipment operability are applied.
I Attachment J
.. . AGM/NPP 89-040
- 2. NRC Page'4 of 10 REQUEST SMUD presented, in B&W Document 86-1153322-00, the results of a benchmark analysis which compared the results of an analysis J using RELAP5 with the Davis Besse turbine trip transient data !
obtained in November, 1982. However, the RELAPS model does not ,
account for any turbine bypass relief capability while the i Davis Besse transient includes the actuation of the turbine bypass and the atmospheric dump valves. The staff does not consider the above stated benchmark analysis valid.
Provide justification, including supporting code verification and appropriate sensitivity analysis demonstrating the .
conservatism of the methodology used, for the use of the P.ELAP5 code for calculating secondary system overpressure transient.
DISTRICT
RESPONSE
The staff's observation that the benchmark case used in reference 3 did not simulate the turbine bypass valves (TBVs) is correct. This.was known prior to making the comparison case and was done so with the following rationale. At the Davis-Besse plant (from which the benchmark data was taken),
which in this regard is similar to all other B&W units, the TBVs have a lift setpoint of approximately 1015 psig for post-reactor trip steam pressure control. With.an anticipatory trip installed, reactor trips occur promptly after a turbine trip and the TBVs immediately revert to a post-trip control mode. The effect of this is that the TBVs do not begin to open until steam pressure has exceeded their setpoint. The TBVs are air-operated valves with a stroke time of approximately 5 seconds; hence, following a turbine trip the TBVs do not reach their full open position for at least 5 seconds after the setpoint is reached. By comparison, the safety valves have opening times measured in hundreds of milliseconds. The effect on the peak steam pressure is relatively minor since it occurs within a few seconds after the turbine trips.
In the case used for the benchmark, the plant data confirms the model's ability to properly account for the rapid change in the primary-to-secondary heat transfer and to calculate the rate of increase in the steam pressure. The comparison clearly shows that RELAPS predicts a conservatively rapid rate of increase in steam pressure which begins promptly after the steam flow to the turbine stops and before any possible TBV's impact on the secondary steam pressure.
In addition, the RELAP5 model shows good agreement with the TRAP 2 predictions for both the rate of steam pressure increase and the peak secondary side pressure that is reached during the event.
In summary, the RELAP5 model bounds the rate of increase of steam pressure following a turbine trip as evidenced by comparison of the model's prediction to plant data. The RELAPS results should be considered valid and conservative estimates of expected plant response for the turbine trip event.
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..c .' 'AGM/NPP 89-040.
Page 5 of.._10.
=3. NRC-REQUEST SMUD stated in a' letter dated November 14, 1985 that.the
_ proposed changes of the technical _ specification regarding the -
MSSV availability do not involve' aisignificant' reduction in a margin of ' safety. Justify this _ determination:in light of the
' reduction in pressure relieving capacity margin of the secondary system due to the proposed technical specification .
change to Section 3.4.2.1, which will permit a 33% reduction in z secondary system pressure: relieving capacity during.ful1~ power operation at Rancho Seco. ;
. DISTRICT RESPONSE-The margin of safety designed into the plant would not be reduced by. permitting plant operation.with up'to three MSSVs out of service because.the safety margin provided by the design remains unchanged. The proposed amendment to the Technical Specifications does not alter the design basis for the i overpressure; protection of.either the RCS or the secondary side.: The design basis _ transient, i.e. the-turbine trip from i overpower conditions, and its attendant conservative- !
assumptions, are the same as used in the original overpressure l protection _ analysis'. The proposed' amendment _ seeks to take ,
advantage of the additional margin in the plant between the l plant's expected response and_ the original acceptance' criteria for the event.
The Rancho Seco plant design was defined prior to the time that -1 any 177 FA plants were operational, thus preventing direct j feedback of plant data into the design process. Sinc 6 that -
-time, several. features of the plant's_ conservative design have become known'either through direct observation of the plant's behavior or enhanced analytical techniques. The main steam j relief system is one in which.large design margins have been ;
identif.ied and equipment reliability has been confirmed. j Since the B&W plant design results in lifting of the MSSVs following all medium to high power reactor trips, there is l
- substantial transient experience to verify the reliability of the MSSVs to lift on demand. Studies performed by the B& HOG l have concluded that the HSSVs are v9ry reliable in their action to lift and relieve steam pressure.t8> To enhance the MSSV reliability, especially with regard to the blowdown characteristics, the B&H0G has developed generic maintenance guidelines for these valves. Thus, MSSV reliability to properly reseat should become commensurate with their reliability to lift. Operating experience in 1988 suggests that this improvement is already being achieved.
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Attachnent AGM/NPP 89-040 Page 6 of 10 Excess MSSV capacity has been quantified through the use of_ improved. computer codes and field data, as discussed in District Response 1,. and has confirmed that there is significant margin to the overpressure protection acceptance criteria of the plant. The TRAP 2 analysis is an independent demonstration of this margin in the overpressure protection equipment.td) .B&H expects that a TRAP 2 analysis using MSSV setpoints as assumed in the Rancho Seco analysis would have confirmed the reported Rancho Seco results.
In summary, six of nine MSSVs on each OTSG are adequate to provide for overpressure protection for the secondary side of the OTSGs and steam lines. With this Technical Specification amendment there will be no loss of safety margin of the plant; hence, plant equipment and the health and safety of the public will be adequately protected against overpressurization events, l
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Attachment
." AGM/NPP 89-040 Page 7 of 10
- 4. NRC
. REQUEST l- In the IE Information Notice No. 86-05 regarding MSSV test failures and ring setting adjustments, NRC has expressed its concern on the MSSV problems. Describe the testing and maintenance program used at Rancho Seco to preclude similar 1 problems from occurring and thereby provide assurance of I.- proper MSSV operation.
DISTRICT
RESPONSE
The District has a number of programs in place to assure proper MSSV operation.
- Increased monitoring of MSSVs during plant transients using acoustic monitors. The acoustic monitors are calibrated using values established during flow tests performed during our 1988 restart and test program.
- a. PMs have been generated for setpoint verification and periodic monitoring (using ultrasonic equipment) to identify leakage.
- The requirements of ANSI /ASME OM-1 have been incorporated into our mainttuance and testing procedures.
- Active participation in the B&WOG Valve Task Force.
Recommendations of the Valve Task Force have been incorporated into our procedures. Valve performance has been evaluated following plant transients using the recommendations of the Valve Task Force.
1
Attachment AGM/NPP 89-040 Page 8 of 10
- 5. NRC-REQUEST In lir' ' of the industrial experiences on the use of MSSV ident sd in the IE Information Notice No. 86-05, provide justii .ation on how the proposed technical specification will provide secondary system overpressure protection with reliable facilities and high confidence level.
DISTRICT
RESPONSE
The District considers that the procedures, programs, and testing for these valves provide high reliability and resolve the valve disc travel issue. The following summarizes this position:
- Procedures and programs currently in place incorporate vendor and industry recommendations with Quality verification points to ensure that the ring settings after maintenance are in accordance with Dresser Industries recommendations.
- The data from recent trips and testing show better responses for setpoint and blowdown. '
e The District's programs (monitoring and testing) have identified one valve that does not respond well. This valve has been gagged and taken out of service.
- Toledo Edison has recently performed extensive testing of Dresser valves of the same size and model as those at Rancho Seco. The ring setting combinations tested did not result in inadequate disc travel. Based on this testing it appears that the disc travel issue is not applicable to Dresser valves.
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. AGM/NPP 89-040 !
Page 9 of 10 L 6. NRC ]
! REQUEST j Section 1.1 of the B&W Document 86-1153322-00, " Rancho Seco: l l Main Steam Safety Valve (HSSV) Analysis," states that. primary I system pressure is not of major concern in this analysis as L events which require MSSV actuation result in primary side L pressurization to the Reactor Protection System (RPS) high i- pressure setpoint and consequent reactor trip within the first ten seconds of the. event. For design basis event, a turbine trip from overpressure, an anticipatory reactor trip prevents ]
any significant RCS pressure increase. This statement is not !
consistent with the assumptions listed in Section 3.1 of the I report which states that no direct trip is assumed to result J from turbine trip. Also, the report does not provide I sufficient discussions to address the effect of the primary 1 system pressure transient due to a reduction of the secondary system pressure relieving capacity. Provide the results of such analysis.
DISTRICT
RESPONSE
Section 1.1 of the referenced report (8) intended to make clear the following points. First, the bounding case for secondary side overpressurization does not cause an over-pressure condition on the primary system due to the RPS shutdown of the reactor on high RCS pressure and the action of the pressurizer safety valves to limit the RCS pressure rise.
Second, Rancho Seco has in place an anticipatory reactor trip on turbine trip that would virtually eliminate any RCS pressure increase following a turbine trip, even if from an initial !
condition of overpower. The RPS high pressure trip function is !
assumed operable in all safety analyses scenarios except ATHS, whereas the anticipatory trip on turbine trip is not. The acceptability of the proposed Technical Specification amendment is not contingent ors continued use of the anticipatory reactor trip on turbine trip.
In the analysis performed to support the District's submittal, the reactor coolant portion of the OTSGs was modeled by including the appropriate reactor coolant flow rate, inlet temperature and pressure for 1121 power as boundary conditions. These inlet conditions were held constant through the transient. This assumption is j% tified by the fact that ;
the peak pressure in the OTSGs, whi.rh accurs in about 4 to 6 seconds, is reached before any tempetture feedback due to the upset can affect these inlet conditions. (The RCS loop transport time is about 12 seconds.) RCS pressure results are not available from that work; however, results from later work done on a similar plant show that the RCS overpressure design limit will not be exceeded during the turbine trip event from overpower. These later analyses were performed with assumptions consistent with BAH-10043 assumptions. u) The .
plots of RCS pressure from this analysis are provided in l conjunction with District Response 1. As previously discussed, the maximum RCS pressure from that evaluation
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. AGM/NPP 89-040 Page 10 of 10 varied between 2681 and 2678 psia for the two cases' involving three and then two MSSVs out of service.. All-functioning MSSVs were set:to begin opening 1% above their lift setpoint of -
1100 psig and did not reach their full open relief capacity until 3% above this setpoint. These results show that the peak'.
RCS pressure reached following a turbine trip is somewhat insensitive to small changes in the MSSV relief capacity and that the peak RCS pressure limit of 2765 psia will not be exceeded. These results should bound the Rancho Seco plant since Rancho Seco will continue to use a staggered set of MSSV setpoints.
_ REFERENCES
- 1. NRC Letter J. F. Stolz to J. E. Hard, " Request for Additional Information Related to the Technical Specification Change Request Regarding Main Steam Safety Valves," Docket No. 50-312, dated February 7, 1987.
- 2. B&W Document No. 51-1171572-00, " Maximum RC Pressure Data: Peak Pressure Events," dated April 13, 1988.
- 3. B&W Document No. 86-1153322-00, "SMUD MSSV Analyses," dated November 13, 1984.
- 4. B&W Document No. 32-1170334-00, "MSSV Analysis," dated 10/27/87.
l l S. B&W Document No. 47-1173004-00, "B&H Owners Group Valve Task Force Final Report," dated August 1988.
l
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- 6. _ BAH-10128,. Topical Report: TRAP 2 - FORTRAN Program for Digital Simulation ]
of the Transient Behavior of the Once-Through Steam Generator and l Associated Reactor Coolant System, dated Ap;ast 1976. {
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