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Category:INTERNAL OR EXTERNAL MEMORANDUM
MONTHYEARML20212H2031999-09-27027 September 1999 Forwards Operator Licensing Exams Administered at Monticello Nuclear Generating Plant During Wk of 990823.Encl Consists of Facility Submitted Outline & Initial Exam Submittal ML20216G4221999-09-27027 September 1999 Forwards NRC Operator Licensing Exam Rept 50-263/99-301 (Including Completed & Graded Tests) for Tests Administered During Wk of 990823 at Monticello Nuclear Generating Plant ML20212F5461999-09-23023 September 1999 Notification of 991004 Meeting with Utils in Rockville,Md to Update Status of Nuclear Mgt Company & Provide Details of Member Licensees Impending License Transfer Applications & Operating Agreement ML20211G9701999-08-30030 August 1999 Notification of 990909 Meeting with Util in Rockville,Md to Discuss Licensing Issues That May Result from Ongoing Merger Activities Between NSP & New Century Energies ML20137H5041999-04-0808 April 1999 Informs That Licensee Requesting Listed Changes to Boilerplate Distribution Lists Used by NRR for Docketed Info.Add Site General Manager to Both Prairie Island & Monticello Lists ML20202H6671999-02-0101 February 1999 Notification of 990223 Meeting with Util in Rockville,Md to Discuss Need for TR on Util Analytical Methods Used for Other NRC Licensees,Epri Schedule for Completion of CPM3/ Coretran Analysis Topical & Util Transition Plan ML20202H6321999-02-0101 February 1999 Notification of 990224 Meeting with Util in Rockville,Md to Discuss Implications of Util Recently Extended Commitment for Plant ITS Submittal,Nrc Initiative Toward risk-informed TSs & Interfacing ITS Conversion with Other Activities ML20154R3691998-10-19019 October 1998 Notification of 981029 Meeting with Listed Utils in Rockville,Md to Discuss Proposed Consortium of Utils ML20206S6521998-08-18018 August 1998 Informs That During 980708-10 ACRS 454th Meeting,Several Matters Were Discussed & Listed Repts & Letters Completed. Executive Director Also Authorized to Transmit Noted Memos ML20236U7851998-07-24024 July 1998 Informs That During 453rd & 454th Meetings of ACRS on 980603-05 & 0708-10,NRC Reviewed GE Nuclear Energy Program Associated W/Extended Power Uprates for Operating BWRs & Application for NSP for Power Level Increase for MNGP NUREG-1635, Informs That During 453rd Meeting on 980603-05,ACRS Discussed Several Matters & Completed Listed Repts & Ltr1998-07-0707 July 1998 Informs That During 453rd Meeting on 980603-05,ACRS Discussed Several Matters & Completed Listed Repts & Ltr ML20249A8041998-06-15015 June 1998 Notification of 980630 Meeting W/Util in Rockville,Md to Discuss Issues Related to Conversion to Improved Std TSs for Monticello Nuclear Generating Plant ML20248D7081998-05-26026 May 1998 Notification of 980604 Meeting W/Util in Rockville,Md to Discuss Proposed License Amend Supporting Monticello Power Uprate Program ML20216C3931998-05-11011 May 1998 Notification of 980521 Meeting W/Northern States Power Co & GE in Rockville,Md to Discuss Status of Staff Review of Proposed Power Uprate Program for Plant ML20217F2561998-03-20020 March 1998 Notification of 980330 Meeting W/Util to Discuss Licensee Response to Staff Request for Addl Info on Licensee Uprate Program.Meeting Will Be Held in Rockville,Md ML20198Q1881998-01-13013 January 1998 Forwards Nonproprietary Version of Montecello & Cooper Trip Rept to PDR ML20198N0251998-01-12012 January 1998 Discusses 971215-16 NRR Audit of Monticello Strainer Test. Tests Were Established to Develop Data for Strainer Design Installed in NPP ML20198G1911997-12-23023 December 1997 Notification of 980107 Meeting W/Util in Rockville,Md to Discuss Plans for Conversion to Improved STS for Plants ML20198R4951997-10-27027 October 1997 Notifies of 971030 Meeting W/Util in Rockville,Md to Discuss Licensee Operability Determinations for Sys & Components w/limited-scope Weld Insps & Licensee Plans for Submitting Formal Relief Requests Per 10CFR50.55a(g)(5)(iv) ML20216F9711997-09-0505 September 1997 Notification of 970911 Meeting W/Util in Rockville,Md to Discuss Current Issues Re Environ Qualification of Equipment at Plant ML20210T1371997-09-0303 September 1997 Notification of 970910 Meeting W/Ge & Southern Co in Rockville,Md to Discuss Status of Boiling Water Reactor Power Uprate Program ML20138J7671997-02-0505 February 1997 Informs That NRC Plans to Administer Generic Fundamentals Exam Section of Written Operator Licenseing Exam on 970409. Ltr W/Copy to Chief,Operator Licensing Branch Must Be Submitted to Listed Address in Order to Register Personnel ML20129B6031996-10-18018 October 1996 Notification of 961105 Meeting W/Util in Rockville,Md to Discuss Contents of NSP License Amend Request Supporting Plant Power Upgrade Program NUREG-1299, Forwards Continuation of Curtiss Papers to Be Filed Under Commission Correspondence in Pdr.Advanced Copy Sent to Pdr. List of Documents Included in Four Boxes Encl1994-06-29029 June 1994 Forwards Continuation of Curtiss Papers to Be Filed Under Commission Correspondence in Pdr.Advanced Copy Sent to Pdr. List of Documents Included in Four Boxes Encl ML20134B5061994-04-13013 April 1994 Submits Plants Which Will Be Discussed in Categories Indicated Re Results of Screening Meetings for June 1994. Partially Deleted ML20059H5891994-01-24024 January 1994 Notification of 940202 Meeting W/Util in Rockville,Md to Discuss Plans for Implementation of Rwl Mod as Required by NRC Bulletin 93-003 ML20057B4721993-09-15015 September 1993 Notification of 930929 Meeting W/Util in Rockville,Md to Discuss Installation of Water Level Monitoring Instrumentation at Plant ML20056E4571993-08-0505 August 1993 Forwards Technical Review Rept Re, Tardy Licensee Actions Initiated Because of Delayed Replacement of Batteries in Uninterruptible Power Supplies at Plant ML20127K6681992-11-20020 November 1992 Undated Memo Discussing Status of Field Erected Reactor Vessel Fabrication Review ML20055D5791990-06-27027 June 1990 Requests Position on Allowability of Radios or Tape Players in Control Room of Nonpower Reactors ML20155G7431988-06-0707 June 1988 Forwards F Miraglia 880527 Memo for Review & Requests Proposed Priorities for Actions on Project Manager Rept by C.O.B. 880609 ML20154Q1661988-05-27027 May 1988 Discusses Updating Project Managers Rept (Pmr) in Accordance W/New Priority Ranking Sys.Mods Have Been Made So That Pmr Will Now Accept New Priority Data.Old Priority Data Will Be Deleted During Wk of 880530.Sample Data Format Encl ML20148A7571988-03-14014 March 1988 Forwards Project Directorate III-1 Slides for 880317 Briefing of Executive Team.Slides Marked P Primary Slides Directorate Plans to Show.Other Slides Backup for Possible Ref ML20236P7451987-11-13013 November 1987 Reviews Latest Performance Indicators to Determine Whether Indicators Can Be Used to Ascertain Quality Performance.Five of Six Plants Achieving Very Good Quality Performance While One Plant Achieving Good Quality Performance ML20211N7961987-02-19019 February 1987 Requests Consideration of Encl Util 861223 Request That 860228 Application for Extension of Duration of Licenses DPR-57 & NPF-5 Be Given Higher Review Priority.Util Requests Completion of Review by 870331 ML20214C6861986-11-17017 November 1986 Forwards List of Missing SALP Evaluation Forms (0516B). Requests Review of Files to Locate Missing Forms.Recognizing That RP 0516B Issued in Mar 1986,request Applies Only to Repts Issued After Mar 1986 ML20210T3791986-10-0202 October 1986 Notification of 861017 Meeting W/Utils in Bethesda,Md to Discuss Issues Affecting Operating Reactors & NRC ML20211N8311986-10-0101 October 1986 Proposes Listed Schedule for Completion of Reviews of OLs Extensions for Listed Facilities Based on Low Priority of Effort.Plant Sys Branch Will Coordinate Responses & Recipients Will Be Provided W/Integrated Plant SERs ML20203H2741986-07-28028 July 1986 Forwards,For Review,Latest Update of SALP Ratings ML20206F1641986-06-21021 June 1986 Requests That Evaluations of Licensee Responses to Encl IE Bulletin 86-001 Re Min Flow Logic Problems That Could Disable RHR Pumps Not Be Closed Until Temporary Instruction for Guidance Issued ML20211E5131986-06-0606 June 1986 Discusses Inputs for SALP 6 Assessment for Dec 1984 - Mar 1986,due on 860618.Inputs Should Be Typed on 5520 Sys & Remain on Sys Until SALP Board Meeting Held.Listing of Insps Conducted During Assessment Period Encl ML20155F1941986-04-10010 April 1986 Summarizes Operating Reactor Events Meeting 86-11 on 860407 Re Events Since 860331.List of Attendees,Discussion of Events,Status of Assignments & Assigned Completion Dates for Items Encl.Response Requested for Incomplete Assignments ML20151Q9751986-01-29029 January 1986 Requests Identification of Div Contact for Regional Insp Team Leaders to Arrange NRR Alternative Shutdown & Fire Protection Reviewer Technical Assistance on region-based post-fire Safe Shutdown Insps.Schedule of Insps Submitted ML20198G2001985-11-0505 November 1985 Recommends Issuance of IE Info Notice Re Possible LOCA at High/Low Pressure Interface After Fire Damage Occurs in Control Room.Problem Discovered During Fire Protection re-review ML20087A8311984-03-0505 March 1984 Forwards Monticello Nuclear Power Plant Site-Specific Offsite Radiological Emergency Preparedness Evaluation & Monticello Nuclear Power Plant Full-Scale Joint Emergency Exercise on 830223, Final Rept ML20207L3491983-11-30030 November 1983 Advises That Scheme Described in Licensee 830929 Request for Extension of Date for Complying w/10CFR50.54 Unsatisfactory. Mods May Result in Shift Supervisor Spending Less Time in Control room.Davis-Besse Proposal Also Unacceptable ML20058G6051982-07-13013 July 1982 Forwards Draft Ltr Clarifying Confusion During LANL 820706-09 Site Visit to Collect Data for Vital Area Analysis Program.Concerns Re Releasing of Data Resolved.Future Visits Will Be Endowed W/Official Imprimatur ML20148F1051978-10-24024 October 1978 Forwards Memos Re Recent Problems in Pipe Support Base Plate design.(ANO:7811020332,7811020336, & 7811020343.) ML20148G2271978-10-24024 October 1978 Forwards 780929 Memo Re Results of Recent Fire Protec Res Test Conducted at Underwriters Lab. (See ANO: 7810050359, 7810050373.) ML20125A4051978-09-0101 September 1978 Responds to Bajwa 780831 Request for Review of Nonradiological Ets.Several Critical Inconsistencies Still Exist Between Fes Findings & ETS 1999-09-27
[Table view] Category:MEMORANDUMS-CORRESPONDENCE
MONTHYEARML20212H2031999-09-27027 September 1999 Forwards Operator Licensing Exams Administered at Monticello Nuclear Generating Plant During Wk of 990823.Encl Consists of Facility Submitted Outline & Initial Exam Submittal ML20216G4221999-09-27027 September 1999 Forwards NRC Operator Licensing Exam Rept 50-263/99-301 (Including Completed & Graded Tests) for Tests Administered During Wk of 990823 at Monticello Nuclear Generating Plant ML20212F5461999-09-23023 September 1999 Notification of 991004 Meeting with Utils in Rockville,Md to Update Status of Nuclear Mgt Company & Provide Details of Member Licensees Impending License Transfer Applications & Operating Agreement ML20211G9701999-08-30030 August 1999 Notification of 990909 Meeting with Util in Rockville,Md to Discuss Licensing Issues That May Result from Ongoing Merger Activities Between NSP & New Century Energies ML20137H5041999-04-0808 April 1999 Informs That Licensee Requesting Listed Changes to Boilerplate Distribution Lists Used by NRR for Docketed Info.Add Site General Manager to Both Prairie Island & Monticello Lists ML20202H6321999-02-0101 February 1999 Notification of 990224 Meeting with Util in Rockville,Md to Discuss Implications of Util Recently Extended Commitment for Plant ITS Submittal,Nrc Initiative Toward risk-informed TSs & Interfacing ITS Conversion with Other Activities ML20202H6671999-02-0101 February 1999 Notification of 990223 Meeting with Util in Rockville,Md to Discuss Need for TR on Util Analytical Methods Used for Other NRC Licensees,Epri Schedule for Completion of CPM3/ Coretran Analysis Topical & Util Transition Plan ML20154R3691998-10-19019 October 1998 Notification of 981029 Meeting with Listed Utils in Rockville,Md to Discuss Proposed Consortium of Utils ML20206S6521998-08-18018 August 1998 Informs That During 980708-10 ACRS 454th Meeting,Several Matters Were Discussed & Listed Repts & Letters Completed. Executive Director Also Authorized to Transmit Noted Memos ML20236U7851998-07-24024 July 1998 Informs That During 453rd & 454th Meetings of ACRS on 980603-05 & 0708-10,NRC Reviewed GE Nuclear Energy Program Associated W/Extended Power Uprates for Operating BWRs & Application for NSP for Power Level Increase for MNGP NUREG-1635, Informs That During 453rd Meeting on 980603-05,ACRS Discussed Several Matters & Completed Listed Repts & Ltr1998-07-0707 July 1998 Informs That During 453rd Meeting on 980603-05,ACRS Discussed Several Matters & Completed Listed Repts & Ltr ML20249A8041998-06-15015 June 1998 Notification of 980630 Meeting W/Util in Rockville,Md to Discuss Issues Related to Conversion to Improved Std TSs for Monticello Nuclear Generating Plant ML20248D7081998-05-26026 May 1998 Notification of 980604 Meeting W/Util in Rockville,Md to Discuss Proposed License Amend Supporting Monticello Power Uprate Program ML20216C3931998-05-11011 May 1998 Notification of 980521 Meeting W/Northern States Power Co & GE in Rockville,Md to Discuss Status of Staff Review of Proposed Power Uprate Program for Plant ML20217F2561998-03-20020 March 1998 Notification of 980330 Meeting W/Util to Discuss Licensee Response to Staff Request for Addl Info on Licensee Uprate Program.Meeting Will Be Held in Rockville,Md ML20198Q1881998-01-13013 January 1998 Forwards Nonproprietary Version of Montecello & Cooper Trip Rept to PDR ML20198N0251998-01-12012 January 1998 Discusses 971215-16 NRR Audit of Monticello Strainer Test. Tests Were Established to Develop Data for Strainer Design Installed in NPP ML20198G1911997-12-23023 December 1997 Notification of 980107 Meeting W/Util in Rockville,Md to Discuss Plans for Conversion to Improved STS for Plants ML20198R4951997-10-27027 October 1997 Notifies of 971030 Meeting W/Util in Rockville,Md to Discuss Licensee Operability Determinations for Sys & Components w/limited-scope Weld Insps & Licensee Plans for Submitting Formal Relief Requests Per 10CFR50.55a(g)(5)(iv) ML20216F9711997-09-0505 September 1997 Notification of 970911 Meeting W/Util in Rockville,Md to Discuss Current Issues Re Environ Qualification of Equipment at Plant ML20210T1371997-09-0303 September 1997 Notification of 970910 Meeting W/Ge & Southern Co in Rockville,Md to Discuss Status of Boiling Water Reactor Power Uprate Program ML20138J7671997-02-0505 February 1997 Informs That NRC Plans to Administer Generic Fundamentals Exam Section of Written Operator Licenseing Exam on 970409. Ltr W/Copy to Chief,Operator Licensing Branch Must Be Submitted to Listed Address in Order to Register Personnel ML20129B6031996-10-18018 October 1996 Notification of 961105 Meeting W/Util in Rockville,Md to Discuss Contents of NSP License Amend Request Supporting Plant Power Upgrade Program NUREG-1299, Forwards Continuation of Curtiss Papers to Be Filed Under Commission Correspondence in Pdr.Advanced Copy Sent to Pdr. List of Documents Included in Four Boxes Encl1994-06-29029 June 1994 Forwards Continuation of Curtiss Papers to Be Filed Under Commission Correspondence in Pdr.Advanced Copy Sent to Pdr. List of Documents Included in Four Boxes Encl ML20134B5061994-04-13013 April 1994 Submits Plants Which Will Be Discussed in Categories Indicated Re Results of Screening Meetings for June 1994. Partially Deleted ML20059H5891994-01-24024 January 1994 Notification of 940202 Meeting W/Util in Rockville,Md to Discuss Plans for Implementation of Rwl Mod as Required by NRC Bulletin 93-003 ML20057B4721993-09-15015 September 1993 Notification of 930929 Meeting W/Util in Rockville,Md to Discuss Installation of Water Level Monitoring Instrumentation at Plant ML20056E4571993-08-0505 August 1993 Forwards Technical Review Rept Re, Tardy Licensee Actions Initiated Because of Delayed Replacement of Batteries in Uninterruptible Power Supplies at Plant ML20127K6681992-11-20020 November 1992 Undated Memo Discussing Status of Field Erected Reactor Vessel Fabrication Review ML20055D5791990-06-27027 June 1990 Requests Position on Allowability of Radios or Tape Players in Control Room of Nonpower Reactors ML20155G7431988-06-0707 June 1988 Forwards F Miraglia 880527 Memo for Review & Requests Proposed Priorities for Actions on Project Manager Rept by C.O.B. 880609 ML20154Q1661988-05-27027 May 1988 Discusses Updating Project Managers Rept (Pmr) in Accordance W/New Priority Ranking Sys.Mods Have Been Made So That Pmr Will Now Accept New Priority Data.Old Priority Data Will Be Deleted During Wk of 880530.Sample Data Format Encl ML20148A7571988-03-14014 March 1988 Forwards Project Directorate III-1 Slides for 880317 Briefing of Executive Team.Slides Marked P Primary Slides Directorate Plans to Show.Other Slides Backup for Possible Ref ML20236P7451987-11-13013 November 1987 Reviews Latest Performance Indicators to Determine Whether Indicators Can Be Used to Ascertain Quality Performance.Five of Six Plants Achieving Very Good Quality Performance While One Plant Achieving Good Quality Performance ML20215L3101987-05-0707 May 1987 Staff Requirements Memo Re Commission 870430 Affirmation/ Discussion & Vote in Washington,Dc on SECY-87-68A Concerning Order to Rescind 861020 Order Directing Licensee to Show Why OL Should Not Be Modified.Order Signed on 870501 ML20211N7961987-02-19019 February 1987 Requests Consideration of Encl Util 861223 Request That 860228 Application for Extension of Duration of Licenses DPR-57 & NPF-5 Be Given Higher Review Priority.Util Requests Completion of Review by 870331 ML20214C6861986-11-17017 November 1986 Forwards List of Missing SALP Evaluation Forms (0516B). Requests Review of Files to Locate Missing Forms.Recognizing That RP 0516B Issued in Mar 1986,request Applies Only to Repts Issued After Mar 1986 ML20210T3791986-10-0202 October 1986 Notification of 861017 Meeting W/Utils in Bethesda,Md to Discuss Issues Affecting Operating Reactors & NRC ML20211N8311986-10-0101 October 1986 Proposes Listed Schedule for Completion of Reviews of OLs Extensions for Listed Facilities Based on Low Priority of Effort.Plant Sys Branch Will Coordinate Responses & Recipients Will Be Provided W/Integrated Plant SERs ML20203H2741986-07-28028 July 1986 Forwards,For Review,Latest Update of SALP Ratings ML20206F1641986-06-21021 June 1986 Requests That Evaluations of Licensee Responses to Encl IE Bulletin 86-001 Re Min Flow Logic Problems That Could Disable RHR Pumps Not Be Closed Until Temporary Instruction for Guidance Issued ML20211E5131986-06-0606 June 1986 Discusses Inputs for SALP 6 Assessment for Dec 1984 - Mar 1986,due on 860618.Inputs Should Be Typed on 5520 Sys & Remain on Sys Until SALP Board Meeting Held.Listing of Insps Conducted During Assessment Period Encl ML20155F1941986-04-10010 April 1986 Summarizes Operating Reactor Events Meeting 86-11 on 860407 Re Events Since 860331.List of Attendees,Discussion of Events,Status of Assignments & Assigned Completion Dates for Items Encl.Response Requested for Incomplete Assignments ML20151Q9751986-01-29029 January 1986 Requests Identification of Div Contact for Regional Insp Team Leaders to Arrange NRR Alternative Shutdown & Fire Protection Reviewer Technical Assistance on region-based post-fire Safe Shutdown Insps.Schedule of Insps Submitted ML20198G2001985-11-0505 November 1985 Recommends Issuance of IE Info Notice Re Possible LOCA at High/Low Pressure Interface After Fire Damage Occurs in Control Room.Problem Discovered During Fire Protection re-review ML20087A8311984-03-0505 March 1984 Forwards Monticello Nuclear Power Plant Site-Specific Offsite Radiological Emergency Preparedness Evaluation & Monticello Nuclear Power Plant Full-Scale Joint Emergency Exercise on 830223, Final Rept ML20207L3491983-11-30030 November 1983 Advises That Scheme Described in Licensee 830929 Request for Extension of Date for Complying w/10CFR50.54 Unsatisfactory. Mods May Result in Shift Supervisor Spending Less Time in Control room.Davis-Besse Proposal Also Unacceptable ML20058G6051982-07-13013 July 1982 Forwards Draft Ltr Clarifying Confusion During LANL 820706-09 Site Visit to Collect Data for Vital Area Analysis Program.Concerns Re Releasing of Data Resolved.Future Visits Will Be Endowed W/Official Imprimatur ML20148F1051978-10-24024 October 1978 Forwards Memos Re Recent Problems in Pipe Support Base Plate design.(ANO:7811020332,7811020336, & 7811020343.) ML20148G2271978-10-24024 October 1978 Forwards 780929 Memo Re Results of Recent Fire Protec Res Test Conducted at Underwriters Lab. (See ANO: 7810050359, 7810050373.) 1999-09-27
[Table view] |
Text
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wasmucros o c. som g 3 1972 Files (Docket No. 50-263)
THRU: Z mann, Chief, ORB #2, DRL Me MONTICELLO NUCLE POWER PLANT - ECCS CONFORMANCE TO AEC INTERIM DESIGN CRITERIA Summarl To continue our evaluation of the Monticello "ECCS Conformance to New AEC Adopted Interim Acceptance Criteria" (NSP report dated September 21, 1971), we require further explanation:
- 1. of the re@c.4 time f rom 8 minutes to 3 minutes to cover the core by the accumulation of core spray water.
- 2. of the reduced flood time for equivalent core spray conditions following the DBA.
- 3. of the peak clad temperature discrepancies within the report for the 0.1 ft2 break.
- 4. of the maximum recirculation line break described in the FSAR as 5.6 ft2 and in the September 1971 report prepared by GE as 4.2 ft2
- 5. of the automatic relief system reliability with regard to operation of two or all three of the Automatic Pressure Relief (APR) valves following small primary system breaks.
Discussion By letter dated September 21, 1971, Northern States Power Company (NSP) submitted an analysis by General Electric Company confirming that the performance of the Monticello ECCS is in compliance with the AEC Interim Design Criteria.
According to the FSAR, the core is protected from excessive temperatures following loss of coolant by two independent full capacity core spray systems that draw water f rom the suppression pool and spray water directly v4 9212030598 DR 720503 ADOCK 05000263 PDR
s'. ,
Files MAY 3 1972 on the core. Each system, rated at 3250 gpm with 165 psig pump dis-charge pressure, starts to deliver when the injection valve opens at 450 psig reactor vessel pressure. The core spray pumps are electrically driven. Either of the core spray systems when operated in conjunction with either the automatic pressure relief or High Pressure Coolant Injection (HPCI) system will, according to the GE reanalysis, limit core clad temperatures to less than 2300*F for various postulated loss-of-coolant accidents f rom the smallest detectable leak to the largest pipe break, the complete double-ended severance of a recirculation loop pipe.
The HPCI system is described as a system designed to pump 3000 gpm into the reactor vessel through the feedwater inlet within a reactor pressure range of about 1125 psig to 150 psig under loss of coolant conditions which do not result in rapid depressurization of the pressure vessel.
The single HPCI pump is driven by a steam turbine with steam f rom the reactor v asel. Initial core cooling water is supplied from the condensate storage tank. When this source of water is exhausted, the pump suction transfers automatically to the suppression pool. Operation of the HPCI system is dependent upon reactor water level signals. Either
" low" reactor water level or "high" drywell pressure starts the system and "high" reactor water level will stop it. Operation of the HPCI system is completely independent of ac power and requires only de power f rom the plant de batteries systems for operation. GE has assumed in reanalysis of the ECCS, as they have for other similar power plants, that the HPCI is not availabic af ter the LOCA.
The automatic pressure relief system accomplishes reactor vessel depres-surization by blowdown through automatic opening of the relief valves which vent steam to the suppression pool. For small breaks, the vessel is depressurized in sufficient time to allow either the core spray subsystem or Low Pressure Coolant Injection (LPCI) system to provide adequate cooling to prevent any clad melting
Two of three relief valves, it is reported in the FSAR, discharge more 3
- The original design basis for the auto relief system was, in conjunction with core spray or low pressure coolant injection, to prevent clad melting for all break sizes (FSAR 6-2.39 and 6-2.3). Two of three j relief valves will reduce the pressure in time to permit rated core
- spray flow before clad melting occurs, t
l
Files MAY 3 1972 than 1.6 x 106 pe.inds of steam per hour at 1125 psig, thereby providing a satisfactory backup for the HPCI system. Automatic actuation requires coincident indication of reactor water " low-low" level and drywell "high" pressure. An additional interlock is provided to assure that at least one of four Residual Heat Removal (RHR) (LPCI) or one of two core spray pumps is delivering output pressure. A time delay provides time for the HPCI or feedwater system to restore the proper reactor vessel coolant level before blowdown activation. The GE reanalysis of the ECCS, in accordance with AEC Interim Policy Statement, assumes automatic depressurization for the small breaks less than 0.1 f t2, The LPCI subsystem, an integral part of the RHR, is designed to deliver
?
water to the recirculation loops at the rate of 4000 gpm for each of four pumps when the reactor vessel prescure is 20 psi above the suppression chamber pressure or 2000 gpm when reactor vessel pressure l 1s 262 psi above the suppression chamber pressure (FSAR 6-2.11). The objective of the LPCI system is to restore and maintain the coolant inventory in the reactor vessel af ter a loss-of-coolant accident by pumping water from the suppression pool to the reactor vessel via the coolant recirculation .aops. The LPCI system is designed to provide reactor core cooling for a large spectrum of loss-of-coolant accidents completely independent of the reactor core spray cooling systems.
Evaluation Peak clad temperatures have been calculated by the General Electric Company for the complete spectrum of break sizes in the pressurized core cooling system up to and including the design basis double-ended recirculation line break. Using the calculational methods and assumptions that have previously been reviewed and accepted by DRL, peak clad temperatures remain below 2300*F for breaks smaller than 0.1 ft2 with automatic depressurization followed by core cooling with both spray systems or one core spray system and low pressure coolant
- inj ection by two pumps. For breaks larger than 0.1 ft2 but smaller than the 4.2 f t2 (DBA double-ended recirculation line break), either two core spray systems or one core spray system and low pressure coolant
, injection by two pumps is sufficient to limit peak clad temperatures
' to less than 2300*F. The active fuel cladding metal water reaction is less than or about 0.1% for all break sizes, acceptably below the 1%
value specified in the AEC Interim Design Criteria. Selected examples of assumed primary coolant breaks show that peak clad temperature
i
(~
4 Files May 3 B72 i
i transients endure above the normal operating 600-650'F temperature i for as long as 350-400 seconds and that the temperatures in some cases remain above 1500'F but below 2300'F for as long as 220 seconds, in our opinion, this is insufficient time, temperature, or volume of clad
' to cause core damage as a result of clad embrittlement that could interfere with the emergency core cooling process.
We have observed that the calculated clad temperature increases, following all LOCAs up to and including the double-ended break of the recirculation l
piping, are arrested by core flooding prior to reaching 2300*F at any j location in the core. The following observations require further eval-uation by DRL:
i
- 1. Figures 11,16,17 and 18 of the subject report show core flooding f rom two core spray systems following the design basis LOCA within approximately 175 seconds, whereas the FSAR states "the accumulation of core spray water in the bottom of the vessel covers the core in about 8 minutes for the maximum design break". There appears to be a time 4
discrepancy.
- 2. Figures 11,16,17 and 18 of the subject report show core flooding following the design basis LOCA within approximately 175 seconds, a period of time that is noticeably shorter than the 215 seconds previously reported in FSAR Figure 6-2.25
- and 185 seconds in FSAR Figure 6-2.49. According to the most recent calculations, core flooding time appears to be 10-40 seconds f aster than for comparable conditions reported in the FSAR.
i
- 3. Figures 13,14 and 15 of the subject report show peak temper-atures that are 400-450*F higher than the temperature reported i in Figure 1 for 0.1 ft primary 2
coolant system break. There appears to be an inconsistency in reporting the peak clad temperatures calculated by GE in accordance with the AEC Interim Policy Statement on Emergency Core Cooling Systems.
- 4. The DBA primary system break while not identified numerically in the subject _ report is estimated f rom Figure 1 to be about-4.2 ft2 This value is in agreement with some of the references in the FSAR but not with Figure 6-2.1 of the FSAR which shows that the maximum recirculation line break size is 5.6 ft2 Resolution of this apparent difference in the i
a Z
i
Files MAY 3 1972 size of the DBA is important because extrapolation of the Figure 1 results to the larger break size results in peak clad temperatures in excess of 2300*F, the limit specified by the AEC Interim Acceptance Criteria for ECCS for light water reactors.
- 5. Table I and Figure 1 of the subject report show that peak clad temperatures occur following a 0.05 f t2 break which is now defined as an intermediate break. When compared with Figure 6-2.34 in the FSAR, it is evident that the reanalysis has resulted in a significant shif t of the peak clad temperature from the approximately 0.15 ft 2 break to
- 0.05 ft2 break. Intermediate and large breaks, i.e. , breaks
- previously defined as a break size greater than 0.1 ft2, required no HPCI or auto depressurization. The peak temper-ature based on the recent GE calculations now occurs in the region formerly described as the small break range of breaks (i.e., 4.0.1 ft2) and HPCI or auto depressurization is a pre-d requisite to satisf actory cooling by LPCl or CSCS. Since it is assumed in the GE analysis that the HPCI is inoperative,
- depressurization in time for effective core cooling by spray and flooding must be accomplished by the Automatic Pressure Relief System. Figure 3 shows the peak clad temperature and Figure 10 shows the reactor vessel water level following a 0.05 ft2 break as a function of time. We cannot determine whether two of the three automatic relief system valves were assumed to open to discharge 1.6 x 106 pounds of steam per hour at 1125 psig as described by the FSAR or if all three valves were assumed to open to depressurize in time for acceptable emergency core cooling; nor can we explain the transient increase in reactor vessel water level at about 180 seconds af ter the break or verify that the two-minute i
time delay in opening the automatic pressure relief valves af ter sensing high drywell pressure and low reactor vessel water has been properly considered by GE.
W" f , 0 1
Files SY 31972 During telecons on April 7, 13 and 14, 1972, NSP was requested to pro-vide responses to the above concerns so that our evaluation of the Monticello "ECCS Conformance to the New AEC Adopted Interim Acceptance Criteria" can be completed.
,_ w James J . Shea Operating Reactors Branch #2 Division of Reactor Licensing cc: DJSkovholt, DRL TJCarter, DRL DLZiemann, DRL JJShea, DRL RMDiggs, DRL t