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Category:CORRESPONDENCE-LETTERS
MONTHYEARML20217M3801999-10-21021 October 1999 Forwards Insp Rept 50-263/99-06 on 990813-0923.Four Violations of NRC Requirements Identified & Being Treated as non-cited Violations Consistent with App C of Enforcement Policy ML20217G0711999-10-13013 October 1999 Forwards Insp Rept 50-263/99-12 on 990913-17.No Violations Noted ML20216J2491999-09-30030 September 1999 Ack Receipt of 980804,990626 & 0720 Ltrs in Response to GL 98-01,suppl 1, Year 2000 Readiness of Computer Sys at Npps. Staff Review Has Concluded That All Requested Info Has Been Provided ML20217B1421999-09-30030 September 1999 Informs That on 990902,NRC Staff Completed mid-cicle Plant Performance Review of Monticello Nuclear Generating Station. Staff Conducted Reviews for All Operating NPPs to Integrate Performance Information & to Plan for Insp Activities ML20212K9131999-09-30030 September 1999 Refers to 990920 Meeting Conducted at Monticello Nuclear Generating Station to Discuss Initiatives in Risk Area & to Establish Dialog Between SRAs & PRA ML20216J8091999-09-24024 September 1999 Informs That New Diaphragm Matl Has Corrected Sticking Problem Associated with Increased Control Rod Drive Scram Times.Augmented Testing of Valves at Monticello Has Been Discontinued ML20216G4341999-09-24024 September 1999 Forwards Exam Rept 50-263/99-301 on 990823-26.Violation Occurred & Being Treated as Ncv,Consistent with App C of Enforcement Policy.Test Was Administered to Two Applicants. Both Applicants Passed All Sections of Exam ML20212G7171999-09-24024 September 1999 Submits Semiannual Status Update on Project Plans for USAR Review Project & Conversion to Its.Conversion Package Submittal Continues to Be Targeted for Aug of 2000 ML20212G9801999-09-23023 September 1999 Refers to Resolution of Unresolved Items Identified Re Security Alarm Station Operations at Both Monitcello & Prairie Island ML20212F0901999-09-21021 September 1999 Confirms Discussion Between M Hammer & Rd Lanksbury to Have Routine Mgt Meeting on 991005 in Lisle,Il.Purpose of Meeting to Discuss Improvement Initiatives in Areas of Operations & Equipment Reliability ML20212A9761999-09-0909 September 1999 Submits 1999 Annual Rept of Any Changes or Errors Identified in ECCS Analytical Models or Applications ML20217A5751999-09-0909 September 1999 Forwards Individual Exam Results for Licensee Applicants Who Took Aug 1999 Initial License Exam.Without Encls ML20211Q6981999-09-0606 September 1999 Informs That NRC Tentatively Scheduled Initial Licensing Exam for Monticello Operator License Applicants During Wks of 010604 & 11.Validation of Exam Will Occur at Station During Wk of 010514 ML20211L1981999-09-0101 September 1999 Forwards Insp Rept 50-263/99-05 on 990702-0812.No Violations Noted ML20211K7971999-09-0101 September 1999 Informs That Util Reviewed Rvid as Requested in NRC .Recommended Corrections Are Listed ML20211K2591999-08-27027 August 1999 Forwards NSP Co Fitness for Duty Program Performance Data for Six Month Period Ending 990630 ML20211F9961999-08-26026 August 1999 Forwards Effluent & Waste Disposal Semi-Annual Rept for 990101-990630, Revised Effluent & Waste Disposal Semi-Annual Rept for 980701-981231 & Revs to ODCM for Monitcello Nuclear Generating Plant ML20211C9501999-08-23023 August 1999 Forwards Rev 17 to Monticello Nuclear Generating Plant USAR, Updating Info in USAR to Reflect Implementation of Increase in Licensed Core Thermal Power from 1,670 Mwt to 1,775 Mwt.Rept of Changes,Tests & Experiments Not Included ML20210U1831999-08-12012 August 1999 Revises 980202 Commitment Re GL 96-06, Assurance of Equipment Operability & Containment Integrity During Design- Basis Accident Conditions ML20210T9601999-08-12012 August 1999 Provides Rept on Status of Util RPV Feedwater Nozzle Insps Performed in Response to USI A-10 Re BWR Nozzle Cracking ML20210Q0341999-08-0404 August 1999 Forwards SE Granting Licensee 980724 Relief Request 10 Re Third 10-year Interval ISI Program Plan,Entitled, Limited Exam ML20210H0861999-07-28028 July 1999 Forwards Insp Rept 50-263/99-04 on 990521-0701.No Violations Noted.Licensee Conduct at Monticello Facility Characterized by safety-conscious Operations,Sound Engineering & Maint Practices & Appropriate Radiological Controls ML18107A7051999-07-20020 July 1999 Provides Suppl Info Which Supersedes Info in 990625 Ltr in Response to NRC RAI Re GL 98-01, Y2K Readiness of Computer Sys at Nuclear Power Plants. ML20212H3191999-07-16016 July 1999 Forwards Aug 1999 Monticello RO Exam Package,Including Revised Outlines.All Changes Are in Blue Font ML20209G5621999-07-14014 July 1999 Forwards Insp Rept 50-263/99-11 on 990621-24.No Violations Noted.Objective of Insp,To Determine Whether Monticello Nuclear Generating Station Emergency Plan Adequate & If Station Personnel Properly Implemented Emergency Plan ML20196J5351999-07-0202 July 1999 Discusses GL 92-01,Rev 1,Supp 1, Rv Integrity, Issued by NRC on 950515 & NSP Responses & 980917 for Monticello Npp.Informs That Staff Revised Info in Rvid & Released Info as Rvid Version 2 ML20196J9681999-07-0101 July 1999 Informs That in Sept 1998,Region III Received Rev 20 to Portions of Util Emergency Plan Under 10CFR50.54(q).Based on Determination That Changes Do Not Decrease Effectiveness of Licensee Emergency Plan,No NRC Approval Required ML20209B6151999-06-25025 June 1999 Responds to NRC Request for Info Re Y2K Readiness at Nuclear Power Plants.Gl 98-01 Requested Response on Status of Facility Y2K Readiness by 990701.Y2K Readiness Disclosure Attached ML20196H2291999-06-24024 June 1999 Responds to Administrative Ltr 99-02,dtd 990603,requesting Licensee to Provide Estimate of Licensing Action Submittals Anticipated.Four New Submittals Per Year Are Anticipated ML20207D5851999-05-25025 May 1999 Submits Info Re Partial Fulfillment of License Conditions Placed on Amend 101,which Approved Use of Ten Exceptions for 24 Months Subject to Listed App C Conditions.Util Will Submit Second Rept to Obtain Approval for Continued Use ML20206S0911999-05-17017 May 1999 Forwards Response to NRC 990324 RAI Re Proposed Amend to pressure-temp Limits & Surveillance Capsule Withdrawal Schedule, .Supporting Calculations Also Encl ML20206N5601999-05-13013 May 1999 Informs That NRC Office of Nuclear Reactor Regulation Reorganized Effective 990328.As Part of Reorganization,Div of Licensing Project Mgt Created.Cm Craig Will Be Section Chief for Monticello Npp.Organization Chart Encl ML20206G2181999-05-0505 May 1999 Discusses Completion of Licensing Action for GL 96-01, Testing of Safety-Related Logic Circuits, Dtd 960110,for Plant ML20206G4901999-05-0404 May 1999 Forwards Staff Review of Licensee 960508 Response to NRC Bulletin 96-002, Movement of Heavy Loads Over Sf,Over Fuel in Rc or Over Safety-Related Equipment, .Overall, Responses Acceptable.Tac M95610 Closed ML20206G7741999-05-0303 May 1999 Forwards Insp Rept 50-263/99-02 on 990223-0408.One Violation Occurred & Being Treated as non-cited Violation,Consistent with App C of Enforcement Policy ML20206D1651999-04-27027 April 1999 Forwards Radiation Environ Monitoring Program for MNGP for Jan-Dec 1998, Per Plant TS 6.7.C.1.Ltr Contains No New NRC Commitments or Modifies Any Prior Commitments ML20205N0821999-04-12012 April 1999 Forwards SE of NSP Response to NRC GL 95-07, Pressure Locking & Thermal Binding of Safety-Related Power-Operated Gate Valves. Licensee Adequately Addressed Actions Requested in GL ML20205N4811999-04-0909 April 1999 Forwards Licensing Requalification Insp Rept 50-263/99-10 on 990308-12.No Violations Noted.However,Inspectors Through Observation of Simulator Scenario Exams Noted Difficulties in Ability of SM to Simultaneously Implement Duties of SM ML20205N5301999-04-0909 April 1999 Discusses Arrangements Made on 990406 for Administration of Licensing Exams at Monticello Nuclear Generating Station During Wk of 990823.Requests That Exam Outlines Be Submitted by 990128 & Supporting Ref Matls by 990719 ML20196K7831999-03-31031 March 1999 Forwards Decommissioning Funding Status Rept for Monticello & Prairie Island Nuclear Generating Plants,Per Requirements of 10CFR50.75(f)(1) ML20205H5731999-03-29029 March 1999 Submits Required 1998 Actual & 1999 Projected Cash Flow Statements for Monticello Nuclear Generating Plant & PINGP, Units 1 & 2.Encl Contains Proprietary Info.Proprietary Info Withheld,Per 10CFR2.790(b)(1) ML20205C4851999-03-26026 March 1999 Informs That on 990203,NRC Staff Completed PPR of Nuclear Plant.Staff Conducts Reviews for All Operating NPPs to Develop an Integrated Understanding of Safety Performance ML20205C6561999-03-26026 March 1999 Submits Semiannual Update on Project Plans for USAR Review Project & Conversion to ITS ML20205A5881999-03-24024 March 1999 Forwards Request for Addl Info Re Submittal Requesting Rev of pressure-temperature Limits & Surveillance Capsule Withdrawal Schedule ML20204H4711999-03-18018 March 1999 Forwards SER Concluding That Util Established Acceptable Program to Verify Periodically design-basis Capability of safety-related MOVs at Monticello & Adequately Addressed Actions Requested in GL 96-05 ML20207H5161999-03-11011 March 1999 Forwards Insp Rept 50-263/99-01 on 990112-0222.No Violations Noted ML20207F4091999-02-28028 February 1999 Forwards Fitness for Duty Program Performance Data for Six Month Period from 980701-981231,IAW 10CFR26.71 ML20207F6741999-02-24024 February 1999 Forwards Summary of Nuclear Property Insurance Maintained at Monticello & Prairie Island Nuclear Generating Plants ML20207F6901999-02-23023 February 1999 Forwards Effluent & Waste Disposal Semi-Annual Rept for 980701-981231, Off-Site Radiation Dose Assessment for 980101-981231 & Revised Effluent & Waste Disposal Semi- Annual Rept for 980101-980630, for Monticello ML20203F2541999-02-10010 February 1999 Informs That Beginning 990216,DE Hills Will Be Chief of Operations Branch Which Includes Operator Licensing Function 1999-09-09
[Table view] Category:INCOMING CORRESPONDENCE
MONTHYEARML20216J8091999-09-24024 September 1999 Informs That New Diaphragm Matl Has Corrected Sticking Problem Associated with Increased Control Rod Drive Scram Times.Augmented Testing of Valves at Monticello Has Been Discontinued ML20212G7171999-09-24024 September 1999 Submits Semiannual Status Update on Project Plans for USAR Review Project & Conversion to Its.Conversion Package Submittal Continues to Be Targeted for Aug of 2000 ML20212A9761999-09-0909 September 1999 Submits 1999 Annual Rept of Any Changes or Errors Identified in ECCS Analytical Models or Applications ML20211K7971999-09-0101 September 1999 Informs That Util Reviewed Rvid as Requested in NRC .Recommended Corrections Are Listed ML20211K2591999-08-27027 August 1999 Forwards NSP Co Fitness for Duty Program Performance Data for Six Month Period Ending 990630 ML20211F9961999-08-26026 August 1999 Forwards Effluent & Waste Disposal Semi-Annual Rept for 990101-990630, Revised Effluent & Waste Disposal Semi-Annual Rept for 980701-981231 & Revs to ODCM for Monitcello Nuclear Generating Plant ML20211C9501999-08-23023 August 1999 Forwards Rev 17 to Monticello Nuclear Generating Plant USAR, Updating Info in USAR to Reflect Implementation of Increase in Licensed Core Thermal Power from 1,670 Mwt to 1,775 Mwt.Rept of Changes,Tests & Experiments Not Included ML20210U1831999-08-12012 August 1999 Revises 980202 Commitment Re GL 96-06, Assurance of Equipment Operability & Containment Integrity During Design- Basis Accident Conditions ML20210T9601999-08-12012 August 1999 Provides Rept on Status of Util RPV Feedwater Nozzle Insps Performed in Response to USI A-10 Re BWR Nozzle Cracking ML18107A7051999-07-20020 July 1999 Provides Suppl Info Which Supersedes Info in 990625 Ltr in Response to NRC RAI Re GL 98-01, Y2K Readiness of Computer Sys at Nuclear Power Plants. ML20212H3191999-07-16016 July 1999 Forwards Aug 1999 Monticello RO Exam Package,Including Revised Outlines.All Changes Are in Blue Font ML20209B6151999-06-25025 June 1999 Responds to NRC Request for Info Re Y2K Readiness at Nuclear Power Plants.Gl 98-01 Requested Response on Status of Facility Y2K Readiness by 990701.Y2K Readiness Disclosure Attached ML20196H2291999-06-24024 June 1999 Responds to Administrative Ltr 99-02,dtd 990603,requesting Licensee to Provide Estimate of Licensing Action Submittals Anticipated.Four New Submittals Per Year Are Anticipated ML20207D5851999-05-25025 May 1999 Submits Info Re Partial Fulfillment of License Conditions Placed on Amend 101,which Approved Use of Ten Exceptions for 24 Months Subject to Listed App C Conditions.Util Will Submit Second Rept to Obtain Approval for Continued Use ML20206S0911999-05-17017 May 1999 Forwards Response to NRC 990324 RAI Re Proposed Amend to pressure-temp Limits & Surveillance Capsule Withdrawal Schedule, .Supporting Calculations Also Encl ML20206D1651999-04-27027 April 1999 Forwards Radiation Environ Monitoring Program for MNGP for Jan-Dec 1998, Per Plant TS 6.7.C.1.Ltr Contains No New NRC Commitments or Modifies Any Prior Commitments ML20196K7831999-03-31031 March 1999 Forwards Decommissioning Funding Status Rept for Monticello & Prairie Island Nuclear Generating Plants,Per Requirements of 10CFR50.75(f)(1) ML20205H5731999-03-29029 March 1999 Submits Required 1998 Actual & 1999 Projected Cash Flow Statements for Monticello Nuclear Generating Plant & PINGP, Units 1 & 2.Encl Contains Proprietary Info.Proprietary Info Withheld,Per 10CFR2.790(b)(1) ML20205C6561999-03-26026 March 1999 Submits Semiannual Update on Project Plans for USAR Review Project & Conversion to ITS ML20207F4091999-02-28028 February 1999 Forwards Fitness for Duty Program Performance Data for Six Month Period from 980701-981231,IAW 10CFR26.71 ML20207F6741999-02-24024 February 1999 Forwards Summary of Nuclear Property Insurance Maintained at Monticello & Prairie Island Nuclear Generating Plants ML20207F6901999-02-23023 February 1999 Forwards Effluent & Waste Disposal Semi-Annual Rept for 980701-981231, Off-Site Radiation Dose Assessment for 980101-981231 & Revised Effluent & Waste Disposal Semi- Annual Rept for 980101-980630, for Monticello ML20203A3081999-01-28028 January 1999 Forwards TS Page 60d,as Supplement 3 to 971125 LAR Re CST Low Level Hpci/Rcic Suction Transfer.Page Includes NRC Approved Amend 103 Changes for Use by NRC in Issuing SER ML20202F7821999-01-27027 January 1999 Forwards 1999 Four Year Simulator Certification Rept for MNGP Simulation Facility, Per 10CFR55.45(b)(5)(ii) & 10CFR55.45(b)(5)(vi).Ltr Contains No New Commitments or Modifies Any Prior Commitments ML20206S0331999-01-20020 January 1999 Submits Annual Rept of Safety & Relief Valves Failure & Challenges ML20206P1221998-12-31031 December 1998 Forwards LAR for License DPR-22,revising TS pressure-temp Curves Contained in Figures 3.6.1,3.6.2,3.6.3 & 3.6.4, Deleting Completed RPV Sample SRs & Requirement to Withdraw Specimen at Next Refueling Outage & Removing Redundant SR ML20198M3271998-12-28028 December 1998 Submits Change to Commitment for Submittal of ITS Application.Util Plans to Provide ITS Conversion Package Submittal to NRC in Dec of 2000 ML20198J7511998-12-22022 December 1998 Informs of Completion of Listed Commitment Made in Re Severe Accident Mgt. Severe Accident Mgt Guidelines Have Been Assessed,Plant Procedures Have Been Modified & Training of Affected Plant Staff Has Been Completed ML20198J4311998-12-21021 December 1998 Forwards Rev 2 to SIR-97-003, Review of Test Results of Two Surveillance Capsules & Recommendations for Matls Properties & Pressure-Temp Curves to Be Used for Monticello Rpv. Under Separate Cover,Licensee Is Providing LAR to Revise Curves ML20198J7711998-12-14014 December 1998 Documents 981214 Discussion with NRC Staff Re Deviation from Emergency Procedure Guidelines ML20195C8781998-11-11011 November 1998 Forwards Supplement to 971125 License Amend Request Re Condensate Storage Tank Low Level Suction Transfer Setpoints for HPCI Sys & Reactor Core Isolation Cooling Sys ML20195C9631998-11-11011 November 1998 Forwards 120-day Response to NRC GL 98-04, Potential for Degradation of ECCS & CSS After LOCA Because of Construction & Protective Coating Deficiencies & Foreign Matl in Containment ML20195E2261998-11-10010 November 1998 Submits Suppl 1 to Util Response to NRC Request for Addl Info Re 981118 Request for Deviation from Emergency Procedure Guidelines ML20155H6591998-11-0404 November 1998 Forwards Response to 980910 RAI Re GL 96-05, Periodic Verification of Design-Basis Capability of Safety-Related Movs ML20155F9091998-10-27027 October 1998 Forwards Master Table of Contents to Rev 16 of Usar.Info Was Inadvertantly Omitted at Time of 981023 Submittal 05000263/LER-1998-005, Forwards LER 98-005-00,re HPCI Being Removed from Service to Repair Steam Leak in Drain Trap Bypass.Commitments Made by Util Are Listed1998-10-21021 October 1998 Forwards LER 98-005-00,re HPCI Being Removed from Service to Repair Steam Leak in Drain Trap Bypass.Commitments Made by Util Are Listed ML20154L9321998-10-12012 October 1998 Forwards Suppl 2 to LAR & Suppl 980319,which Proposed Changes to Ts,App a of Operating License DPR-22 for Mngp.Number of Addl Typos & One Title Change on Pages Associated with Amend Request Have Been Identified 05000263/LER-1998-004, Forwards LER 98-004-00 Re Manual Scram Inserted Following Pressure Transient Closes Air Ejector Suction Isolation Valves & Trips Offgas Recombiners.Ler Contains Listed Commitment1998-10-0909 October 1998 Forwards LER 98-004-00 Re Manual Scram Inserted Following Pressure Transient Closes Air Ejector Suction Isolation Valves & Trips Offgas Recombiners.Ler Contains Listed Commitment ML20154L8671998-10-0909 October 1998 Forwards Suppl 1 to LAR for License DPR-22, Replacing Exhibits B & C of Original Submittal to Reflect Item 2 & Subsequent Changes.Request for APRM Flow Converter Calibr Interval Extension,Withdrawn ML20154J6201998-10-0505 October 1998 Forwards Rev 49 to Monticello Security Plan.Encl Withheld, Per 10CFR73.21 ML20153F5351998-09-25025 September 1998 Submits Semiannual Status Update on Project Plans for USAR Review Project & Conversion to Improved TS ML20153F0051998-09-25025 September 1998 Forwards Suppl 1 to 971031 Application for Amend to License DPR-22,replacing Exhibit C Which Contains TS Pages Incorporating Proposed Changes Described in Original 971031 Request ML20153D8561998-09-17017 September 1998 Forwards Rev 17 to EPIP A.2-414, Large Vol Liquid Sample &/ or Dissolved Gas Sample Obtained at Post Accident Sampling Sys. Superseded Procedures Should Be Destroyed.Ltr Contains No New NRC Commitments,Nor Does It Modify Prior Commitments ML20153D1441998-09-17017 September 1998 Informs NRC That Listed Commitments 1 & 3 Were Completed by End of 1998 Refueling Outage.Commitments Involved Final Disposition of Remaining Outlier Components Re All Known Outstanding Work Associated with GL 87-02,Suppl 1,USI A-46 ML20153E0331998-09-17017 September 1998 Forwards Response to NRC 980629 RAI Re RPV Weld Chemistry Values Previously Submitted as Part of Plant Licensing Basis.Next Monticello RPV Sample Capsule Scheduled to Be Removed During 1999/2000 Refueling Outage ML20153E9011998-09-0909 September 1998 Forwards Rev 1 to MNGP Colr,Cycle 19, Incorporating Changes to power-flow Maps in Figures 6 & 7.Changes Made to Correct Errors in Stability Exclusion Region & Stability Buffer Region Shown on Rev 0 ML20151S7401998-08-28028 August 1998 Responds to NRC Re Violations Noted in Insp Rept 50-263/98-09.Corrective Actions:Procedure 4 AWI-04.04.03 Will Be Revised to Eliminate Term Urgent from Section 4.3.1.D ML20238E8201998-08-26026 August 1998 Forwards Effluent & Waste Disposal Semi-Annual Rept for Jan-June 1998 & Revised Effluent & Waste Disposal Semi- Annual for Jul-Dec 1997. Ltr Contains No New NRC Commitments,Nor Does It Modify Any Prior Commitments ML20237E9741998-08-26026 August 1998 Forwards Rev 4 to EWI-09.04.01, Inservice Testing Program. Rev of Inservice Testing Program Reflects Valves Added as Result of Component Mods Recently Performed ML20237E6821998-08-25025 August 1998 Forwards fitness-for-duty Program Performance Data for Six Months Period Ending 980630 1999-09-09
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Northem States Power Company 414 Nicollet Mail Minneapolis, Minnesota 55401 1927 Telephone (612) 330-5500 November 28,1994 US Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555 MONTICELLO NUCLEAR GENERATING PLANT Docket No. 50-263 License No. DPR-22 Response to Unresolved items Contained in inspection Report 50-263/94008 Inspection report 50-263/94008 reported the result of an Inspection conducted by Messrs. A Dunlop and J Colaccino on August 8-26,1994. The inspection consisted of a review of the Monticello program for inservice testing (IST) of pumps and valves and of the effectiveness of the program regarding check valves. Two concems were identified regarding the implementation of American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI requirements into the IST program. These items concerned testing relief valves in accordance with OM-1 and the removal of the Residual Heat Removal Service Water control valves from the IST program; unresolved items 94008-01 and 94008-02 respectively. The inspection report requested that Monticello provide a 60 day written response concerning these items.
Attachment A of this submittal provides the requested 60 day response to the two unresolved items contained in Inspection Report 50-263/94008. This letter contains the following new commitments to the NRC:
Following the Code Committee's response to our inquiry concerning the intent of OM-1, Section 4.3, Monticello will provide a supplemental response concerning Unresolved Item 94008-01.
We commit to revise our Inservice Testing Program document to include valves CV-1728 and CV-1729 to be exercised in accordance with IWV-3412.
Please contact Marv Engen, Sr Licensing Engineer, at (612) 295-1291, if you require further information.
11/23/94 NSP H:\ WORD \1STi1R94000. DOC 9412050244 DR 041128 ADOCK 05000263 PDR }I g
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q USNRC- NORTHERN STATES POWER COMPANY November 28,1994 Page 2
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W J Hill Plant Manager 1 Monticello Nuclear Generating Plant j 1
c- Regional Administrator- 111, NRC . !
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- NRR Project Manager, NRC
- Sr Resident inspector, NRC State of Minnesota j Attn
- Kris Sanda j . J Silberg i
l Attachments: A - Response to Unresolved items 94008-1 and 94008-2 i
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TRANSMITTAL MANIFEST NORTHERN STATES POWER COMPANY j NUCLEAR LICENSING DEPARTMENT MONTICELLO NUCLEAR GENERATING PLANT Response to Unresolved items Contained in Inspection Report 50-263/940088 Manifest Date: November 28,1994 Monticello intemal Site Distribution Special Instructions Kaleen Hilsinhoff.. . .USAR File... . .. . ... ..Yes No_X_
Steve Ludders ... . . .NRC Commitment . ...Yes_X_ No Lila Imholt.. . . . . .Monti OC Sec .. .. . .Ye s No X_ - 10, No dist to OC members below if YES Mal Opstad... . ...... ..Monti SAC Sec. . ..Yes_X_ No -6
- Mail Room .... . .Monti Posting. .. . .. .Ye s No_X_ - 7 Monticello Internal Site Distributico:
- Monti Document Control File . L H Waldinger, GM MNGP, SAC W J Hill, Plunt MGR, SAC, OC J C Grubb, NGSS, OC C A Schibonski, GSE, OC L L Nolan, GSSA, OC M F Hammer, GSM, OC B D Day, MGR MTC, OC J E Windschill, GSRS, OC Operating Experience Coord M W Onnen, GSO, OC W A Shamla, NQD Dean Carstens Monti Site Lic File l
Al Wojchouski Steve Hammer Jim Freborg NSP Internal Distribution (Ren Square,8th Floor)
E L Watzl R O Anderson, Dir LMI, SAC T E Amundson, Dir NQD, SAC Communications Dept Yes No_X_
External NSP Distribution
- Doc Control Desk, NRC Kris Sanda, State of Minn Regional Admin-lll,NRC J E Silberg, Beth Wetzel, NRR-PM, NRC Steve Ray, SR Res Insp, NRC
- Advance Distribution made by Site Licensing Correspondence Date : November 28,1994
Attachment A Response to Unresolved items 94008-1 and 94008-2 Unresolved item 94008-1 "It was unclearif the requirements for testing relief valves per OM-1 were adequately incorporatedinto test procedure 0255-02-1B, " Relief Valve Setpoint and Leak Checks." Relief valve testing as required by OM-1 stated that valves should be tested under the similar conditions that they would be expected to see during operation or accident conditions. If testing was performed at ambient conditions, then certified correlations for setpoint testing need to be developed. Several valves had correlations for testing under ambient conditions, however, sufficient information was not available to verify they were pmperty certified as required by the Code. A certified conelation as required by OM-1 should be based on either vendor orlicensee test results for similar valves under similar test conditions. This is considered an unresolved item (263/94008-1) pending furtherinformation on the setpoint correlation."
Response
Monticello test procedure 0255-02-1B, " Relief Valve Setpoint and Leak Checks" performs relief valve testing per OM-1-1981 for liquid relief valves in various ASME Class 2 and 3 safety systems. The procedure is a bench test of the relief valves after they have been removed from the system. Upon satisfactory test completion, the valves are reinstalled. We understand that the above identified unresolved item pertains to the following specific valves tested per procedure 0255-02-1B:
Relief Valve No. System Setpoint (PSIG) Cold Setpoint (PSIG)
RV-1990 A RHR 185 189 RV-1991 B-RHR 185 189 RV-1992 A-RHR 185 189 RV-1993 B-RHR 185 189 RV-2004 A-RHR 500 510 RV-2005 B-RHR 500 510 RV-2025 A-RHR 500 510 RV-2031 A-RHR 185 189 RV-4281 A-RHR 500 515 RV-4282 B-RHR 500 515 Per the standard industry practice, a relief valve is provided by the manufacturer at its design cold set pressure if the operating ambient (environment) temperature or operating fluid temperature exceeds a certain value determined by the manufacturer. For the above listed valves, the ambient temperature does not require application of a setpoint temperature correction factor. However, since the operating fluid temperature has a maximum design value ,
of 281*F, these valves are designed with the cold set pressure shown above. This allows the 11/?3/94 NSF H:\ WORD \1ST\IR94008. DOC
Attachment A Page 2 November 28,1994 valve to lift at the same nominal setpoint for any fluid temperature up to its maximum design value of 281*F.
The Monticello inservice Testing Program for testing of Class 1,2, and 3 valves is governed by the 1986 Edition of the ASME Boiler and Pressure Vessel Code (the Code),Section XI, Subsection iWV, and portions of ASME Operations and Maintenance Standards Part 10 (OM-
- 10) pursuant to 10 CFR 50.55a(f)(4)(iv). Subsection IWV, article IWV-3510, " Safety Valve and Relief Tests", specifies testing of these valves in accordance with ANSI /ASME OM-1-1981 (OM-1). OM-1, section 4, provides test methods for set pressure testing and seat tightness testing. OM-1 subsection 4.1.3.1 provides test media requirements for the testing of relief devices for liquid service and states:
" Valves shall be tested with the normalsystem operating fluid and temperature for which' they are designed. Altemate liquids ordifferent temperatures may be used, provided the requirements of 4.3 are met."
Section 4.3, " Alternative Test Media" states:
" Pressure relief devices may be subjected to set pressure tests and seat tightness tests using a test media (fluid and temperature) other than that for which they are designed, provided the testing complies with 4.3.1, 4.3.2, and 4.3.3".
OM-1, Sections 4.3.1,4.3.2, and 4.3.3 provide requirements for establishing certified correlations for the testing of pressure relief devices with altemate test media.
We consider the use of cold set pressure to be subjecting the valves to a set of test conditions for which they are designed and paragraphs 4.3.1,4.3.2, and 4.3.3 do not apply. The fluid temperature is a design input required by the manufacturer when procuring the valve. The valve is then supplied set at its cold set pressure. Therefore, it is evident that the cold set pressure is a design condition of the valve. The cold setpoint established in procedure 0255-02-1B is provided by the manufacturer and is stamped on the valve or is determined using correction factors provided by the manufacturer. Therefore, cold setpoint testing of these valves is not performed using alternate test media and the requirements of OM-1, section 4.3, are not applicable to the above valves of concern.
Our discussions with other licensees and valve manufacturers has confirmed that our procedure satisfies the Code. In order to provide further confirmation as to the Code's intent we have submitted the following inquiry to the Code committee:
'Does performing a cold setpoint bench test on a Class 2 liquid relief valve, where the
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cold setpoint is provided by the manufacturer, test the valve as it is designed such that 1 paragraph 4.3 and subparagraphs 4.3.1, 4.3.2, and 4.3.3 do not apply?"
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Attachment A Page 3 November 28,1994 We understand that the Code Committee will be meeting in December of 1994 and we have requested that this inquiry be placed on the agenda for the December meeting. Following the Code Committee's review of this inquiry we will provide a supplemental response concerning this issue.
Unresolved item 94008-2
"(Close) Violation (263/92010-05e). This item concemed unacceptable test results for RHRSW control valves CV-1728 and CV-1729 without being documented and evaluated as required by the Code and the IST administrative procedure. Although Form 3107 and Form 3108 were not completed at the time of the deficiency, an operability determination was performed on the valves. The required forms were subsequently completed; however, the deletion of the valves from the IST program appeared incorrect.
The evaluation for Form 3108 determined that the valves' fail-safe function was not safety-related and was beyond the plant's design basis such that the valves were deleted from the ISTprogram. This was based on the control valves being supplied by a safety-related air supply system. The licensee's evaluation determined the valves could be excluded fmm the program based the fail-safe function being non-safety related and on IWV-1200, which excludes control valves. The inspectors agreed the fail-safe function was not required to be tested since it was not safety-related. However, the valves stillperformed a specific safety function that would require theirinclusion in the program. Since the valves were normally closed, the valves were required to open to perform theirsafety-function. This function requires the valves to be exercised to the open position perIWV-3412 and stroke timed per IWV-3413.
\ The licensee was still evaluating this issue at the conclusion of the inspection. This will be considered an unresolved item (253/94008-2) pending inclusion of the valves in the IST program or adequatejustifications for their exemption. Based on the ta nresolveditem, this example of the violation is considered closed."
Response
The Residual Heat Removal Service Water (RHRSW) System function is to supply strained river water to the RHR heat exchangers which remove heat rejected by the Residual Heat Removal (RHR) System during normal reactor shutdown cooling, during reactor isolation or accident conditions, and during the time when the fuel pool has an emergency heat load from ;
spent fuel. )
RHRSW valves CV-1728 and CV-1729 are air operated control valves on the outlet line of the j I
RHRSW side of the "A" and "B" RHR heat exchangers, respectively. These control valves maintain a differential pressure between the RHRSW process stream and the RHR process
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- Attachment A Page 4 November 28,1994 stream during RHRSW system operation. The valves are controlled by a positioner, controlled by a differential pressure indicating controller which senses pressure on the RHRSW discharge line and the RHR inlet line to the RHR heat exchanger. The desired differential pressure control point, and thus the desired valve position for system flow, is manually set by the operator. The normal air supply to these valves is plant instrument air. However, in the event of loss of plant air, air is supplied by safety related auxiliary air compressors energized from essential buses.
The Monticello inservice Testing Program for testing of Class 1,2, and 3 valves is govemed by the 1986 Edition of the ASME Boiler and Pressure Vessel Code (the Code),Section XI, i Subsection IWV, and portions of ASME Operations and Maintenance Standards Part 10 (OM-
- 10) pursuant to 10 CFR 50.55a(f)(4)(iv). RHRSW control valves CV-1728 and CV-1729 are Quality Group "C" components which corresponds to the ASME Boiler and Pressure Vessel l Code Class 3 classification. The valves operate in conjunction with the RHRSW pumps to i provide proper service water flow to the RHR heat exchanger, while maintaining a differential pressure between the RHRSW and the RHR process stream.Section XI, Subsection IWV, i Article IWV-1200 of the Code provides the following criteria for exemn9 ;; valves from testing: ;
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(a) valves that have no specific function in shutting down a mactor orin mitigating the consequences of an accident and used only for: ,
I (1) operating convenience (such as manual vent, drain, instrument, or test valves); ,
(2) system control (such as pressure regulating valves); or (3) maintenance.
(b) extemal control and protection systems responsible for sensing plant conditions l and providing signals for valve operation.
The function of the control valves to open, providing a flow path for RHRSW through the RHR heat exchanger, does not allow exemption of these valves from inservice testing.
IWV-3412 states that valves shall be exercised to the position required to fulfill their function. t The required position to fulfill valve function varies for valves CV-1728 and CV-1729 based on the desired RHRSW pump combination, desired system flow conditions, and RHRSW to RHR I system differential pressure. Based on these factors, the required valve position to perform valve function will be partially open and not a full open position. IWV-3412 provides for demonstrating the necessary valve disk movement by observing indirect evidence (such as ,
changes in system pressure, flow rate, level, or temperature), which reflect stem or disk
- position. The most representative test of the capability of valves CV-1728 and CV-1729 to perform their intended function is performed during inservice testing of the RHRSW pumps.
Quarterly testing of the RHRSW pumps verifies the capability of the valves to operate properly l
, Attachment A Page5 November 28,1994 to pass the maximum required accident flow as well as the valve position necessary to achieve required flow conditions. Testing of the valves in this manner demonstrates valve performance capability as well as monitors for valve degradation.
lWV-3413 requires that a limiting value of full stroke time be established for a power operated valve. Establishment of a limiting stroke time for these valves would be purely arbitrary for code compliance and would not provide any meaningfulinformation regarding valve degradation or capability to perform their required function. Upon receipt of an automatic ECCS initiation signal, the RHRSW pumps will be automatically tripped, if they are running. If the ECCS initiation signal is coincident with the loss of off-site AC power, then the RHRSW pumps will be automatically tripped and locked out. Under conditions of a design basis accident, the low pressure coolant injection mode of RHR is used to restore the water level in the reactor core. Upon restoration of reactor water level by the RHR system, the RHRSW system must be manually started and flow established by the Reactor Operator. There are no automatic initiation signals associated with the RHRSW system. Plant procedures direct the initiation of RHRSW largely at the discretion of the Reactor Operator. Monticello Updated Safety Analysis Report, section 6.2.3.2.1, states; "The containment pressure analysis assumes that the RHR containment spray / cooling mode and the RHR service water pumps are not initiated until 600 seconds after the beginning of the accident....There is not a fixed requirement as to when the containment cooling system (RHR system with heat exchangers to remove accident and decay heat) must be placed in operation since at least eight hours are available before the containment design pressure (56 psi) is reached without cooling under DBA-LOCA condition."
RHRSW control valves CV-1728 and CV-1729 are manually positioned, from the control room, by adjustment of the air operated control valve differential pressure indicating controller to establish the desired RHRSW flow conditions. These valves do not receive an actuation signal (neither by manual handswitch nor by automatic logic) to stroke to the position required to fulfill their safety function. Performing full stroke time testing of these valves is inconsistent with the control scheme design of the valves, the functional requirements of the valves, and the manual positioning of the valves. As these valves are manually positioned and do not receive an actuation signal to full stroke to perform their specific safety function, it is not appropriate to classify them as " power operated valves" within the context of the Code, thus the requirements of IWV-3413 (as well as IWV-3417(a)) are not applicable.
We acknowledge that exclusion of these valves from the plant's Inservice Testing Program, based solely on the fail-safe function being a non-safety related function, was incorrect. Our exclusion of these valves from the program was not a conservative interpretation of the revised exclusion criteria wording of the ASME code implemented by our third ten year interval program. Operability of valves CV-1728 and CV-1729 has been demonstrated and compliance with the Code has been maintained via the existing testing requirements performed on these valves in conjunction with quarterly RHRSW pump testing. We commit to l
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Attachment A Page 6 November 28,1994 revise our Inservice Testing Program document to include valves CV-1728 and CV-1729. The valves are to be included in the program as Category "B" valves with the following requirements of Section XI applied: IWV-3411, IWV-3412, IWV-3416, and IWV-3417(b).
These valves are not subject to the requirements of IWV-3413 nor IWV-3417(a) because they do not act as power operated valves and are not classified as power operated valves as described above. We have performed a review of plant systems to determine if any additional control valves with specific safety functions are not included in the plant's Inservice Testing Program. No other valves were identified.