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MONTHYEARML20236C3041976-09-14014 September 1976
Notification of 760928 Meeting W/Util in Bethesda,Md to Discuss Util Concepts for Proposed Neutron Shield for Plant. Summary of Neutron Streaming Problem Encl
Project stage: Meeting
ML19344A8901978-07-31031 July 1978
Forwards Addl Info Re Control of Heavy Loads Near Spent Fuel,Proposed Neutron Shield Design & Withdrawal of Applicants Tech Spec Proposed Change Re Revision to Containment Leak Rate Testing
Project stage: Withdrawal
ML20064C2341978-10-11011 October 1978
Forwards Proposed Tech Specs to Require Diesel Generator Start Time of Less than or Equal to 20
Project stage: Other
05000336/LER-1978-024-03, /03L-0 on 781004:during Surveillance Testing, Setpoint for Reactor Protec Sys Reactor Coolant Low Flow Trip on Channel B Was Out of Spec.Setpoint Was Readjusted. Rev Will Req Low Flow Trip Unit Setpoint Be Reset1978-11-0101 November 1978
/03L-0 on 781004:during Surveillance Testing, Setpoint for Reactor Protec Sys Reactor Coolant Low Flow Trip on Channel B Was Out of Spec.Setpoint Was Readjusted. Rev Will Req Low Flow Trip Unit Setpoint Be Reset
Project stage: Request
ML20064D4921978-11-0101 November 1978
Presents Details of Stretch Power Effort in Advance of Execution to Enable Facility to Operate at 2,700 Mwt Following Second Refueling Outage
Project stage: Other
ML20049A1301978-11-13013 November 1978
Forwards Nuc Shield Design as Alternate to That Proposed in .Design Incorp ALARA Considerations & Philosophy. Requests Expeditious Review to Allow Target Installation Date.Oversize Drawing Available in Central File
Project stage: Other
ML20027A4181978-11-15015 November 1978
Notification of 781121 Meeting W/Ne Nuc Energy Co to Discuss Tech Review Effort & Schedule Necessary to Authorize Stretch Pwr Level Rating of 2700MWT
Project stage: Meeting
ML20064E4531978-11-18018 November 1978
Forwards Sys Description for Reactor Coolant Pump Shaft Speed Trip Function Being Installed by Util to Replace Steam Generator Differential Pressure Sys.Believes Proposal Is Exempt from Amend Fee
Project stage: Other
ML20062G3871978-12-14014 December 1978
Sleeved Guide Tube Inspec Prog, CEN-104(N)-NP.Describes Inspec Prog to Show Performance Re Wear.Edited to Delete Info W/Held from Pub Disclosure IAW 10CFR2.790
Project stage: Other
ML20064H4291978-12-15015 December 1978
Appl to Amend Oper Lic DPR-65 by Increasing Maximum Allowable Thermal Output of Subj Facil Over 2700MWt to Commence W/Cycle 3 on 790515.W/att Environ Impact Appraisal, Eval of Radiological Consequences & Site Meteorology
Project stage: Other
ML20062G3811978-12-18018 December 1978
Forwards non-proprietary Version of Sleeved CEA Guide Tube Inspec Prog,CEN-104(N)-NP,relating to Amend 38 to DPR-65. IAW 10CFR2.790.W/encl Affidavit
Project stage: Other
ML19256A5231978-12-21021 December 1978
Summary of 781221 Meeting Re Neutron Shield Design.Applicant Will Forward Qualifications of Design by mid-January 1979 & Request a Safety Evaluation by 790331
Project stage: Approval
ML19267A2841978-12-28028 December 1978
Requests Acceptance of Proposed Changes to Tech Specs for License DPR-65 Adding Requirement of Operable Charging Pump & Operable Flow Path as ECCS Subsys.W/Att Description of Proposed Changes
Project stage: Request
ML19274D2061978-12-29029 December 1978
Pursuant to 781215 Request by Applicant to Increase Thermal Power of Unit to 2700 Mwt,Or Requests IE Evaluation of Utils Past Mgt & Discussion of Outstanding Insp Items Re Safe Operation at Stretch Power Level
Project stage: Other
ML19274D2521979-01-0303 January 1979
Summary of Meeting on 781121 in Bethesda,Md to Discuss Stretch Power Program Including Control Element Assembly Guide Tube Wear Basis,Cycle 2 & 3 Assemblies,Eccs, & Low Flow Trip.W/Tentative Review Schedule & Submittal Schedule
Project stage: Meeting
ML19259A8121979-01-0808 January 1979
Ack Receipt of 781101,781108 & 781215 Requests for Review & Comments on Stretch Power,Core Reload & Reactor Coolant Pump Speed Sensing Sys.Staff Considers Requests as First Submittals for Amend & Are Subj to Class V Fee
Project stage: Other
ML19256A8481979-01-0909 January 1979
Discusses Status of Proposed Tech Spec Revision Increasing Response Time of Encl Bldg Filtration Sys.Disagrees W/Nrc Dose Calculations.Contends 151 Rems Thyroid Dose & 3.8 Rems Whole Body Does Are Conservative
Project stage: Other
ML19263B9261979-01-17017 January 1979
Forwards $25,800 Payment for Licensing Fee Per NRC
Project stage: Other
ML19269C8461979-02-0606 February 1979
Responds to NRC Verbal Request for Addl Info on Proposed Mod to Tech Specs Re ECCS Performance Calculations for Charging Pump Flow.No Plant Mod Is Necessary & There Is No Change in Valve Motion or Operability Requirements
Project stage: Request
ML19263C4881979-02-12012 February 1979
Proposes Amend to License DRR-65.Proposed Changes Include Core Performance Characteristics & Safety Analyses Re Operation at 2,700 Mwt
Project stage: Request
ML19283B6651979-02-23023 February 1979
Forwards Rept on Neutron Shield Design Qualification. NRC Concerns Raised During 781128 Meeting Are Addressed in Rept. Items on Shield Movement,Leakage & Pressure Relief Valves Discussed in Forwarding Ltr
Project stage: Meeting
ML19281A0961979-02-27027 February 1979
Supports Proposed License Amend to Increase Rated Thermal Power to 2,700 Mwt,Starting W/Cycle 3 Operation.Forwards Addl Info Re Control Rod Ejection,Main Steam Line Rupture & Seized Rotor Event
Project stage: Other
ML19276E4911979-03-0202 March 1979
Forwards Proposed Amend to License DPR-65,changing Tech Specs to Incorporate Reactor Coolant Pump Speed Sensing Sys. W/Affidavit & Oversized Drawing
Project stage: Other
ML19276F5721979-03-14014 March 1979
Request for Addl Info Re Utils Cycle 3 Reload/Stretch Power Request of 790212
Project stage: RAI
ML19282C6011979-03-22022 March 1979
Forwards Results of non-LOCA Safety Analysis Necessary to Support Cycle 3 Operation at 2,700 Mwt
Project stage: Other
ML19282C5631979-03-23023 March 1979
Util Determined Certain stem-mounted Limit Switches on safety-related Valves May Not Be Suitable for Svc in LOCA Containment.Forwards List of Specific Switch Types & Valves on Which Switches Are Installed
Project stage: Other
ML20244A5481979-03-23023 March 1979
Provides Trip Setpoint for Proposed Tech Spec Revision to Incorporate Reactor Coolant Pump Speed Sensing Sys as Addition to Reactor Protection Sys
Project stage: Other
ML19273B5681979-03-27027 March 1979
Responds to NRC 790314 Request for Addl Info Re Cycle 3 Operation
Project stage: Request
ML19282C6151979-03-28028 March 1979
Forwards Info Requested in .Desribes Loop Current Step Response Method of in Situ Response Time Testing of Resistance Temperature Detectors
Project stage: Other
ML20037A2041979-03-29029 March 1979
Forwards Proposed Insp Program for Sleeved Guide Tubes. Program Withheld (Ref 10CFR2.790)
Project stage: Other
ML19273B5491979-03-30030 March 1979
Forwards Results of Large Break Loca/Eccs Performance Analysis.Qa Verification by C-E,NSSS & Fuel Supplier Is Expected on 790410.W/basis for no-fee Determination
Project stage: Other
ML20076A8021979-04-0606 April 1979
In Answer to IE Bulletin 79-01,environmentally Qualified stem-mounted Limit Switches W/Appropriate Documentation Can Be Procured.Util Will Replace Switches W/Qualified Switches Prior to Start of Cycle 3 Operation Scheduled for May 1979
Project stage: Other
ML19276F8081979-04-0909 April 1979
Advises That QA Verification of Small & Large Break Loca/Eccs Performance Analysis Has Been Completed.Ltrs to NRC & 790330 Are Sufficient to Show Compliance w/10CFR50.46 Criteria
Project stage: Other
ML20244A5851979-04-12012 April 1979
Forwards Info Re Proposed Revisions to Tech Specs to Reflect Installation of Reactor Coolant Pump Speed Sensing Sys to Provide Protection for 4-pump Loss of Flow Event
Project stage: Other
ML20076A8431979-04-13013 April 1979
Steam Generator Insp
Project stage: Other
ML20076A8361979-04-13013 April 1979
Eddy Current Test Has Been Performed.No Tube Defects Discovered & No Corrective Action Based on Tube Defects Required.Preliminary Analysis of Results of Insp Indicates Denting Rate Has Been Reduced
Project stage: Other
ML20076A9041979-04-18018 April 1979
Responds to IE Bulletin 79-01.One Solenoid Installed on safety-related Valve in Plant Containment Has Not Been Qualified for Svc in LOCA Environ.Solenoid Will Be Replaced Before Start of Cycle 3 Operation
Project stage: Other
ML19270G0861979-04-24024 April 1979
Responds to IE Bulletin 79-07 Re Seismic Stress Analyses of safety-related Piping
Project stage: Other
ML20076A9551979-04-26026 April 1979
Responds to NRC 780419 Request for Insulation Resistance Tests.Forwards Electrical Penetration Testing Evaluation. Modified Conductor Configuration Is Acceptable for Cycle 3 Operation
Project stage: Request
ML20076A9521979-04-26026 April 1979
Discusses Guide Tube Insp Program.Pull Tests Performed on 35 Sleeves.Evaluation of Data Indicates Sleeves Are Functioning as Expected
Project stage: Other
ML19289F4721979-04-30030 April 1979
Responds to IE Bulletin 79-04, Incorrect Weights for Swing Check Valves Mfg by Velan Engineering Corp. Lists Subj Valves Used in Seismic Category I Piping Sys.Investigation Indicates No Drawing Discrepancies Re Valves
Project stage: Other
ML20037A2291979-05-0202 May 1979
Response to NRC Question 2.2 on Cycle 3 Safety Analysis, Nonproprietary Version
Project stage: Other
ML20244A6131979-05-0303 May 1979
Forwards Responses to Request for Addl Info Re Cycle 3 Safety Analysis.Nonproprietary Rept Encl.Proprietary Version Withheld (Ref 10CFR2.790)
Project stage: Request
ML20037A2281979-05-0303 May 1979
Forwards Responses to Request for Addl Info Re Cycle 3 Safety Analysis.Nonproprietary Rept Encl.Proprietary Version Withheld (Ref 10CFR2.790)
Project stage: Request
ML20076B0161979-05-0707 May 1979
Discusses Addl Info Re Control Element Assembly Guide Tube Insp.Fuel Assemblies Suitable for Continued Use During Cycle 3 Operation
Project stage: Other
ML19259C2731979-05-0707 May 1979
Supplements 790507 Response to IE Bulletin 79-07. Safety-related Piping Meets Seismic Sys Criteria
Project stage: Supplement
ML20076B0871979-05-0808 May 1979
Notifies That stem-mounted Limit Switches on Valves SI-614, SI-624,SI-634 & SI-644 Will Not Be Replaced;Valves Do Not Perform safety-related Function
Project stage: Other
ML20076B0881979-05-11011 May 1979
Discusses Results of Six Tasks Which Utilized Adlpipe Computer Program for Reanalysis of Piping Sys.Mods Involved Minor Changes to Small Portions of Existing Sys
Project stage: Other
ML19269E3551979-05-12012 May 1979
Amend 52 to License DPR-65 Authorizing Cycle 3 Operation at 2,560 Mw W/Modified Guide Tubes for Control Element Assemblies
Project stage: Request
ML20076B0891979-05-17017 May 1979
Forwards Addl Info Re Performance of Control Element Assembly Guide Tubed Sleeves,In Response to
Project stage: Other
1979-03-14
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Text
INHrrHI!AST tmuTHIS P o BOX 270 q
HARTFORD, CONNEC11 CUT 06101 1
(M3) 6f66911 k
L J
April 26 1979 '
Docket N 50-336 5c Director of Nuclear Reactor Regulation Attn:
Mr. R. Reid, Chief Operating Reactors Branch #4 U. S. Nuclear Regulatory Commission Washington, D. C.
20555 Q
Reference:
(1)
W. G. Counsil letter to R. Reid dated April 17, 1979.
D Gentlemen:
Millstone Nuclear Power Station, Unit No. 2 n
CEA Guide Tube Inspection Program i
In Reference (1), Northeast Nuclear Energy Company (NNECO) responded to verbal NRC Staff questiens concerning the guide tube inspection program. Reference (1) also indicated that selected sleeves would be subjected to a pull test to the load indicated in the response to question one.
Subsequently, additional: pull tests have been completed; the results of these tests are summarized below.
A total of 35 sleeves in 7 discharge fuel assemblies were subjected to the pull test.
Of these, 34 showed no axial movement. On one of the sleeves in Assembly B047, the pull force inadvertently exceeded the specified limit and reached approximately twice the prescribed load. This sleeve moved axially upward n) app'rosiiiiiifely 0.25 inches.
!x NNECO has evaluated these data and has concluded that the sleeves are functioning mechanically as expected and sleeved assemblies are acceptable for additional in-core use.
Our NSSS vendor and fuel supplier, Combustion Engineering, has concurred in this determination. The pull force was originally established to represent a conservative upper limit on CEA drag force and on the axial force due to flow. The fact that the sleeve moved at a higher force is not considered to be a negative result.
In fact, the load required for movement in the hot condition is expected to increase significantly since dif ferential thermal expan-sion causes an intimate contact between the sleeve and guide tube that was not present in the test.
NNECO and CE reviewed the results of unitradiated pull test laboratory results for both " original" (i.e., not thermally cycled) and thermally cycled samples.
The results of these laboratory tests indicated some slight motion of the sleeve during pull tests for both cycled and uncycled samples, approximately twice the vertical displacement at the same pull force for the thermally cycled 7005040053 G
)
i case when compared to the uncycled casr. Although more motion at the same pull force level was noted for sleeve / guide tube crimp joints which had been thermally cycled, the effect of additional thermal cycles af ter the initial cycle is not significant. This phenomenon is attributable to the 3,
relatively rapid relaxation rate of zircaloy.
1 Sleeve development tests also revealed that the peak force required to remove a sleeve completely from the end fitting post was not significantly reduced by thermal cycling. This indicates that the potential for complete sleeve removal is not affected by long-term operation. Although irradiation effects were not directly incorporated into these tests, they were simulated by the use j
of cold worked materials. As expected, the pull to withdrawal force was higher for the cold worked material.
Because one sleeve in one assembly was displaced when subjected to the higher load, NNECO's test program was expanded to a total of 35 sleeves. No additional movement was evidenced in any other sleeve. Since both irradiation effects and O'-
thermal cycling effects saturate early (approximately 90% complete in one operating cycle), NNECO and CE have concluded that no significant loss of joint integrity will occur during further plant operation. Accordingly, NNECO has concluded that the Millstone Unit No. 2 fuel assemblies are suitable for continued plant operation. We trust the above information, coupled with previously submitted material regarding CEA guide tubes and sleeves, is sufficient for you to concur with our evaluation.
Very truly yours, NORTHEAST NUCLEAR ENERGY COMPANY
/
/ ?'-
.ffl4142L '
/
W. G. CounsL1 Vice President O
e
