Letter Sequence Other |
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Results
Other: 05000344/LER-1985-013-01, Forwards Rev 0 to 01-0300-1471, Preliminary Thermal Expansion Evaluation of Reactor Coolant Loop for Trojan Nuclear Plant, as Clarification of LER 85-013-01 in Response to 860430 Telcon, ML19267A316, ML19274E248, ML19274E250, ML20197E758, ML20198Q962, ML20198R016, ML20198R908, ML20198R985, ML20198S019, ML20198S051, ML20199B502, ML20199B510, ML20202C354, ML20204F441, ML20206F998, ML20210S227, ML20210V194, ML20211G407, ML20211G419, ML20212D368, ML20235K098, ML20235Z787
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MONTHYEARML19267A3161978-12-29029 December 1978
Informs That Due to Control Bldg Proceeding,The Mod to Operating Status Alarms of Diesel Generators Will Not Be Completed Until 790131
Project stage: Other
ML19261B7371979-01-26026 January 1979
Forwards Request for Addl Info in Order to Complete Review Re Revised Westinghouse Thermal Design Procedure & WRB-1 Critical Heat Flux Correlation.Requests Response as Soon as Possible
Project stage: RAI
ML19259B6041979-02-0707 February 1979
Forwards Request for Addl Info Re Revised Westinghouse Thermal Design Procedure & Critical Heat Flux Correlation for Facility
Project stage: RAI
ML19274E2481979-03-15015 March 1979
Forwards License Change Application 50 to Amend License NPF-1 Re Radiological Tech Specs.Certificate of Svc & Class III Fee Encl
Project stage: Other
ML19274E2501979-03-15015 March 1979
License Change Application 50 to Amend License NPF-1 Re Radiological Tech Specs
Project stage: Other
ML19254F2471979-10-31031 October 1979
Supplements 790315 License Change Application 50 Incorporating 10CFR50,App I,Radiological Effluent Tech Specs.Includes Spec Implementing 40CFR50.Util to Implement 10CFR190 Until Application 50 Approved & Issued as Amend
Project stage: Supplement
ML20197E7581986-01-31031 January 1986
Rev 0 to Preliminary Thermal Expansion Evaluation of Reactor Coolant Loop for Trojan Nuclear Plant
Project stage: Other
05000344/LER-1985-013-01, Forwards Rev 0 to 01-0300-1471, Preliminary Thermal Expansion Evaluation of Reactor Coolant Loop for Trojan Nuclear Plant, as Clarification of LER 85-013-01 in Response to 860430 Telcon1986-05-0909 May 1986
Forwards Rev 0 to 01-0300-1471, Preliminary Thermal Expansion Evaluation of Reactor Coolant Loop for Trojan Nuclear Plant, as Clarification of LER 85-013-01 in Response to 860430 Telcon
Project stage: Other
ML20210S2271986-05-15015 May 1986
Trip Rept of 860510-11 Site Visit to Participate in Licensee Cold Walkdown of Reactor Coolant Loop Piping & Components, Including Supports,To Assess Whether Cause of LER 85-013, Rev 1 Determined
Project stage: Other
IR 05000344/19860171986-05-19019 May 1986
Insp Rept 50-344/86-17 on 860505-09.No Noncompliance or Deviations Noted.Major Areas Inspected:Onsite Review Committee,Qa Test & Measurement Equipment Program & Surveillance Program
Project stage: Request
ML20198S0511986-05-23023 May 1986
Interim Rept of Independent Review of Trojan Nuclear Plant Reactor Coolant Loop Thermal Movements Evaluation Program. Supporting Info Encl
Project stage: Other
ML20198Q9621986-05-31031 May 1986
PG&E - Trojan - 1986 Snubber Valve Failure Analysis
Project stage: Other
ML20198Q9281986-06-0303 June 1986
Forwards Test Rept 17822, PG&E - Trojan - 1986 Snubber Valve Failure Analysis, Rev 0 to Temporary Plant Test TPT-166, RCS Thermal ... & Westinghouse Amend 3,Rev 2 to Procedure ISI-205,per NRC 860529 Meeting Request
Project stage: Meeting
ML20198R9081986-06-0404 June 1986
Forwards Impell Summary Rept, Evaluation of Reactor Coolant Loop for Restrained Thermal Expansion,Trojan Nuclear Plant
Project stage: Other
ML20206G7531986-06-0505 June 1986
Concludes That Adequate Assurance of RCS Integrity Demonstrated for Mode 3,hot Standby,Based on Review of 860521,0603 & 04 Ltrs & 860521 Meeting.Safety Evaluation Will Be Issued After Certification Received
Project stage: Approval
ML20198S0191986-06-0606 June 1986
Forwards Documentation from Bechtel Corp Re Review & Evaluation of Final Results of RCS Thermal Expansion Analysis & Hot Leg Shell Analysis Performed by Impell Corp. Encl Info Should Close Out Concerns in
Project stage: Other
ML20199B5101986-06-13013 June 1986
Reactor Coolant Thermal Expansion Test,Summary of Data & Predictive Curves,Heatup of 860608-10
Project stage: Other
ML20199B5021986-06-13013 June 1986
Forwards Temporary Plant Test TPT-166, Reactor Coolant Thermal Expansion Test,Summary of Data & Predictive Curves, Heatup of 860608-10, Per .Future Monitoring Program Will Be Documented by 861001
Project stage: Other
ML20211G4191986-06-16016 June 1986
Rev 0 to Reactor Coolant Thermal Expansion Test Final Predictive Curves,Heatup of 860615-16
Project stage: Other
ML20211G4071986-06-16016 June 1986
Forwards Rev 0 to Temporary Plant Test TPT-166, Reactor Coolant Thermal Expansion Test Final Predictive Curves Heatup of 860615-16, Demonstrating That Acceptance Criteria Met
Project stage: Other
ML20202C3541986-06-26026 June 1986
Informs That Remote Monitoring of Electronic Instrumentation Will Be Used to Assure That Steam Generator Snubbers Allow RCS Thermal Movement,Per 860616 Safety Evaluation.Only Remote Monitoring Will Be Repeated During Each Heatup
Project stage: Other
ML20198R9851986-06-30030 June 1986
Rev 0 to Evaluation of Reactor Coolant Loop for Restrained Thermal Expansion,Trojan Nuclear Plant, Summary Rept
Project stage: Other
ML20198R0161986-06-30030 June 1986
Rev 0 to Temporary Plant Test TPT-166, RCS Thermal Expansion Test. Addl Info on Insp of Cast Stainless Steel Hot Leg Elbows Encl
Project stage: Other
ML20206F9981986-06-30030 June 1986
SER in Support of Conclusion That Util Adequately Addressed Technical & Programmatic Aspects of Cause & Effects of Restrained Thermal Growth of Rcs.Ser Re Util Response to TMI Item III.D.3.4, Control Room Habitability Also Encl
Project stage: Other
ML20204F4411986-07-22022 July 1986
Rev 0 to Temporary Plant Test Procedure TPT-175, RCS Thermal Expansion Long-Term Monitoring
Project stage: Other
ML20210L7201986-09-24024 September 1986
Summary of 860529 Meeting W/Util in Bethesda,Md Re Inoperable Steam Generator Hydraulic Snubbers Relationship to Restrained Thermal Growth of Rcs.Supporting Info Encl
Project stage: Meeting
ML20210V1941986-09-25025 September 1986
Informs of Results of RCS Thermal Expansion Monitoring Performed During Wk of 860902.Measurements Acceptable.Data Will Be Taken When RCS Temp Reaches 340 F,In Lieu of Holding Temp at 340 F for Analysis & Evaluation
Project stage: Other
ML20235Z7871986-10-15015 October 1986
Discusses Insp on 860318-0428 & Two Special Insps on 860513- 0612 & 860804-14 & Forwards Notice of Violation & Proposed Imposition of Civil Penalty
Project stage: Other
ML20212D3681986-12-19019 December 1986
Submits Assessment of Impact of Replacing Graphite Shims W/ Carbon Steel Shims on Hot Leg Pipe Whip Restraints.Potential Effect on RCS Piping in Event of Future Contact Between Hot Leg Piping & Whip Restraints Addressed
Project stage: Other
ML20235K0981987-09-30030 September 1987
Informs That Periodicity of RCS Thermal Expansion Monitoring Surveillances Returned to Normal Following Completion of Twice Per Shift Exam to Detect Possible RCS Leakage as Reported to NRC on 870724
Project stage: Other
1986-05-09
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PORTLAND GENERAL ELEcTHic COMPANY 121 S.W. S ALM O N STREET Po RT LA N o. O R EGO N 97204 WILLIAM J. LIND8 lao I
"5" csos)aze-sa7s December 19, 1986 l
Trojan Nuclear Plant Docket 50-344 License NPF-1 Director of Nuclear Reactor Regulation ATTN:
Mr. Steven A. Varga
[
Director, PWR-A Project Directorate No. 3 U.S. Nuclear Regulatory Comission Washington DC 20555
Dear Mr. Varga:
Use of Carbon Steel vs Graphite Shims in Reactor Coolant System Your safety evaluation of June 16, 1986 for the inoperable steam generator snubbers on the Trojan Nucleer Plant reactor coolant loop, directed PCE to provide an assessment of the impact of replacing graphite shims with carbon steel shims on the hot les pipe whip restraints. This assessment is pro-vided herein and addresses the potential effect on the Reactor Coolant System (RCS) piping in the event of future contact between the hot leg piping and the whip restraints.
Since 1982, when a pressurizer surge line thermal sleeve was removed, unexpected movements of the pressurizer surge line had been observed. As a result of steam generator hydraulic snubber testing in 1985, all 16 snubbers were declared inoperable. One of the hypotheses from the evaluation of the pressurizer surge line movements was that the snubbers might have failed in such a manner so as to restrain normal RCS thermal expansion, thereby causing the observed movement.
Subsequently, a thorough examination of RCS components was performed during the 1986 refueling outage. The examination suggested that the RCS piping had experienced abnormal movement in that damage to the graphite shims on each of the hot leg whip restraints was observed. The examination also revealed the gaps between the steam genera-tors and the upper and lower seismic support rings were insufficient to accomodate the expected thermal expansion of the RCS loops.
Following extensive snubber control valve testing and further inspection and evaluation of the RCS loop, it was determined that the inadequate gaps discussed above were the most likely cause of the damage to'the hot leg whip restraint shims. A detailed analysis of RCS thermal expansion was conducted and the whip restraint gaps were adjusted to the proper clearances. As part of this adjustment, the hot leg whip restraint graphite shims were replaced with carbon steel shims. The modified clearances were monitored during subsequent heatups and cooldowns and it was verified the RCS did not contact the whip restraints.
v 8612310345 861219 N
PDR ADOCK 05000344 P
pyg I
e PORTLAND GENE!*AL ELECTRIC COMPANY Mh. Steven A. Varga i
December 19, 1986 Page 2 Since the cause of the original RCS contact with the hot leg whip restraints was determined to be inadequate clearances on the steam generator support rings and possibly at the hot leg whip restraints themselves, and the cause has been corrected, it is not expected that the RCS will again contact the whip restraints. A temporary monitoring program has been developed to demonstrate the thermal expansion of the RCS occurs within predictable limits.
This monitoring program will continue until this predictability has been established.
The pipe whip restraints are designed to limit movement of the pipe in the event of a pipe rupture.
During normal operations, it is not intended for the pipe to contact the restraint.
In the event the pipe was to contact the carbon steel shims during normal operations, finite element analyses have demonstrated under worst-case conditions the faulted load stress limits of the pipe would not be exceeded.
s A parceived benefit of the graphite shims was that they " crushed" under load, thereby relieving the stress on the RCS loop. Since it is not intended for the pipo to contact the restraint and our efforts have been to preclude such contact, this apparent advantaga has become a moot issue. Under pipe rupture conditions, it is not necessarily desirable for the shims to " crush".
In conclusion, it has been determined that the replacement of the graphite shims in the RCS hot leg whip restraints with carbon steel shims will not have an adverse effect. The RCS whip restraint gaps have been adjucted in order to prevent any further contact between the shims and the RCS piping and a tamporary monitoring program is being conducted to confirm contact does not occur.
In the unusual event that contact should occur, the stresses impesed on the RCS would be acceptable.
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Sincerely, eg sMe l
c:
Mr. John E. Martin Regional Administrator, Region V U.S. Nuclear Regulatory Commission Mr. David Kish, Acting Director State of Oregon Department of Energy
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