ML17056C079
| ML17056C079 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 10/16/1992 |
| From: | Terry C NIAGARA MOHAWK POWER CORP. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML17056C080 | List: |
| References | |
| GL-88-11, GL-92-01, GL-92-1, NMP1L-0707, NMP1L-707, TAC-M83486, NUDOCS 9210220181 | |
| Download: ML17056C079 (8) | |
Text
ACCEI IRATED 'DISTRIBUTIONDEMCSTRATION SYSTEM
~.s REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)
SUBJECT:
Forwards final rept, "Elastic-Plastic Fracture Mechanics Assessment of Nine Mile Point Unit 1 Beltline Plates for Service Level A 6 B Loadings," re GL 92-0l,rev 1.
DISTRIBUTION CODE:
A047D COPIES RECEIVED:LTR ENCL SIZE:
TITLE: OR Submittal: Inservice Inspection/Testing/Relief from ASME Code NQTEs:S~ 84 4'<~Q RECIPIENT COPIES COPIES ID CODE/NAME LTTR ENCL LTTR ENCL PD1-1 LA 1
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RECIPIENT ID CODE/NAME PD1-1 PD ACCESSION NBR-9210220181 DOC.DATE: 92/10/16 NOTARIZED: NO DOCKET FACIL:50-220 Nine Mile Point Nuclear Station, Unit 1, Niagara Powe 05000220 AUTH.NAME AUTHOR AFFILIATION TERRYiC.D.
Niagara Mohawk Power Corp.
RECIP.NAME RECIPIENT AFFILIATION Document Control Branch (Document Control Desk)
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NIAGARAMOHAWKPOWER CORPORATION/301 PLAINFIELDROAD, SYRACUSE, N.Y. 13212/TELEPHONE (315) 474-1511 October 16, 1992 NMP1L 0707 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C.
20555 Re:
Nine Mile Point Unit 1 Docket No. 50-220 DPR-63 TAC N M
4 Subject; Generic Letter 92-01, Revision 1, Reactor Vessel Structural Integrity, 10 CFR 50.54(f), Supplemental Analysis in Accordance with Appendix X to Section XI of the ASME Code Gentlemen:
The Nuclear Regulatory Commission issued Generic Letter 92-01 on March 6, 1992, as part of a program to evaluate reactor vessel integrity and to ensure that licensees are complying with 10CFR50.60 and 10CFR50.61, and are fulfillingcommitments made in response to Generic Letter 88-11.
In response to this request, Niagara Mohawk submitted a report for Nine MilePoint Unit 1 (NMP1) on July 2, 1992 documenting compliance with 10CFR50.60 and 10CFR50.61 and the guidance provided in Regulatory Guide 1.99 (Revision 2).
Our July 2, 1992 response identified two reactor vessel beltline plates (G-307-4 and G-8-1) with predicted Charpy uppershelf energies (USE) below the 50 ft-lb screening criterion at the present time, based on the conservative models of Regulatory Guide 1.99 (Revision 2).
By letter dated August 12, 1992, the Staff provided the results of its preliminary review of our July 2, 1992 response.
The Staff requested that Niagara Mohawk perform an elastic-plastic fracture mechanics analysis to demonstrate that the two beltline plates with USE less than 50 ft-lb have margins of safety against fracture equivalent to those required by Appendix G of the ASME Code.
The Staff recommended that the criteria in the latest revision of Appendix X to Section XIof the ASME Code and the material properties from NUREGICR-5729," Multivariable Modeling of Pressure Vessel and Piping J-R Data," be used in the analysis.
Enclosed is a copy of our report entitled, "Elastic-Plastic Fracture Mechanics Assessment of Nine Mile Point Unit 1 Beltline Plates for Service Level A and B Loadings".
This report documents the calculations reported during our September 30, 1992 meeting with your Staff in Washington, D.C.
As requested by the Staff, the calculations in the enclosed report are based on a revised A302B material model.
The A302B J-R curves are assumed to be flat pOO Qaftg the initial 0.1 inches of crack extension.
Based on the ASME Appendix X analysis, it has been concluded that the NMP1 vessel is safe in terms of ductile fracture failure through end-of-license for Service Level A and B loadings.
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Niagara Mohawk willalso be performing an analysis for Service Level C and D loadings.
The acceptance criteria given in the draft ASME Appendix X willalso be applied to the G-8-1 and G-307-4 plates for Service Level C and D loadings.
The revised A302B material model willbe used in the analysis and only the limiting axial flaw orientation reviewed to
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- identify the limiting events from a ductile fracture perspective.
Based on this review, three candidate Level C transients and two candidate Level D transients have been identified.
These transients willbe screened by performing a thermal transient heat transfer analysis to identify the limiting transient.
Once the limiting Level C and D transients are identified, the ASME Section III, Appendix G, Subsection G-2214.3(b) approach for calculation of the thermal stress intensity factor (Kir)willbe followed. In particular, the stress distribution in the vicinityof the flaw willbe calculated using a one dimensional finite element model.
The Raju-Newman stress intensity factor influence coefficient method willthen be used to determine Krr. Influence coefficients for surface flaws recently proposed by the ASME Working Group on Flaw Evaluation [CI92] willbe used in the analysis, Zahoor's [ZA91]
stress intensity factor solution for a finite length axial part-throughwall flaw willbe used to calculate K~. The results of the Service Level C and D loading calculations willbe documented in a separate report to be submitted by January 29, 1993.
During our September 30, 1992 presentation, the Staff requested that calculations be performed to accurately determine the uncertainty in the USE estimates for the beltline welds.
The analytical procedure used, which was briefly described at the meeting, willbe applied to and verified using data from the latest version of the NRC Power Reactor Embrittlement Data Base.
A separate report, addressing the USE estimates for the beltline welds, willbe included in our January 29, 1993 submittal.
The elastic-plastic fracture mechanics analysis performed has shown that the axial flaw is the limiting orientation.
Since 1972, the transverse-longitudinal (T-L) orientation has been required by ASME and used in the nuclear industry for analysis of pressure vessels.
The 50 ft-lb screening criterion is also evaluated based on the T-L orientation.
However, a more consistent approach would be to evaluate the axial flaw using L-T Charpy USE data and to evaluate the circumferential flaw using T-L Charpy USE data.
When the appropriate orientation is considered, the margin between the maximum allowable USE and the predicted actual USE at end-of-license is on the order of 38 ft-lbs. This margin of safety is in addition to the safety factors applied to the ASME Appendix X equations.
Therefore, it has been concluded that the NMP1 vessel is safe in terms of ductile fracture failure through end-of-license for Service Level A and B loadings.
Based on analyses performed to date, we expect that the Level A and B margins are more limiting than the Service Level C and D loading margins.
We also believe that the weld uncertainty analysis willdemonstrate that the USE estimates used are conservative.
During the presentation, the Staff also asked how Niagara Mohawk had conformed to the regulations regarding USE at the time of issue (10CFR50, Appendix G, May 1983).
Niagara Mohawk has no documentation to indicate that the 50 ft-lb USE screening criterion was evaluated at that time.
Niagara Mohawk personnel and our reactor vessel material consultant involved at the time do not recall any actions taken to address the 50 ft-lb screening criterion of Appendix G. During 1983, surveillance capsule C specimens (removed from the reactor vessel in 1982) were being tested by Battelle.
The results of this testing showed an
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unexpected large shift in the 30 ft-lb index temperature and a negligible drop in the upper shelf energy.
Because of this large shift and the fact that the shelf drop was negligible, brittle fracture (RT>>T) rather than ductile fracture (USE) became the focus of Niagara Mohawk's efforts.
These results prompted Niagara Mohawk to begin an investigation into a potential material mix-up in the surveillance program and to reinsert two capsules to generate additional surveillance data in the future.
The result of this investigation was a new baseline surveillance document for NMP1 which was issued in January 1991 and transmitted to the Staff on August 30, 1991.
References:
[CI92] Cippola, R., Proposed Changes to ASME Code Article A-3000, "Method for K, Determination", ASME Working Group on Flaw Evaluation, Chicago meeting notes, 1992
[ZA91] Zahoor, A., "Ductile Fracture Handbook", Volume 3, EPRI Research Project 1757-69, January, 1991 Very truly yours, C. D. Terry Vice President Nuclear Engineering AER/mls 003237GG Enclosure XC:
Regional Administrator, Region I Mr. R. A. Capra, Director, Project Directorate I-l, NRR Mr. D. S. Brinkman, Senior Project Manager, NRR Mr. W. L. Schmidt, Senior Resident Inspector Mr. K. Cozens, NUMARC Records Management
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