ML20085A666
| ML20085A666 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 06/07/1995 |
| From: | Keaten R GENERAL PUBLIC UTILITIES CORP. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 5000-95-0076, 5000-95-76, C321-95-2175, TAC-M90104, NUDOCS 9506140251 | |
| Download: ML20085A666 (4) | |
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GPU Nuclear Corporation
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E Gu Parsippany, New Jersey 07054 201-316-7000 TELEX 136-482 June 7, 1995 Writers Direct Dial Numbec 5000-95-0076 C321-95-2175 U. S. Nuclear Regulatory Commission Att: Document Control Desk Washington, DC 20555 Gentlemen:
Subject:
Oyster Creek Nuclear Generating Station (OCNGS)
Docket No. 50-219 Facility Operating License No. DPR-16 Core Shroud Enhancement - Inservice inspection and XM-19 Material Testing Programs
References:
NRC Letter dated November 25,1994, " Safety Evaluation Regarding the Oyster Creek Core Shroud Repair,"
TAC No. M90104 The referenced NRC Safety Evaluation requested GPU Nuclear to submit, within 6 months after plant start up from the 15R outage, an inservice inspection program for the core shroud repair / tie rod assemblies and a material testing program for hot rolled XM-19 in a simulated BWR environment. Attachment 1 provides the requested augmented inservice inspection program and Attachment 2 provides the proposed hot rolled XM-19 material test program.
GPU Nuclear plans to initiate the XM-19 material testing immediately upon receipt of NRC acceptance of the proposed testing program.
If you have any questions or comments on this submittal, please contact Mr. Michael Laggart, Manager, Corporate Nuclear Licensing at (201) 316-7968.
Sincerely, h
R. W. Keaten Vice President and Director Technical Functions Attachments RT71plp 950614 hh19 c: Administrator, Regior PDR A
PDR O
Senior Resident inspecm.
Oyster Creek NRC Project Manager l
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l GPU Nuclear Corporation is a subsid;ary of General Pubhc Utilities Corporation
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.9 Pcge I of 2 ATTACilMENT 1 OYSTER CREEK NUCLEAR GENERATING STATION AUGMENTED INSERVICE INSPECTION (ISI) i CORE SIIROUD & TIE ROD ASSEMBLIES Purpose By letter dated November 25,1994, the NRC requested GPU Nuclear to submit an inspection program for the ten core shroud tie rod assemblies placed in service during the 15R outage.
These tic rod assemblies stabilize and reinforce the core shroud should cracking of weld (s) occur and exceed acceptable limits. The proposed augmented ISI program for the Oyster Creek shroud enhancement follows.
Tie-Rods Perform a VT-3 examination of two (2) tie-rod assemblies each refueling outage with the initial inspection taking place during the 16R outage scheduled for the fall of 1996. The specific tie rod assemblies selected will vary each refueling outage. The two (2) selected tie rod assemblies should be approximately 180" apart. The examination results shall be compared to the preservice base line inspection. The examination to include all accessible areas with emphasis on the following:
- Ilook positioning.
- Lateral supports (bumpers) orientation and gaps (vessel side and shroud side), including keeper plate gap.
- Top bracket attachment to rod and spacer ring mating with top flange of shroud.
- Crimp on top bracket nut.
Bracket Attachment to Conical Support (Clevis Assembly) i Perform a VT-1 examination of the accessible attachment welds (cone to bracket and bracket to shroud) and adjoining clevis assemblies on each side of the two (2) selected tie rods (see above) each scheduled refueling outage with the initial inspection taking place during the 16R outage scheduled for the fall of 1996.
i Conical Suonort to Vessel Weld (119)
Perform a VT-1 cxamination of the areas inspected during the 15R outage (4 segments approximately 56" long each) each scheduled refueling outage with the initial inspection taking place during the 16R outage scheduled for the fall of 1996.
Page 2 of 2 Shroud Vertical Welds Perform a VT-1 examination of one (1) vertical weld (shroud O.D. only) each scheduled refueling outage with the initial inspection taking place during the 16R outage scheduled for the fall of 1996. The weld selected will vary each refueling outage.
Too Guide Rine Seement Radial Welds Verify the integrity of both radial top guide ring segment welds by direct or indirect method (s) each scheduled refueling outage with the initial inspection taking place during the 16R outage scheduled for the fall of 1996.
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Page 1 of 1 s
ATTACIIMENT 2 OYSTER CREEK NUCLEAR GENERATING STATION (OCNGS)
IIOT ROLLED XM-19 MATERIAL TEST PROGRAM Purpose By letter dated November 25, 1994, the NRC requested GPU Nuclear to submit a material testing program for hot rolled XM-19 used in the core shroud enhancement placed in service during the 15R outage. The purpose of this test program is to demonstrate the resistance to intergranular stress corrosion cracking (IGSCC) of hot rolled XM-19 materials in a simulated BWR environment under a creviced condition such as in a threaded configuration. The proposed material testing program for the Oyster Creek shroud enhancement follows.
Test Procram Testing will be accomplished utilizing cylindrical CERT type specimens containing a threaded section cut to the same geometry as that utilized in the core shroud tie rods. A crevice geometry will be established around the threaded section to simulate the condition existing at the tie rod ends. Materials will be archive specimens from the same heats of material utilized in the core shroud tie rods installed at Oyster Creek (GPUN) and Fitzpatrick (NYPA) Nuclear Power Plants.
The test medium will be simulated BWR reactor coolant at 550 F with a minimum 8 ppm oxygen.
Contaminant levels will be controlled to maintain conductivity of 0.3 to 0.5 i
microsiemens/cm.
Test acceleration will be accomplished by subjecting the specimens to slow strain rate testing at a low rate (approximately 5x10' sec~ ) until failure. Prior to straining, specimens will be preconditioned for approximately seven days in the elevated temperature test environment.
Two specimens each of GPUN and NYPA heats of hot rolled XM-19 will be tested in the BWR coolant environment and one specimen each will be tested in air as a control. In addition, one specimen of sensitized 304SS will be tested in the test environment as a control to assure adequacy of the test environment to produce IGSCC.
Following the_ test, specimens will be examined using conventional light microscopy and scanning electron microscopy. The specimens will be examined for indications of stress corrosion cracking on the fracture surface and along the gauge section. A minimum of two metallographic mounts will be evaluated for each specimen.
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