ML17264A831
| ML17264A831 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 03/10/1997 |
| From: | Mecredy R ROCHESTER GAS & ELECTRIC CORP. |
| To: | Vissing G NRC (Affiliation Not Assigned), NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9703190117 | |
| Download: ML17264A831 (6) | |
Text
CATEGORY 1 REGULATOr INFORNATION DISTRIBUTION STEM (RIDS)
ACCFSSIQN,NBR: 9703190117 DOC. DATE: 97/03/10 NOTARIZED:
NO DOCKET FACIL: 50-244 Robert Emmet Ginna Nuc lear Plantr Unit ii Rochester G
05000244 AUTH. NARE AUTHOR AFFILIATION NECREDYi R. C.
Rochester Gas Zc Electric Corp.
REC IP. NANE RECIPIENT AFFILIATION VISSINGj G.
SUBJECT:
Forwards relief request 32 For review Zc action. Request has been developed to address technical limitations in ability to perform inner radius exams on regenerative heat exchanger due to size 5 configuration of nozzles to be examined.
DISTRIBUTION CODE:
A047D COPIES RECEIVED: LTR ENCL SIZE:
TITLE:
OR Submitt'al:
Inservice/Testing/Relief from ASNE Code GL-S9-04 NOTES: License Exp date in accordance with 10CFR2i 2. 109(9/19/72).
05000244
.RECIPIENT ID CODE/NANE PD1-1 LA VISS INGa G.
INTERNAL: *EOD/SPD/RAB NRR/DE/ENEB OGC/HDS3 RES/DET/ENNEB COPIES LTTR ENCL 1
1 1
1 1
1 1
1 1
0 1
1 RECIPIENT ID CODE/NANE PD1-1 PD ILE CENTER 0
N
-ABSTRACT RES/DET/EIB COPIES LTTR ENCL 1
1 0
EXTERNAL: LITCO ANDERSON NRC PDR 1
1 1
NOAC D
U E
N NOTE TO ALL "RIDS" RECIPIENTS:
PLEASE HELP US TO REDUCE WASTE!
CONTACT THE DOCUMENT CONTROL DESK, ROOM OWFN SD-5(EXT. 415-2083)
TO ELIMINATE YOUR NAME FROM DISTRIBUTION LISTS FOR DOCUMENTS YOU DON'T NEED!
TOTAL NUNBER OF COPIES REQUIRED:
LTTR 13 ENCL 12
AND ROCHESTER GASANDEIECTRIC CORPORATION ~ 89 EASTAVENUE, ROCHESTER, N. Y 14649 0001 AREACODE716 Sd6-27M ROBERT C. MECREDY Vice President Nuclear Operations March 10, 1997 U.S. Nuclear Regulatory Commission Document Control Desk Attn:
Guy Vissing Project Directorate I-1 Washington, D.C.
20555
Subject:
Relief Request No.
32 Ginna Nuclear Power Plant Inservice Inspection ASME Section XI Required Examinations R.E. Ginna Nuclear Power Plant Docket No. 50-244
Dear Mr. Vissing:
The purpose of this letter is to transmit the attached Relief Request for NRC review and action.
The request has been developed to address technical limitations in the ability to perform inner radius examinations on the regenerative heat exchanger due to the size and configuration of the nozzles to be examined.
Your action on this item is requested to support the refueling outage due in
- October, 1997.
Very truly yours, Robert C.
Me ed REJE453 xc:
Mr. Guy Vissing (Mail Stop 14C7)
Project Directorate I-1 Washington, D.C.
20555 U.S. Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia, PA 19406 Ginna Senior Resident Inspector
'77031'F0117 F70310 PDR ADQCK 05000244 P
l Et
-I t
l
Relief Request No.
32 Regenerative Heat Exchanger (RHE), Inner Radius Examinations I.
Components for Which Relief is Requested:
Chemical and Volume Control Systems Regenerative Heat Exchanger (RHE), Class 1 Inner Radius Examinations; RHE-N1 RHE-N5 II.
ASME Requirements for Which Relief is Requested:
In accordance with Table IWB-2500-1, Examination Category B-D, Item Number B3.160.
III. Basis:
The Regenerative Heat Exchanger (RHE) consists of three (3) shell and tube heat exchangers connected in series.
The RHE is designed to recover heat from the reactor coolant system letdown stream during normal operation.
The letdown stream flows through the shell side of the heat exchanger.
The shell side of the RHE is Class 1 while the tube side is Class 2.
The RHE provides the major single source of radiation exposure accumulated during a
normal refueling outage inservice inspection.
Inner Radius examinations were scheduled to be performed on Class 1 side of the bottom heat exchanger that experience the most extreme conditions as specified by our approved Relief Requests (18-1 and 18-2).
The RHE inner radius examinations were to be performed on small heavy wall nozzles that are connected to 2" piping.
An indepth investigation was initiated by RG&E to determine the feasibility of performing an acceptable Code examination.
The initial investigation reviewed the nozzle type, weld placement and actual OD weld profiles as well as ultrasonic measurements to verify ID configuration.
To assist in the evaluation of performing an acceptable Code ultrasonic inner radius examination, Computer Modeling and mockups of the nozzle to vessel configuration were initiated.
Computer modeling was performed by SouthWest Research Institute, AEA Technology and EPRI.
Computer modeling performed by the various organizations compared favorably.
The computer modeling initiative indicated that several different transducers would be required to be used and that the inner radius examination results would be questionable at best due to the size and configuration of the nozzles.
The modelling also indicated that beam spread and mode conversion at the notches and neighboring surfaces
would seriously reduce the signal to noise ratios, causing confusing spurious signals.
Based upon the computer modeling results, EPRI NDE Center personnel were utilized to perform actual "hands-on" inner radius examination evaluations on the mockups.
An area was selected on both mockups of the nozzle and suitable transducers and wedges were selected to perform the examination.
The inspection was performed from the boss region of the nozzle because inspection from the shell surface proved to be greatly affected by attenuation and.scattering from the nozzle-to-shell weld material.
A variety of inspection frequencies were attempted of which none provided what was considered successful for the detection of the notches on these nozzles.
It should be noted that the transducer positions for detecting the selected notch was nearly optimum.
Since the attempts made were unsuccessful it was decided to increase the depth of the notch from 10'to 304.
The increase of the notch is greater than code allowable.
Attempts were made on the greater notch depth but detection was not achievable.
In conclusion, it has been proven that an acceptable Code examination of the RHE nozzle inner radius region is not possible utilizing current technology.
The testing demonstrates that signals from different. depth notches located in the most optimum position of the mock-up could not be differentiated from noise and geometric reflectors without the aid of finger damping on the ID surface.
The evaluations have also shown that the limitation in the boss area, in the 0 and 180 degree position, along with the transition to these areas, limits the size of transducers that can be employed and where the transducers can be placed.
IV.
Proposed Alternate Method:
None.
RG&E will continue to evaluate new emerging inspection technology as they become available.
The Code required leakage tests with associated VT-2 examinations shall be performed.
References:
2 ~
EPRI Report Title, "Evaluation of Ultrasonic Examination Technology for Inspection of Regenerative Heat Exchanger Nozzles at the Rochester Gas
& Electric, Ginna Nuclear Plant",
dated April 1996 by Douglas E. MacDonald and E. Kim Kietzman.
AEA Reactor Services Report Title, "Report on the Mathematical Modeling of Two Nozzle Specimens for Rochester Gas
& Electric Corporation", dated November 1992 by P.
D, Birchall and L. Z.
Poulter.
0 0
~
v
%i ll
~oi l ~ II 1
gl