ML19331B640
| ML19331B640 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood, LaSalle  |
| Issue date: | 08/01/1980 |
| From: | Naughton W COMMONWEALTH EDISON CO. |
| To: | Snaider R Office of Nuclear Reactor Regulation |
| References | |
| REF-GTECI-A-12, REF-GTECI-ES, RTR-NUREG-0577, RTR-NUREG-577, TASK-A-12, TASK-OR NUDOCS 8008120526 | |
| Download: ML19331B640 (2) | |
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Commonwealth Edison One First National Plaza. Chicago, Illinois Address Reply to: Post Office Box 767 Chicago, lilinois 60690 August 1, 1980 Mr. Richard P. Snaider l
Generic Issues Branch U.S. Nuclear Regulatory Commisson Washington, DC ~ 20555
Subject:
Comments on NUREG-0577,
" Potential for Low Fracture Toughness and Lamellar Tearing on PWR Steam Generators and Reactor Coolant
_ Pump Supports Reference (a):
May 20, 1980 letter from D. G. Eisenhut to All Pending Operating Licensees and Construction Permit Applicants and All Licensees of Plants under Construction
Dear Mr. Snaider:
Commonwealth Edison Company is pleased to submit the following comments in response to the subject NUREG regarding the reactor coolant pumps, reactor vessels, pressurizer and steam generators for its Byron, Braidwood and LaSalle Stations.
Specifically, (1) The Charpy-V-Notch test and Acceptance Criteria in the proposed evaluation are different than ASME Section NF 2300 criteria.
These supports were designed to Section III of the ASME Code.
Per NF 2300, the Charpy-V-Notch (CVN) test is permitted for all material thickness, and the acceptance criteria are given in terms of mils lateral expansion, which is considered to be a better indication of fracture toughness.
(2) The proposed evaluation procedure should allow for plant specific analysis and plant specific maximum NOTT (Nil Ductility Transition Temperature) much in the same vein as the original implementation plan contained in Section 4.0 o f NUREG-0577.
Since different plants have different operating temperatures in the support region, an acceptance criteria of 750F for NOTT is not well understood.
/8 (3) The stress corrot on cracking for materials with I
minimum yield strength greater than 120 KSI is
)
inconsistent with the Sandia Report (Page C-23).
The report states that "...as long as the specified yield strength is less than 180 KSI, this problem is not 8008120 Mb
)
Mr. Richard P. Snaider August 1, 1980 Page 2 considered to be present."
The 120 KSI yield level will preclude the use of pretensioned A490 bolts which have a yield strength of 130 KSI.
Commonwealth Edison and its A-E, Sargent & Lundy, recommend that 180 KSI be established as the minimum yield strength used to determine the susceptibility of a material to stress corrosion cracking.
Such a change would be consistent with the Sandia Report.
The acceptance of higher strength materials will also be consistent with NRC Regulatory Guide 1.85, Code Case 1644.
Commonwealth Edison also endorses comments from EPRI on tha proposed implementation plan which include:
(1) Ancillary heating of the supports as a proposed cure assuming that adequate fracture toughness is missing, is expensive, troublesome and is not always effective.
(2) Some materials that do not pass the proposed evaluation cannot be replaced such as embedded bolting.
(3) Improper interpretation of a radiation damage model could cause needlessly expensive analysis of reactor vessel supports.
(4) Many of the support structures in question see only compression loading; however, the failure mode assumes tensile loading.
Commonwealth Edison Company appreciates the opportunity to submit these comments and is hopeful that these comments will be of assistance to the Commission in its evaluation.
Very truly yours, W. F. Naugh n
Nuclear Licensing Administrator Pressurized Water Reactors Commonwealth Edison Company 5727A a.m
-oiaew