Protecting People and the EnvironmentUNITED STATES NUCLEAR REGULATORY COMMISSION
GL80046
Distribution
May 19 1980 Docket ACRS (16)
NRC PDR CR BCs
LOCAL PDR L;ADs
NRR Reading L:LAs
TIppolito DEisenhut
SNorris RPurple
ALL POWER REACTOR LICENSEES Jolshinski RSnaider
Atty, OELD (5) VNoonan
OI&E (5) KWichman
NSIC DSellers
TERA TNovak
In November 1979, you were sent, for your review and comment, the "For
Comment" edition of NUREG-0577, "Potential for Low Fracture Toughness and
Lamellar Tearing on PWR Steam Generator and Reactor Coolant Pump Supports".
The report provides the staff's resolution of the NRC's Generic Technical
Activity A-12, which is an "Unresolved Safety Issue" pursuant to Section 210
of the Energy Reorganization Act of 1974.
The generic study resulted from questions raised during the licensing review
of two pressurized water reactors (PWRs). The specific concern was the
capability of the supports to maintain their structural integrity under
severe environmental and accident conditions.
NUREG-0577 describes the technical issues, the technical studies performed
by an NRC consultant, the NRC staff's tentative plans for implementing its
technical positions. It also provides recommendations for further generic
research into the subject of lamellar tearing. The Electric Power Research
Institute has been requested to conduct such research.
Comments on the report, including the proposed review procedure and
implementation schedule contained in this letter, are being solicited from
interested organizations, groups and individuals. These comments will be
evaluated by the NRC staff prior to final implementation of this subject.
all comments should be forwarded to Mr. Richard Snaider, Generic Issues
Branch, Nuclear Regulatory Commission, Washington, D.C. 20555, by July 7,
1980.
At the completion of the 60 day period, the staff will evaluate the
components received and, if needed, will issue a supplement or revision to
NUREG-0577. The target for issuing the supplement or revision is late
September 1980.
Subsequent to issuance of NUREG-0577, the NRC has decided to modify
significantly the implementation plan presented in Section 4 of NUREG-0577.
Basically, we propose that licensees be tasked with demonstrating the
adequacy of the support structures of their facility(ies) from a fracture
toughness standpoint. They will then be required to submit a report to the
NRC describing in detail the conclusions drawn and any action taken or
planned. The following steps will be required under the proposed
implementation plan:
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May 19 1990
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1. Licensees of operating PWRs should refer to Section 3 (pages 9 and 10)
of NUREG-0577 to ascertain their classification with regard to
potential susceptibility to low fracture toughness. Plants not reviewed
during the generic study are already undergoing a complete review by
the Franklin Research Center (FRC). This includes Arkansas Nuclear One
Unit 2. San Onofre Unit 1, and Turkey Point Units 3 and 4. Indian Point
Units 2 and 3, which are Group I plants, are included in the FRC
review. The NRC technical contact for the FRC review is C. D. Sellers.
2. These licensees in Group III are not required to take any further
action with regard to steam generator and reactor coolant pump supports
and may consider this matter resolved. Evaluation of other supports
should be performed as set forth in Step 6.
3. These licensees in Group II are not required to take any action at this
time with regard to steam generator and reactor coolant pump supports.
Future action on these supports may be necessary based on information
received from the reports of Group I licensees. The NRC staff will make
this determination and inform Group II licensees if any additional
action is necessary for those classes of equipment. However, these
licensees should implement Step 6 for action concerning other supports.
Group I licensees, with the exception of Indian Points Units 2 and 3,
are to undertake plant specific evaluation. The specific questions of
Appendix D to NUREG-0577 are to be disregarded. The materials/parts
listed below are to be analyzed in accordance with the General
Operating Reactor Review Procedure enclosed as Attachment 1.
A. Crystal River Unit 3: A-515 in the flange and material of
steam generator skirts; material used in the upper steam generator
supports
B. Davis-Besse Unit 1: A-36; A-516, A-515 and A-53 in steam generator
lower lateral supports.
C. J. M. Farley Units 1 and 2: Carpenter Custom 455 steel bolts used
in Clevis attachments of the vertical columns.
D. Fort Calhoun: A-307 nuts and bolts.
E. Kewanee: 250 CYM 0.5-inch diameter "Heli-coil screws into S.
G." and 1.0-inch diameter "upper support ring girder wall bolts"
F. Maine Yankee: Steam generator base castings.
G. Millstone Unit 2: A-106 and A-515 steel in pump supports.
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H. Palisades: A-212 in the base flange of the steam generator
supports.
I. Point Beach Units 1 and 2: 12-inch diameter A-53 Schedule 100 pip
columns.
J. Prairie Island Units 1 and 2: 250 CYM 0.5-inch diameter "heli-
coil screws into S. G." and 1.0-inch diameter "upper support ring
girder wall bolts".
K. Rancho Seco: Materials used in coolant pump horizontal supports;
A-515 base flange on steam generator support skirts; materials
used in steam generator upper horizontal restraints.
L. St. Lucie Unit No. 1: A-515 in coolant pump snubber clevises; A-
27 in steam generator base castings.
M. Surry Units 1 and 2: Vascomax 300 and 350 steels in all location;
A-106 pipes; A-235 plants; A-105 pipe and forgings.
N. Three Mile Island Unit 1: Materials of coolant pump horizontal
supports A-515 base flange on steam generator horizontal
restraints.
O. Yankee Rowe: Materials used in upper part of steam generator
support structures; material used in coolant pump hanger rod
supports; A-7 and C-1020 steels.
If the support materials cannot be shown to have adequate fracture
toughness or the capability to withstand stress corrosion cracking,
licensees must immediately inform the applicable regional office of the
NRC's Office of Inspection and Enforcement (OIE), with copies to
Director, OIE and Director, Office of Nuclear Reactor Regulation. In
this report, licensees are to recommend appropriate action and provide
a schedule for such action. All Group I licensees, including Indian
Point Units 2 and 3, must also perform the evaluations of additional
supports as set forth in Step 6 below.
5. PWR licensees whose plants were not included in NUREG-0577 must also
perform the review of steam generator and reactor coolant pump
supports. These licensees should begin with the materials
classification of Table 4.6 of Appendix C to NUREG-0577 (page C-38) and
proceed with the evaluation call for in Attachment 1. If adequate
fracture toughness or the ability to withstand stress corrosion
cracking (where applicable) cannot be demonstrated, the licensee must
immediately inform the applicable regional office of NRC's Office of
Inspection and Enforcement (OIE), with copies to Director, OIE, and
Director, Office of Nuclear Reactor Regulation. In this report,
licensees
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are to recommend appropriate action and provide a schedule for such
action. If any support contains a material not evaluated by (Table
4.5 of Appendix C), the licensee must perform an evaluation of this
material and demonstrate the adequacy of its toughness in the
particular application. These licensees must also perform the
evaluation of additional supports as set forth in Step 6.
6. Licensees of boiling water reactors (BWRs) PWRs must review the
material of applicable major component supports not included in the
NUREG-0577 review and determine if the analysis of Attachment 1 is
called for. For example, reactor vessel support members of material
listed in Group I of Table 4.5 (page C-38) of NUREG-0577 certainly
should be reviewed. These supports are:
BWR
Reactor vessel
Reactor coolant recirculation pumps (non-pipe supported)
PWR
Pressurizer
Reactor vessel
The review method of Step 5 above is to be utilized for those materials
deemed by the licensee to require analysis. The report at the
completion of this review (see 7. below) must present the results of
such analysis or the rationale behind exemption of support materials
from analysis.
7. The date for completion of the evaluations, and the submittal of
detailed reports discussing methods used, results of evaluations, and
subsequent actions to resolve problems, is December 31, 1981. As with
the reports discussed above, submittal should be made to the Director
of the Regional Office, OIE, with copies to Director, OIE, and
Director, Office of Nuclear Reactor Regulation. If any support material
was evaluated to have inadequate fracture toughness, the licensee must
state what appropriate action has been taken or provide a schedule for
such action. The NRC staff will use the information of these reports to
determine if any additional action is necessary, in particular on Group
II plants. The information will also be used to determine if inservice
inspection of supports is necessary.
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Because of the importance of resolving this issue, plants undergoing
Systematic Evaluation Program (SEP) review must be reviewed and any
necessary changes implemented in the same time frame as non-SEP plants.
Sincerely,
Darrell G. Eisenhut, Director
Division of Licensing
Office of Nuclear Reactor Regulation
Attachment:
"General Operating Reactor
Review Procedure and
Acceptance Criteria"
cc w/attachment:
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