[Federal Register Volume 63, Number 45 (Monday, March 9, 1998)]
[Notices]
[Pages 11460-11462]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 98-5946]
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NUCLEAR REGULATORY COMMISSION
[Docket No. 50-346]
Toledo Edison Company Centerior Service Company and the Cleveland
Electric Illuminating Company; Notice of Consideration of Issuance of
Amendment to Facility Operating License, Proposed No Significant
Hazards Consideration Determination, and Opportunity for a Hearing
The U.S. Nuclear Regulatory Commission (the Commission) is
considering issuance of an amendment to Facility Operating License No.
NPF-3 issued to the Toledo Edison Company, Centerior Service Company,
and The Cleveland Electric Illuminating Company (the licensees) for
operation of the Davis-Besse Nuclear Power Station, Unit No. 1, located
in Ottawa County, Ohio.
The application requests that tube repair roll, as described in
proprietary Framatome Technologies Incorporated Topical Report BAW-
2303P, Revision 3, ``OTSG Repair Roll Qualification Report,'' dated
October 1997, be included as a repair option for steam generator tube
defects in the upper tubesheet. The application further requests that
the pressure boundary joint be defined as the tube-to-tubesheet
expansion joint that is closest to the secondary face of the tubesheet.
Additionally, the application proposes several associated
administrative changes.
Before issuance of the proposed license amendment, the Commission
will have made findings required by the Atomic Energy Act of 1954, as
amended (the Act) and the Commission's regulations.
The Commission has made a proposed determination that the amendment
request involves no significant hazards consideration. Under the
Commission's regulations in 10 CFR 50.92, this means that operation of
the facility in accordance with the proposed amendment would not (1)
Involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. As
required by 10 CFR 50.91(a), the licensees have provided their analysis
of the issue of no significant hazards consideration, which is
presented below:
1a. Not involve a significant increase in the probability of an
accident previously evaluated because the proposed changes described
for Surveillance Requirements (SR) 4.4.5.2.a.1, SR 4.4.5.4.a.4, SR
4.4.5.4.a.6, SR 4.4.5.4.a.7, SR 4.4.5.4.b, SR 4.4.5.4.a.9, SR
4.4.5.5.b.3, and Table 4.4-2 add a repair process defined as
``repair roll'' and redefine the pressure boundary joint for a tube
repaired by the repair roll process. The application of the repair
roll process is limited to repairs in the upper tube sheet. The new
pressure boundary joint created by the repair roll process has been
shown by testing and analysis to provide structural and leakage
integrity equivalent to the original design and construction for all
normal operating and accident conditions. Furthermore, the testing
and analysis demonstrate the repair roll process creates no new
adverse effects for the repaired tube and does not change the design
or operating characteristics of the steam generators. Similarly, the
design and operating characteristics of the systems interfacing with
the steam generators are preserved by the repair roll process.
Accordingly, tubes repaired by the repair roll process will not
increase the probability of the tube rupture accident previously
analyzed.
The proposed change to SR 4.4.5.3.c.1 and the proposed addition
of SR 4.4.5.9 define additional required inspections for the primary
system to secondary system joints created by the repair roll
process. The addition of this inspection does not change any
accident initiators and, therefore, does not increase the
probability of an accident previously evaluated.
The proposed change to Limiting Condition for Operation (LCO)
3.4.6.2.c reduces the maximum allowed primary-to-secondary leakage
through the steam generators from 1 gallon per minute (1440 GPD) to
150 GPD through any one steam generator. The reduction in allowed
primary-to-secondary leakage does not change any accident initiators
and, therefore, does not increase the probability of an accident
previously evaluated.
The proposed additional requirements of SR 4.4.6.2.1.e describe
the method and frequency that will be used for monitoring the
reduced leakage limit. This additional monitoring of primary to
secondary leakage through the steam generators does not change any
accident initiators and, therefore, does not increase the
probability of an accident previously evaluated.
The proposed changes to Bases B 3/4.4.5 add reference to the
repair roll method and change the description of the allowed primary
to secondary leakage through the steam generators to the reduced
limit of 150 GPD through any one steam generator. It is noted that
in Bases 3/4.4.5 the leakage limit established is defined as an
inservice indicator of the structural integrity of the tubes. The
reduction in the allowed primary to secondary leakage continues to
provide inservice indication of tube structural integrity such that
adequate margins of safety exist to withstand the loads imposed by
normal operations and postulated accidents. Each of these changes to
the Bases does not change any accident initiators and, therefore,
does not increase the probability of an accident previously
evaluated.
The proposed changes to Bases 3/4.4.6.2 also change the
description of the maximum allowed primary-to-secondary leakage to
the lowered limit of 150 GPD through any one steam generator. The
reduction of allowed primary-to-secondary leakage does not increase
the probability of an accident previously evaluated.
The proposed changes to SR 4.4.5.2.a and SR 4.4.5.3.a are
administrative changes and do not affect the probability of
accidents previously evaluated.
1b. Not involve a significant increase in the consequences of an
accident previously evaluated because the proposed changes described
for SR 4.4.5.2.a.1, SR 4.4.5.4.a.4, SR 4.4.5.4.a.6, SR 4.4.5.4.a.7,
SR 4.4.5.4.b, SR 4.4.5.4.a.9, SR 4.4.5.5.b.3, and Table 4.4-2 add a
repair process defined as ``repair roll'' and redefine the pressure
boundary joint for a tube repaired by the repair roll process. The
application of the repair roll process is limited to repairs in the
upper tube sheet. The new pressure boundary joint created by the
repair roll process has been shown by testing and analysis to
provide structural and leakage integrity equivalent to the original
design and construction for all normal
[[Page 11461]]
operating and accident conditions. Furthermore, the testing and
analysis demonstrate the repair roll process creates no new adverse
effects for the repaired tube and does not change the design or
operating characteristics of the steam generators. Similarly, the
design and operating characteristics of the systems interfacing with
the steam generators are preserved by the repair roll process.
Accordingly, tubes repaired by the repair roll process will not
increase the consequences of an accident previously analyzed. At
worst, tubes repaired by the repair roll process will result in
primary-to-secondary leakage. Should a tube leak occur, it would be
bounded by the steam generator tube rupture accident consequences,
which have been analyzed previously.
The proposed change to SR 4.4.5.3.c.1 and the proposed addition
of SR 4.4.5.9 define additional required inspections for the primary
system to secondary system joints created by the repair roll
process. The addition of this inspection requirement does not
increase the consequences of an accident previously evaluated.
The proposed change to LCO 3.4.6.2.c reduces the maximum allowed
primary-to-secondary leakage through the steam generators from 1440
GPD to 150 GPD through any one steam generator. This change provides
additional conservatism in the operation of the DBNPS and does not
increase the consequences of an accident previously evaluated.
The proposed additional requirements of SR 4.4.6.2.1.e describe
the method that will be used for monitoring the reduced leakage
limit. This additional method of monitoring primary to secondary
leakage through the steam generators does not change any accident
and, therefore, does not increase the consequences of any accident
previously evaluated.
The proposed changes to Bases B 3/4.4.5 add reference to the
repair roll method and change the description of the allowed primary
to secondary leakage through the steam generators to the reduced
limit of 150 GPD through any one steam generator. It is noted that
in Bases 3/4.4.5 the leakage limit established is defined as an
inservice indicator of the structural integrity of the tubes. The
reduction in the allowed primary to secondary leakage continues to
provide inservice indication of tube structural integrity such that
adequate margins of safety exist to withstand the loads imposed by
normal operations and postulated accidents. These changes to the
Bases do not change any accident and, therefore, will not increase
the consequences of any accident previously evaluated.
The proposed changes to Bases 3/4.4.6.2 also change the
description of the maximum allowed primary-to-secondary leakage to
the lowered limit of 150 GPD through any one steam generator. The
reduction of allowed primary-to-secondary leakage does not increase
the consequences of any accident previously evaluated.
The changes to SR 4.4.5.2.a and SR 4.4.5.3.a are administrative
changes and do not affect the consequences of accidents previously
evaluated.
2. Not create the possibility of a new or different kind of
accident from any accident previously evaluated because there is no
change in the operation of the steam generators or connecting
systems with the repair roll process added by the proposed changes
in SR 4.4.5.2.a.1, SR 4.4.5.4.a.4, SR 4.4.5.4.a.6, SR 4.4.5.4.a.7,
SR 4.4.5.4.a.9, SR 4.4.5.4.b, SR 4.4.5.5.b.3 and Table 4.4-2. The
physical changes in the steam generators associated with the repair
roll process have been evaluated and do not create the possibility
for a new or different kind of accident from any accident previously
evaluated, i.e., the physical change in the steam generators is
limited to the location of the primary to secondary boundary within
the tubesheet and does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The reduction in maximum allowed primary-to-secondary leakage
defined by the proposed change to LCO 3.4.6.2.c does not create the
possibility of a new or different kind of accident from any
previously evaluated accident. The additional testing of tubes
repaired by the repair roll process as required by the proposed
change to SR 4.4.5.3.c.1 and the addition of SR 4.4.5.9 does not
create the possibility of a new or different kind of accident from
any previously evaluated accident. Similarly, the monitoring of
primary to secondary leakage as specified in the proposed SR
4.4.6.2.1.e does not create the possibility of a new or different
kind of accident from any previously evaluated accident.
The proposed changes to Bases 3/4.4.5 and 3/4.4.6.2 reflect the
changes proposed to their associated LCOs and SRs, and are not
involved with any accident. The changes made to SR 4.4.5.2.a and SR
4.4.5.3.a are administrative changes and do not create the
possibility of new or different kinds of accidents from any accident
previously evaluated.
3. Not involve a significant reduction in a margin of safety
because all of the protective boundaries of the steam generator are
maintained equivalent to the original design and construction with
tubes repaired by the repair roll process. Furthermore, tubes with
primary system to secondary system boundary joints created by the
repair roll have been shown by testing and analysis to satisfy all
structural, leakage, and heat transfer requirements.
The additional testing of tubes repaired by the repair roll
process provides continuing inservice monitoring of these tubes such
that inservice degradation of tubes repaired by the repair roll
process will be detected. Therefore, the changes to SR 4.4.5.2.a.1,
SR 4.4.5.4.a.4, SR 4.4.5.4.a.6, SR 4.4.5.4.a.7, SR 4.4.5.4.b, SR
4.4.5.5.b.3 and Table 4.4-2 to add repair roll as a repair process
do not reduce a margin of safety. Similarly, the proposed change to
SR 4.4.5.4.a.9 to redefine the pressure boundary for a tube with a
repair roll is based upon eddy current testing demonstrating the
adequacy of the repair roll to provide this pressure boundary and
maintain the present margin of safety.
The proposed reduction of allowed primary to secondary leakage,
as defined in the changes to LCO 3.4.6.2.c, constitutes additional
conservatism in the operation of the DBNPS and does not reduce a
margin of safety. Similarly, the additional testing and monitoring
defined in the changed SR 4.4.5.3.c.1 and the proposed SR 4.4.5.9
and SR 4.4.6.2.1.e constitute additional conservatism in the
operation of the DBNPS and do not reduce a margin of safety.
The proposed changes to Bases \3/4\.4.5 and \3/4\.4.6.2 reflect
the changes pro posed to their associated LCOs and SRs, and do not
reduce a margin of safety.
The changes to SR 4.4.5.2.a and SR 4.4.5.3.a are administrative
changes and do not reduce the margin of safety.
The NRC staff has reviewed the licensees' analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received.
Should the Commission take this action, it will publish in the Federal
Register a notice of issuance and provide for opportunity for a hearing
after issuance. The Commission expects that the need to take this
action will occur very infrequently.
Written comments may be submitted by mail to the Chief, Rules and
Directives Branch, Division of Administrative Services, Office of
Administration, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and should cite the publication date and page number of
this Federal Register notice. Written comments may also be delivered to
Room 6D59, Two White Flint North, 11545 Rockville Pike, Rockville,
Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of
written comments received may be examined at the NRC Public
[[Page 11462]]
Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC.
The filing of requests for hearing and petitions for leave to
intervene is discussed below.
By April 8, 1998 the licensees may file a request for a hearing
with respect to issuance of the amendment to the subject facility
operating license and any person whose interest may be affected by this
proceeding and who wishes to participate as a party in the proceeding
must file a written request for a hearing and a petition for leave to
intervene. Requests for a hearing and a petition for leave to intervene
shall be filed in accordance with the Commission's ``Rules of Practice
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested
persons should consult a current copy of 10 CFR 2.714 which is
available at the Commission's Public Document Room, the Gelman
Building, 2120 L Street, NW., Washington, DC, and at the local public
document room located at the University of Toledo, William Carlson
Library, Government Documents Collection, 2801 West Bancroft Avenue,
Toledo, OH 43606. If a request for a hearing or petition for leave to
intervene is filed by the above date, the Commission or an Atomic
Safety and Licensing Board, designated by the Commission or by the
Chairman of the Atomic Safety and Licensing Board Panel, will rule on
the request and/or petition; and the Secretary or the designated Atomic
Safety and Licensing Board will issue a notice of hearing or an
appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) The nature of the petitioner's right under the
Act to be made party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: Rulemakings and
Adjudications Staff, or may be delivered to the Commission's Public
Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC,
by the above date. A copy of the petition should also be sent to the
Office of the General Counsel, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, and to Jay E. Silberg, Esquire, Shaw,
Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC
20037, attorney for the licensees.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for hearing will not
be entertained absent a determination by the Commission, the presiding
officer or the presiding Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of the
factors specified in 10 CFR 2.714(a)(1)(I)-(v) and 2.714(d).
For further details with respect to this action, see the
application for amendment dated February 26, 1998, which is available
for public inspection at the Commission's Public Document Room, the
Gelman Building, 2120 L Street, NW., Washington, DC, and at the local
public document room located at the University of Toledo, William
Carlson Library, Government Documents Collection, 2801 West Bancroft
Avenue, Toledo, OH 43606.
Dated at Rockville, Maryland, this 3d day of March 1998.
For the Nuclear Regulatory Commission.
William O. Long,
Senior Project Manager, Project Directorate III-3, Division of Reactor
Projects--III/IV, Office of Nuclear Reactor Regulation.
[FR Doc. 98-5946 Filed 3-6-98; 8:45 am]
BILLING CODE 7590-01-P
