[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