[Federal Register Volume 60, Number 66 (Thursday, April 6, 1995)]
[Notices]
[Pages 17590-17592]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 95-8575]
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NUCLEAR REGULATORY COMMISSION
[Docket No. 50-374]
Commonwealth Edison Co.; 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-18, issued to Commonwealth Edison Company (the licensee), for
operation of the LaSalle County Station, Unit 2, located in LaSalle
County, Illinois.
The proposed amendment would revise the safety/relief valve (SRV)
safety function lift setting allowable tolerance band from (-3% to +1%)
to (-3% to +3%) and include as-left SRV lift setting tolerances of (-1%
to +1%).
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.
Section 50.91(a)(6) of Title 10 of the Code of Federal Regulations
specifies that the Commission may, where exigent circumstances exist,
allow less than the 30 days for public comment. Exigent circumstances
have been found to exist for this proposed amendment. On March 18,
1995, with LaSalle Unit 2 in a shutdown condition for the current
refueling outage, the licensee learned that two of the six SRVs tested
had lift settings that were not within the current tolerance band
allowed by the technical specifications. This resulted in three
additional SRVs being tested and two additional SRVs found to lift at
pressures slightly outside the existing tolerance band. The remaining
nine SRVs are required to be tested based on the current technical
specifications. This testing would involve a significant financial
cost, the collection of approximately 11 person-rem of radiation
exposure by plant workers, and a delay in the restart of Unit 2. The
history of the safety relief value testing at LaSalle is such that the
licensee did not anticipate the immediate need for an increased
tolerance band. However, as part of a longer range plan to reduce the
number of SRVs and increase the allowable lift setting tolerances, the
licensee had performed much of the analyses required to justify the
proposed amendment request. On March 27, 1995, the licensee decided to
expedite the SRV lift setting technical specification change for
LaSalle Unit 2. The licensee completed the review and submitted the
request on March 31, 1995. To avoid the radiation exposures and restart
delays associated with testing the remaining nine SRVs, the proposed
amendment would need to be issued before April 22, 1995, and therefore
the request does not afford the normal 30-day comment period.
Pursuant to 10 CFR 50.91(a)(6) for amendments to be granted under
exigent circumstances, the NRC staff must determine 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 licensee has provided its analysts of
the issue of no significant hazards consideration. The staff has
reviewed the licensee's analysis against the standards of 10 CFR
50.92(c). The NRC staff's review is presented below.
1. The proposed change does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The probability of an accident previously evaluated will not
increase as a result of this change, because the only change are the
tolerances for the SRV opening setpoints and the speed of
[[Page 17591]] the reactor core isolation cooling system (RCIC) turbine
and pump. Changing the maximum allowable opening setpoint for the SRVs
does not cause any accident previously evaluated to occur, or degrade
valve or system performance in any way so as to cause an accident to
occur with an increased frequency. In addition, the increased speed of
the RCIC turbine and pump are within the design limits of the system.
RCIC operability and failure probabilities are not impacted by this
change.
The consequences of an ASME overpressurization event are not
significantly increased and do not exceed the previously accepted
licensing criteria for this event. GE has calculated the revised peak
vessel pressure for LaSalle Station to be 1341 psig, which is well
below the 1375 psig criterion of the ASME Code for upset conditions,
referenced in Section 5.2.2, Overpressurization Protection, of the
Updated Final Safety Analysis Report (UFSAR), and NUREG-0519 (Safety
Evaluation Report related to the operation of LaSalle County Station,
Units 1 and 2, March 1981), and Section 15.2-4, Closure of Main Steam
Isolation Valves (BWR) of NUREG-0800 (Standard Review Plan).
GE has also performed an analysis of the limiting anticipated
transient without scram (ATWS) event, which is the main steam isolation
valve (MSIV) closure event. This analysis calculated the peak vessel
pressure to be 1457 psig, which is well below the 1500 psig criterion
of the ASME Code for emergency conditions.
Per NUREG-0519, listed above, Section 5.4.1 and Technical
Specification 4.7.3.b, the RCIC pump is required to develop flow
greater than or equal to 600 gpm in the test flow path with a system
head corresponding to reactor vessel operating pressure when steam is
supplied to the turbine at 1000 +20, -80 psig. Increasing the turbine
and pump speed ensures these criteria will still be met and the
consequences of an accident will not increase.
Therefore, there is not a significant increase in the consequences
of an accident previously evaluated.
2. The proposed change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The only physical changes are to increase the allowable tolerances
for SRV opening setpoints and to increase the RCIC pump and turbine
speeds. These changes do not result in any changed component
interactions. The SRVs and RCIC will still provide the functions for
which they were designed. Since all of the systems evaluated will
continue to function as intended, the proposed changes do not create
the possibility of a new or different kind of accident from any
previously evaluated.
3. The proposed change does not involve a significant reduction in
the margin of safety.
While the calculated peak vessel pressures for the ASME
overpressurization event and the MSIV closure ATWS event are larger
than that previously calculated without the proposed setpoint tolerance
increases, the new peak pressures remain far below the respective
licensing acceptance limits associated with these events. These
licensing acceptance limits have been previously evaluated as providing
a sufficient margin of safety. For other accidents and transients, the
increased setpoint tolerances have a negligible, if any, effect on the
results, so the margin of safety is preserved.
Based on the 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 15 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 15-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 15-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. The
Commission expects that the need to take this action will occur very
infrequently.
Written comments may be submitted by mail to the Rules Review and
Directives Branch, Division of Freedom of Information and Publications
Services, Office of Administration, U.S. Nuclear Regulatory Commission,
Washington, DC 20555, and should cite the publication date and page
number of this Federal Register notice. Written comments may also be
delivered to Room 6D22, 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
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 May 8, 1995, the licensee 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 Public Library of Illinois Valley
Community College, Rural Route No. 1, Oglesby, Illinois 61348. 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 a 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 [[Page 17592]] 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 the amendment is issued before the expiration of the 30-day
hearing period, the Commission will make a final determination on the
issue of no significant hazards consideration. If a hearing is
requested, 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, Attention: Docketing and Services
Branch, or may be delivered to the Commission's Public Document Room,
the Gelman Building, 2120 L Street, NW., Washington, DC, by the above
date. Where petitions are filed during the last 10 days of the notice
period, it is requested that the petitioner promptly so inform the
Commission by a toll-free telephone call to Western Union at 1-(800)
248-5100 (in Missouri 1-(800) 342-6700). The Western Union operator
should be given Datagram Identification Number N1023 and the following
message addressed to Robert A. Capra, Director, Project Directorate
III-2: petitioner's name and telephone number, date petition was
mailed, plant name, and publication date and page number of this
Federal Register notice. A copy of the petition should also be sent to
the Office of the General Counsel, U.S. Nuclear Regulatory Commission,
Washington, DC 20555, and to Michael I. Miller, Espire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60690, attorney for
the licensee.
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 March 31, 1995, 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 Public Library of Illinois Valley
Community College, Rural Route No. 1, Oglesby, Illinois 61348.
Dated at Rockville, Maryland, this 4th day of April 1995.
For the Nuclear Regulatory Commission.
William D. Reckley,
Project Manager, Project Directorate III-2, Division of Reactor
Projects--III/IV, Office of Nuclear Reactor Regulation.
[FR Doc. 95-8575 Filed 4-5-95; 8:45 am]
BILLING CODE 7590-01-M
