[Federal Register Volume 61, Number 23 (Friday, February 2, 1996)]
[Notices]
[Pages 3951-3953]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 96-2206]
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NUCLEAR REGULATORY COMMISSION
[Docket No. 50-440]
The Cleveland Electric Illuminating Company, et al.; 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-58 issued to The Cleveland Electric Illuminating Company, et al.
(the licensee), for operation of the Perry Nuclear Power Plant, Unit
No. 1 located in Lake County, Ohio.
The proposed amendment would change the Technical Specification
surveillance frequency for the drywell bypass leakage rate test from 18
months to 120 months (10 years) with a more frequent testing
requirement if performance degrades. Additionally, specific leakage
limits would be deleted for the air lock seal and barrel tests. Also,
surveillance frequencies for the air lock interlock test and seal
pneumatic system leak test would be changed from 18 months to 24
months. Finally, the surveillance frequencies for the air lock barrel
test would be changed from ``each COLD SHUTDOWN if not performed within
the previous 6 months'' to ``at least once per 24 months'' and from 18
months to 24 months. The licensee requested that this amendment be
approved for use during the current refueling outage which began on
January 27, 1996.
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 licensee has provided its analysis of
the issue of no significant hazards consideration, which is presented
below:
I. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed changes in frequency for the drywell bypass leakage
and drywell air lock surveillances will continue to ensure that no
paths exist through drywell boundary components that would permit
gross leakage from the drywell to bypass the containment pressure-
suppression feature (the suppression pool) and result in exceeding
the primary design basis limit. The Mark III primary containment
system satisfies General Design Criterion 16 of Appendix A to 10 CFR
Part 50. Maximum drywell bypass leakage was determined previously by
reviewing the full range of postulated primary system break sizes.
The limiting case was a primary system small break LOCA that yielded
a design allowable drywell bypass leakage rate limit of
approximately 58,000 scfm. The Technical Specification acceptable
limit for the bypass leakage following a surveillance is less than
10% of the design basis value. The most recent bypass leakage value
was approximately 0.2% of the design allowable leakage rate limit
for the limiting event. Programmatic and oversight controls are
maintained that ensure drywell bypass leakage remains a fraction of
the design allowable leakage limit.
The drywell is exposed to essentially 0 psig during normal plant
operation and 2.5 psig during drywell bypass leak rate testing.
These pressures are considerably lower than the structural integrity
test pressure and are not likely to initiate a crack or cause an
existing crack to grow. Visual inspections of the accessible drywell
surfaces that have been performed since the structural integrity
tests have not revealed the presence of abnormal cracking or other
abnormalities. Therefore, drywell degradation is not expected due to
testing or operation and it is not considered credible for the
passive drywell structure to begin to leak sufficiently to impact
the design drywell bypass leakage limit.
The primary containment's ability to perform its safety function
is fairly insensitive to the amount of drywell bypass leakage,
thereby providing a margin to loss of the drywell safety function
that is not normally available for safety systems. This
insensitivity is demonstrated by the extremely high limiting event
design basis allowable leakage for the drywell (approximately 58,000
scfm as discussed above). An even higher allowable leakage can be
accommodated by the primary containment due to containment design
margin. It would take valves in multiple penetration flow paths
leaking excessively to cause the primary containment to fail as a
result of overpressurization. Therefore, the probability that
drywell isolation valve leakage will result in primary containment
failure due to excessive drywell bypass leakage is not significant
and this drywell/primary containment failure mode is not credible.
The proposed Technical Specification changes have no significant
impact on the IPE conducted in accordance with NRC Generic Letter
88-20. The IPE considered
[[Page 3952]]
primary containment overpressurization failure as part of the primary
containment performance assessment. Due to the magnitude of
acceptable drywell bypass leakage and the extremely low
probabilities of experiencing excessive leakage, preexisting
excessive drywell bypass leakage was considered a non-significant
contributor to primary containment failure. In a beyond-design-basis
``severe accident,'' the surveillance frequencies for the air lock
failure can occur with or without preexisting excessive drywell
bypass leakage. This is due to physical phenomena associated with
potentially extreme environmental conditions inside primary
containment following a severe accident. However, the calculated
frequency of such extreme conditions is very small. The proposed
changes do not impact the IPE evaluated phenomena causing primary
containment overpressurization failure and do not significantly
increase the probability that the drywell has preexisting excessive
leakage. The proposed changes therefore, would not contribute to
these accident scenarios.
The movement of the air lock leakage rate tests to the Drywell
Specification and the elimination of the Notes in the Improved
Technical Specifications are proposed because drywell leakage rate
requirements are the essence of drywell operability. Leakage rates
discovered outside limits will always clearly result in entering the
actions for drywell inoperability. Additionally, the requirements
for the drywell air lock seal and barrel tests to meet specific
leakage limits are deleted since the ability of the drywell to
perform its safety function is not dependent on the air lock meeting
a specific leakage limit. The limiting case for drywell bypass
leakage is based on total leakage through all drywell paths other
than the suppression pool vents. Total drywell bypass leakage from
such paths (including the air lock) should not exceed the acceptable
design limit of drywell bypass leakage. The proposed Technical
Specifications will still require performance of seal and barrel
leak tests. Additionally, the proposed changes include minor
administrative changes which clarify the requirement format or
change the requirement to match the plant design bases.
For the reasons discussed above, the proposed changes do not
have any significant risk impact to accidents previously evaluated
and do not significantly increase the consequences of an accident
previously evaluated. Additionally, drywell bypass leakage is not
the initiator of any accident evaluated; therefore, changes in the
frequency of the surveillance for drywell bypass leakage does not
increase the probability of any accident evaluated.
II. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed changes will impact the test frequencies and will
not result in any change in equipment response in the unlikely event
of an accident. The changes do not alter equipment design or
capabilities. The changes do not present any new or additional
failure mechanisms. The drywell is passive in nature and the
surveillance will continue to verify that its integrity has not
degraded. Therefore, the proposed change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
III. The proposed change does not involve a significant
reduction in a margin of safety.
Drywell integrity and reliability have been demonstrated during
past drywell bypass leakage surveillances. Appropriate design basis
assumptions will be maintained. Drywell integrity will continue to
be tested by the proposed periodic drywell bypass leakage test, the
drywell air lock door latching and interlock mechanism surveillance,
and additional surveillances including exercising the drywell
isolation valves. In combination, these surveillances will provide
adequate assurance that drywell bypass leakage will not exceed the
design basis limit. Margins of safety will not be reduced.
Therefore, the proposed change does not cause a reduction in the
margin of safety.
The NRC staff has reviewed the licensee's 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 preventing startup 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 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 March 4, 1996, 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 Perry Public Library, 3753 Main Street,
Perry, Ohio. 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
[[Page 3953]]
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, 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 Gail H. Marcus: 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 Jay E. Silberg,
Shaw, Pittman, Potts & Trowbridge, 2300 N Street NW., Washington, DC
20037, 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 January 16, 1996, 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 Perry Public Library, 3753 Main
Street, Perry, Ohio.
Dated at Rockville, Maryland, this 29th day of January 1996.
For the Nuclear Regulatory Commission.
Jon B. Hopkins, Sr.,
Project Manager, Project Directorate III-3, Division of Reactor
Projects--III/IV, Office of Nuclear Reactor Regulation.
[FR Doc. 96-2206 Filed 2-1-96; 8:45 am]
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