[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]
BILLING CODE 7590-01-P