[Federal Register Volume 62, Number 189 (Tuesday, September 30, 1997)]
[Notices]
[Pages 51165-51167]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 97-25899]


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NUCLEAR REGULATORY COMMISSION

[Docket No. 50-302]


Florida Power Corporation; 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. 
DPR 72, issued to the Florida Power Corporation, (FPC or the licensee), 
for operation of the Crystal River Nuclear generating Unit 3 (CR3) 
located in Citrus County, Florida.
    The proposed amendment involves a revision to the Emergency Diesel 
Generator (EDG) protective relaying scheme at CR3, as described in the 
Final Safety Analysis Report (FSAR) Chapter 8. FPC has evaluated the 
proposed modifications pursuant to 10 CFR 50.59 and has determined that 
these modifications constitute an unreviewed safety question (USQ) 
based on a resulting increase in the probability of a malfunction of 
equipment important to safety. Therefore, FPC is requesting amendment 
of the CR3 license to resolve that USQ. The proposed modification will 
add new protective relays to each EDG generator output breaker to 
provide additional protection for a potential electrical fault or 
overpower condition.
    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:

    1. Does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The EDGs perform a support function for Design Basis Accident 
mitigation by providing a source of emergency AC electrical power 
for the Engineered Safeguards loads. For most Design Basis 
Accidents, a coincident Loss-of-Offsite-Power is postulated to occur 
and any single random electrical failure is considered credible 
including complete failure for one EDG to energize the associated 
4160V ES bus. The failure of an EDG to energize the associated 4160V 
ES bus is not a precursor for any postulated Design Basis Accident 
except Station Blackout (SBO). The failure of both EDGs concurrent 
with a Loss-of-Offsite-Power causes a Station Blackout. Therefore, 
any increase in the probability that an EDG will not energize the 
associated 4160V ES bus will increase the probability of a Station 
Blackout.
    The new relaying added to each EDG has a small probability of 
spuriously actuating, resulting in a small increase in the 
probability of an EDG failing to energize the associated 4160V ES 
bus. Spurious actuation of the overcurrent relaying for the load 
carrying 4160V ES bus offsite power source breaker will cause a loss 
of power on the 4160V ES bus and prevent the EDG from re-energizing 
the bus. In addition, a spurious actuation of the device-32X 
directional power auxiliary relay can cause a loss of offsite power 
for the associated 4160V ES bus. This spurious actuation also 
increases the probability of a Station Blackout. The only new system 
interfaces are between the EDG and 4160V ES bus systems. The 
modified relaying will not directly affect the fuel cladding, the 
Reactor Coolant System (RCS) pressure boundary, or the containment 
building.
    The increase in the probability of a Station Blackout is 
negligible. Although EDG availability is a contributor to the risk 
of Station Blackout, the CR-3 licensing basis assumes this event 
without regard to EDG reliability. Therefore, the probability of 
previously evaluated accidents is not significantly increased. The 
new protective relaying could shorten the duration of an actual 
Station Blackout if a 4160V ES bus fault or other similar problem 
was a contributor to the event by limiting the damage to the station 
power systems.
    The modified relaying will not increase the consequences of a 
Station Blackout since both EDGs and offsite power are assumed to be 
unavailable. The new protective relaying will not create any new 
timing or sequencing impact to the ES loads supplied from the 4160V 
ES bus. The small increase in probability that an EDG will not 
energize the associated 4160V ES bus does not invalidate the Design 
Basis Accident assumption that one EDG successfully energizes the 
associated 4160V ES bus (single failure proof). Therefore, the 
conclusions concerning fission product releases in the FSAR will not 
be changed.
    2. Does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    The modified relaying will not directly affect the fuel 
cladding, the Reactor Coolant System (RCS) pressure boundary, or the 
containment building. The modifications only impact the EDGs and 
4160V ES buses.
    The failure of one of the EDGs to energize the associated 4160V 
ES bus during a Design Basis Accident is a standard ``single 
failure'' for determining the acceptability of an accident 
mitigation system. A standby EDG and the associated 4160V ES bus are 
not capable of creating an accident such as a Loss-of-Coolant 
Accident (LOCA) or Main Steam Line Break (MSLB).
    There is a small increase in the probability that an EDG will 
not successfully energize the associated 4160V ES bus. However, the 
Design Basis Accident assumption that one EDG does successfully 
energize the bus remains valid. Therefore, no new accident involving 
the failure of both EDGs other than a Station Blackout needs to be 
postulated. The proposed modifications to the EDG relaying and the 
small increase in the probability that an EDG will not energize the 
associated 4160V ES bus do not introduce any new interfaces or 
mechanisms that could challenge any fluid system or fission product 
barrier in a different way than previously evaluated. Therefore, the 
modifications cannot create the possibility of an accident of a 
different type than previously evaluated in the FSAR.
    3. Does not involve a significant reduction in the margin of 
safety.
    The Bases of the CR-3 technical specifications do not identify a 
``margin of safety'' for the EDGs or 4160V ES buses that is 
applicable to the proposed EDG relaying modifications. Therefore, 
the plant response to Design Basis Accidents was evaluated. The 
accident analysis assumptions remain valid with the existing and 
proposed changes to the EDG and 4160V ES bus protective relaying. 
Plant response will remain as evaluated in the accident analysis and 
the calculated primary and secondary pressures and temperatures 
during evaluated accidents will not be increased by the changes. The 
reliability of each EDG and associated 4160V ES bus is being 
insignificantly reduced in order to increase the availability of the 
EDG and associated 4160V ES bus after a fault or overcurrent 
condition occurs. A spurious actuation of one of the added relays 
might cause one EDG to fail to energize one 4160V ES bus but would 
not result in failure of the other EDG to perform its function. 
Therefore, the changes do not reduce the margin of safety in the 
bases for any Improved Technical Specification.

    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

[[Page 51166]]

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. 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 
Review and Directives Branch, Division of Freedom of Information and 
Publications 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 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 October 30, 1997, 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 Coastal Region Library, 8619 W. Crystal 
Street, Crystal River, Florida 34428.
    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 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 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 R. Alexander Glenn, General Counsel, 
Florida Power Corporation, MAC--A5A, P. O. Box 14042, St. Petersburg, 
Florida 33733-4042, 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 September 12, 1997, which is 
available for public inspection at the Commission's Public Document 
Room, the Gelman

[[Page 51167]]

Building, 2120 L Street, NW., Washington, DC, and at the local 
public document room, located at the Coastal Region Library, 8619 W. 
Crystal Street, Crystal River, Florida 34428.

    Dated at Rockville, Maryland, this 23rd day of September 1997.

    For The Nuclear Regulatory Commission.
L. Raghavan, Sr.,
Project Manager, Project Directorate II-3, Division of Reactor 
Projects--I/II, Office of Nuclear Reactor Regulation.
[FR Doc. 97-25899 Filed 9-29-97; 8:45 am]
BILLING CODE 2590-01-P