29140 Federal Register / Vol. 61, No. 111 / Friday, June 7, 1996 / Notices

NUCLEAR REGULATORY

COMMISSION

[Docket No. 50–335]

Florida Power and Light 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. DPR–

67, issued to Florida Power and Light

Company, (the licensee), for operation

of the St. Lucie Plant Unit No. 1 located

in St. Lucie County, Florida.

The proposed amendment would

reduce the stated value of design reactor

coolant flow from 355,000 gpm to

345,000 gpm, revise the reactor core

thermal margin safety limits shown in

FIGURE 2.1–1, and modify the reactor

coolant system total water and steam

volume described in the design features.

The amendment also reduces the

Limiting Safety System Setting for the

reactor coolant low flow trip function

from greater than or equal to 95% to

greater than or equal to 93% of design

reactor coolant flow. Finally, TS 2.1.1 is

modified to limit reactor power to less

than or equal to 90% rated thermal

power for Cycle 14 operation exceeding

mid-cycle fuel burn up conditions. The

revisions are being made to support

changes in the safety analyses which

accommodate a larger number of

plugged steam generator tubes.

On April 29, 1996, St. Lucie Unit 1

entered a scheduled refueling outage. A

margin of approximately 14% existed

between the average number of steam

generator (SG) tubes that had been

previously removed from service and

the number of plugged tubes assumed in

the safety analyses. Based on a 10-year

history of 100% Eddy Current Testing

(ECT), and including additional

inspection commitments pursuant to

generic letter (GL) 95–03,

‘‘Circumferential Cracking of Steam

Generator Tubes,’’ the number of tubes

conservatively estimated to be removed

from service during this outage was far

less than the remaining analytical

margin.

Based on meetings and conversations

with NRC staff subsequent to entry into

the outage, concerns involving the

qualification of techniques for sizing SG

tube crack-like indications were

identified, resulting in the staff

questioning the SG tube repair criteria

which have been in place at Florida

Power and Light Company (FPL) since

1985. On May 14, 1996, FPL agreed to

implement a more conservative criteria

for the Cycle 14 inspection. The

licensee’s assessment of the impact of

implementing this criteria indicates that

the number of SG tubes to be plugged

may exceed the existing 25% (average)

analyses limit.

The change in repair criteria and the

magnitude of resultant SG tube plugging

could not have been reasonably

anticipated prior to NRC staff concerns

having been communicated to FPL

during the recent meeting and

discussions. The need for an

amendment to implement revised St.

Lucie Unit 1 power and RCS flow limits

could not have been anticipated prior to

assessing the impact of the change in

repair criteria following FPL’s meeting

and discussions with the NRC staff. The

necessary evaluations and preparation

of the proposed license amendment

were initiated without delay and at the

earliest practical time. Analyses and

quality assurance verifications to

support the proposed license

amendment were completed in an

expeditious manner, and were

performed in parallel with the ongoing

tube examinations.

FPL expects to complete the refueling

overhaul and the required startup

preparations by June 20, 1996. Until a

license amendment is issued to

authorize operation with the proposed

changes, resumption of St. Lucie Unit 1

power operations will be prevented by

the current Technical Specifications.

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.

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

analysis of the issue of no significant

hazards consideration, which is

presented below:

(1) Operation of the facility in accordance

with the proposed amendment would not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed amendment defines reactor

core thermal margin safety limits for a

reduced value of design reactor coolant flow,

and establishes a revised Limiting Safety

System Setting (LSSS) for the protective

system low flow trip. As core protection

variables, these limiting parameters are not

accident initiators and do not affect the

frequency of occurrence of previously

analyzed transients. The design features’ total

water and steam volume revision accounts

for steam generator tube plugging and is

simply administrative in nature. Evaluations

performed to assess the impact of the

proposed amendment conclude that, when

considering a unit derate to 90% rated

thermal power for operation beyond 7000

EFPH in Cycle 14 as required by the

proposed change to TS 2.1.1, the potential

radiological consequences of previously

analyzed transients will conservatively

remain within established acceptance

criteria. Therefore, operation of the facility in

accordance with this amendment would not

involve a significant increase in the

probability or the consequences of any

accident previously evaluated.

(2) Operation of the facility in accordance

with the proposed amendment would not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

The proposed amendment revises limiting

parameters to assure safe operation

commensurate with the impact of steam

generator tube plugging, and will not change

the modes of operation defined in the facility

license. The analysis of transients associated

with steam generator failures are part of the

design and licensing bases. Therefore,

operation of the facility in accordance with

the proposed amendment would not create

the possibility of a new or different kind of

accident from any accident previously

evaluated.

(3) Operation of the facility in accordance

with the proposed amendment would not

involve a significant reduction in a margin of

safety.

The proposed amendment allows full

power operation at an RCS flow

commensurate with 30% (average) steam

generator tube plugging for Cycle 14 fuel

batch average burn up conditions

corresponding to mid-cycle. For operation

beyond mid-cycle, reactor power will be

restricted to less than or equal to 90% rated

thermal power. An evaluation of limiting

events to established acceptance criteria for

Specified Acceptable Fuel Design Limits

(SAFDL), primary and secondary over

pressurization transients, 10 CFR 50.46(b)

emergency core cooling systems acceptance

criteria, peak containment pressure, potential

radiation dose during accidents, and to TS

Limiting Conditions for Operation has been

completed in support of this amendment

request. The evaluation concludes, when

considering the proposed LSSS for the Low

Flow trip, that a conservative margin to

acceptable limits remains available.

Therefore, operation of the facility in

accordance with this proposed amendment

would not involve a significant reduction in

a margin of safety.
Federal Register / Vol. 61, No. 111 / Friday, June 7, 1996 / Notices 29141

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 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 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 June 24, 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 Indian

River Junior College Library, 3209

Virginia Avenue, Fort Pierce, Florida,

34954–9003. 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 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 20555–0001, 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

Frederick J. Hebdon: petitioner’s name
29142 Federal Register / Vol. 61, No. 111 / Friday, June 7, 1996 / Notices

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 20555–0001, and to

Harold F. Reis, Esquire, Newman; and

Holtzinger, 1615 L Street, NW.,

Washington, DC 20036, 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 June 1, 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

Indian River Junior College Library,

3209 Virginia Avenue, Fort Pierce,

Florida 34954–9003.

Dated at Rockville, Maryland, this 3rd day

of June 1996.

For the Nuclear Regulatory Commission.

Leonard A. Wiens,

Senior Project Manager, Project Directorate

II–3, Division of Reactor Projects—I/II, Office

of Nuclear Reactor Regulation.

[FR Doc. 96–14391 Filed 6–6–96; 8:45 am]

BILLING CODE 7590–01–P

NUCLEAR REGULATORY

COMMISSION

[Docket No. 50–244]

Exemption

In the Matter of Rochester Gas and Electric

Corporation, R.E. Ginna Nuclear Power Plant)

I

On December 10, 1984, the Nuclear

Regulatory Commission issued Facility

Operating License No. DPR–18 to

Rochester Gas and Electric Corporation

(RG&E) for the R.E. Ginna Nuclear

Power Plant (Ginna). The license

stipulated, among other things, that the

facility is subject to all rules,

regulations, and orders of the

Commission.

II

The Code of Federal Regulations,

Paragraph I.D.3, ‘‘Calculation of Reflood

Rate for Pressurized Water Reactors

[PWRs],’’ of Appendix K to Part 50 of

Title 10 of the Code of Federal

Regulations (10 CFR) requires that the

refilling of the reactor vessel and the

time and rate of reflooding of the core

be calculated by an acceptable model

that considers the thermal and

hydraulic characteristics of the core and

of the reactor system. In particular,

Paragraph I.D.3 requires, in part, that,

‘‘The ratio of the total fluid flow at the

core exit plane to the total flow at the

core inlet plane (carryover fraction)

shall be used to determine the core exit

flow and shall be determined in

accordance with applicable

experimental data.’’ The purpose of this

requirement is to assure that the core

exit flow during the post-loss-of-coolant

accident (LOCA) refill/reflood phase is

determined using a model that accounts

for appropriate experimental data.

Paragraph I.D.5, ‘‘Refill and Reflood

Heat Transfer for Pressurized Reactors,’’

of Appendix K to 10 CFR Part 50

requires that for (1) reflood rates of 1

inch per second or higher, the reflood

heat transfer coefficients be based on

applicable experimental data for

unblocked cores, and (2) reflood rates

less than 1 inch per second during refill

and reflood, heat transfer calculations

be based on the assumption that cooling

is only by steam.

License Condition 2.D provided an

exemption from 10 CFR 50.46(a)(1) that

the emergency core cooling system

(ECCS) performance be calculated in

accordance with an acceptable

calculational model which conforms to

the provisions of Appendix K (SER

dated April 18, 1978). The exemption

will expire upon receipt and approval of

revised ECCS calculations.

By letter dated November 5, 1992, as

supplemented on June 19, 1995, RG&E

(the licensee) requested an exemption

from 10 CFR Part 50, Appendix K,

Paragraphs I.D.3 and I.D.5 based on

revised ECCS calculations.

The November 5, 1992, exemption

request was supported first by a plant

specific ECCS evaluation model (EM)

using a methodology not yet approved

by NRC (WCAP–10924–P, Volume 2,

Revision 2, Addendum 3). The proposed

EM would have supported the May

1993, 1994, and 1995 core reloads.

However, the WCAP–10924–P, Revision

2, Volume 2, Addendum 3 methodology

has not yet been approved by NRC. On

June 19, 1995, the licensee supported

the November 5, 1992, exemption

request by an updated plant specific EM

using a methodology approved by NRC

(WCAP–10924–P, Volume 1, Revision 1,

Addendum 4). The proposed June 19,

1995, EM includes larger peaking factors

necessary to support conversion to an

18-month fuel-cycle reload to begin in

May 1996.

The specific provision of Paragraph

I.D.3 from which the licensee requested

an exemption, is the calculation of core

exit flow based on carryover fraction.

The licensee stated that the

prescriptions for this calculation given

in Paragraph I.D.3 were based on data

for a bottom-flooding configuration

design. The Ginna design relies on

upper plenum injection (UPI) for the

ECCS injection during the reflood phase

of a large-break LOCA. UPI is not a

‘‘lower flooding design;’’ its ECCS flow

patterns, flow magnitudes, core cooling

mechanisms, and, in fact, the meanings

and impacts of the terms ‘‘inlet’’ and

‘‘exit’’ are different than those of bottom

flooding plants. This EM described in

WCAP 10924–P, Volume 1, Revision 2,

Addendum 4, ‘‘Westinghouse UPI

Model Improvements,’’ dated August

1990, which has been generically

approved in a staff SER of February 8,

1991, determines core flow, including

flow ‘‘exiting’’ the core, flow ‘‘entering’’

the core, and flow within the core and

elsewhere within the reactor coolant

system (RCS) in accordance with

applicable experimental data. The data

are different than that referenced in

paragraph I.D.3, however, they were

found acceptable because they are

specifically applicable to UPI designs.

Because of the differences between UPI

design considerations and those for

bottom flooding designs mentioned

above, the ‘‘carryover fraction’’ as

defined in paragraph I.D.3 is not

calculated in the approved EM and

would not have the same technical

significance if it were. The licensee,

therefore, concludes that, in using the

approved UPI model with its technical

improvements for Ginna, it will not

comply with Paragraph I.D.3. The staff

SER of February 8, 1991, finds WCAP–

10924–P EM contains an empirically

verified model more directly applicable

to top flooding situations to calculate

core exit flow, which satisfies the

technical purpose of this Appendix K,

paragraph I.D.3 requirement to

determine the core exit flow, but does

not comply with the letter of the

requirement.

In more detail, the intent of the

Appendix K, paragraph I.D.3, is to

assure that the calculation of core exit

flow is performed using an EM code

model which has been verified against

appropriate experimental data for LOCA

accident analyses. The Westinghouse

COBRA/TRAC code (WCOBRA/TRAC)

consists of (1) Westinghouse LargeBreak LOCA Best Estimate

Methodology, Volume 1: Model