Federal Register / Vol. 62, No. 189 / Tuesday, September 30, 1997 / Notices 51165

Dated: September 26, 1997.

Bea Hardesty,

Federal Register Liaison Officer.

[FR Doc. 97–26056 Filed 9–26–97; 2:57 pm]

BILLING CODE 7533–01–P

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-OffsitePower 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 reenergizing 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
51166 Federal Register / Vol. 62, No. 189 / Tuesday, September 30, 1997 / Notices

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
Federal Register / Vol. 62, No. 189 / Tuesday, September 30, 1997 / Notices 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

NUCLEAR REGULATORY

COMMISSION

[Docket Nos. 50–338]

In the Matter of Virginia Electric and

Power Company North Anna Power

Station, Unit Nos. 1 and 2; Exemption

and 50–339

I

The Virginia Electric and Power

Company (the licensee) is the holder of

Facility Operating License Nos. NPF–4

and NPF–7, which authorize operation

of the North Anna Power Station

(NAPS), Unit Nos. 1 and 2. The licenses

provide, among other things, that the

licensee be subject to all rules,

regulations, and Orders of the Nuclear

Regulatory Commission (the

Commission) now or hereafter in effect.

The facility consists of two

pressurized-water reactors at the

licensee’s site located in Louisa County,

Virginia.

II

The Code of Federal Regulations at 10

CFR 70.24, ‘‘Criticality Accident

Requirements,’’ requires that each

licensee authorized to possess special

nuclear material (SNM) shall maintain a

criticality accident monitoring system in

each area where such material is

handled, used, or stored. Subsections

(a)(1) and (a)(2) of 10 CFR 70.24 specify

detection, sensitivity and coverage

capabilities of the monitors required by

10 CFR 70.24(a). Subsection (a)(3)

requires licensees to maintain

emergency procedures for each area in

which this licensed SNM is handled,

used, or stored.

Subsection (d) of 10 CFR 70.24 states

that any licensee who believes that there

is good cause why it should be granted

an exemption from all or part of 10 CFR

70.24 may apply to the Commission for

such an exemption and shall specify the

reasons for the relief requested.

III

By letter dated January 28, 1997, as

supplemented March 24, 1997, Virginia

Electric and Power Company requested

an exemption from 10 CFR 70.24(a). The

Commission technical staff has

reviewed the licensee’s submittal and

has determined that inadvertent

criticality is not likely to occur in SNM

handling or storage areas at NAPS, Units

1 and 2.

At North Anna, SNM is present

principally as nuclear fuel. Other small

quantities of SNM are used on site.

However, the total amount used in nonfuel applications is significantly less

than the quantity specified in 10 CFR

70.24(a). The small quantity of non-fuel

SNM present, and the form in which it

is stored and used, precludes an

inadvertent criticality. Therefore, SNM

used as nuclear fuel is the only material

on site subject to the requirements of 10

CFR 70.24(a).

Nuclear fuel is stored in the new fuel

storage area and the spent fuel pool.

New fuel is stored dry (in air) in the new

fuel storage area. The spent fuel pool is

used to store irradiated fuel under water

after its discharge from the reactor, and

new fuel prior to loading into the

reactor.

The new fuel storage area is used to

receive and store new fuel in a dry

condition upon arrival on site and prior

to loading in the reactor or spent fuel

pool. The spacing between new fuel

assemblies in the storage racks is

sufficient to maintain the array in a

subcritical condition even under

accident conditions assuming the

presence of moderator. The maximum

nominal enrichment of 4.3 wt% U–235

for the new fuel assemblies results in a

maximum keff of less than 0.95 under

conditions of accidental flooding by

unborated water and keff less than 0.98

under conditions of low-density

optimum moderation. The staff has

found the design of the licensee’s new

fuel storage racks to be adequate to store

fuel enriched to 4.3 wt% U–235.

Consistent with Technical

Specification Section 5.6.1.1, the spent

fuel pool is designed to store the fuel in

a geometric array that precludes

criticality. The spent fuel racks are

designed such that the effective neutron

multiplication factor, keff, will remain

less than or equal to 0.95 under all

normal and accident conditions for fuel

of maximum nominal enrichment of 4.3

wt% U–235.

Nuclear fuel is moved between the

shipping container, the new fuel storage

racks, the reactor vessel, and the spent

fuel pool to accommodate refueling

operations. In all cases, fuel movements

are procedurally controlled and

designed to preclude conditions

involving criticality concerns.

The purpose of the criticality

monitors required by 10 CFR 70.24 is to

ensure that if a criticality were to occur

during the handling of nuclear material,

personnel would be alerted to that fact

and would take appropriate action.

Although the staff has determined that

such an accident is not likely to occur,

the licensee has radiation monitors, as

required by General Design Criterion 63,

in fuel storage and handling areas.

These monitors have associated area

alarms and control room annunciators

and would detect excessive radiation

levels and will alert personnel to allow

them to initiate appropriate emergency

procedures and safety actions. The low

probability of an inadvertent criticality

together with the licensee’s adherence

to General Design Criterion 63

constitute good cause for granting an

exemption to the requirements of 10

CFR 70.24(a).

IV

The Commission has determined that,

pursuant to 10 CFR 70.14, this

exemption is authorized by law, will not

endanger life or property or the common

defense and security, and is otherwise

in the public interest; therefore, the

Commission hereby grants Virginia

Electric and Power Company the

exemption from the requirements of 10

CFR 70.24(a) for North Anna Power

Station, Unit Nos. 1 and 2, relating to

criticality accident monitoring

requirements.

V

Pursuant to 10 CFR 51.32, the

Commission has determined that the

granting of this exemption will not

result in any significant adverse

environmental impact (62 FR 49540).

This exemption is effective upon

issuance.

Dated at Rockville, Maryland, this 23rd day

of September 1997.

For The Nuclear Regulatory Commission.

Frank J. Miraglia,

Acting Director, Office of Nuclear Reactor

Regulation.

[FR Doc. 97–25900 Filed 9–29–97; 8:45 am]

BILLING CODE 7590–01–P