15942 Federal Register / Vol. 62, No. 64 / Thursday, April 3, 1997 / Notices

[Docket No. 50–286]

Power Authority of the State of New

York (Indian Point Nuclear Generating

Unit No. 3); Exemption

I

The Power Authority of the State of

New York (the licensee) is the holder of

Facility Operating License No. DPR–64,

which authorizes operation of the

Indian Point Nuclear Generating Unit

No. 3 (IP3). The license provides that

the licensee is subject to all rules,

regulations, and orders of the Nuclear

Regulatory Commission (the

Commission) now or hereafter in effect.

The facility consists of a pressurizedwater reactor at the licensee’s site

located in Westchester County, New

York.

II

The Code of Federal Regulations at

subsection (a) of 10 CFR 70.24,

‘‘Criticality Accident Requirements,’’

requires that each licensee authorized to

possess special nuclear material shall

maintain in each area where such

material is handled, used, or stored, a

criticality monitoring system ‘‘using

gamma- or neutron-sensitive radiation

detectors which will energize clearly

audible alarm signals if accidental

criticality occurs.’’ Subsection (a)(1) of

10 CFR 70.24 specifies the detection,

sensitivity, and coverage capabilities of

the monitors required by 10 CFR

70.24(a). The specific requirements of

subsection (a)(1) are that ‘‘the

monitoring system shall be capable of

detecting a criticality that produces an

absorbed dose in soft tissue of 20 rads

of combined neutron and gamma

radiation at an unshielded distance of 2

meters from the reacting material within

one minute.’’ Subsection (a)(3) of 10

CFR 70.24 requires that the licensee

shall maintain emergency procedures

for each area in which this licensed

special nuclear material is handled,

used, or stored and provides (1) that the

procedures ensure that all personnel

withdraw to an area of safety upon the

sounding of a criticality monitor alarm,

(2) that the procedures must include

drills to familiarize personnel with the

evacuation plan, and (3) that the

procedures designate responsible

individuals for determining the cause of

the alarm and placement of radiation

survey instruments in accessible

locations for use in such an emergency.

Subsection (d) of 10 CFR 70.24 states

that any licensee who believes that there

is good cause why he 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.

The purpose of 10 CFR 70.24 (a),

(a)(1), and (a)(3) is to ensure that any

inadvertent criticality is detected and

that action is taken to protect personnel

and correct the problem. By letter dated

December 20, 1996, as supplemented

March 5, 1997, and March 19, 1997, the

licensee requested an exemption from

the requirements of 10 CFR 70.24. The

licensee proposes to handle and store

unirradiated fuel without having the

criticality monitoring system specified

in 10 CFR 70.24. The licensee also

proposes to handle and store

unirradiated fuel without the speicfic

emergency procedures detailed in 10

CFR 70.24. The licensee believes that

fuel handling procedures and design

features make an inadvertent criticality

unlikely. The licensee believes that a

portable radiation monitoring system

and existing plant procedures will

provide adequate protection in the

unlikely event of an accidental

criticality. The licensee also believes

that current emergency procedures and

training are adequate to meet the intent

of 10 CFR 70.24(a)(3).

III

Special nuclear material, as nuclear

fuel, is stored in the spent fuel pool or

the new (unirradiated) fuel storage

racks. 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 racks are used to store new

fuel in a dry condition upon arrival on

site.

Special nuclear material is also

present in the form of fissile material

incorporated into fission chambers for

nuclear instrumentation, primary source

assemblies, and Health Physics

calibration sources. The small quantity

of special nuclear material present in

these items precludes an inadvertent

criticality.

Consistent with Technical

Specification Section 5.4, the spent fuel

pool is designed to store the fuel in a

geometric array using a solid neutron

absorber 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 normal and accident

conditions for fuel of maximum

enrichment of 5.0 wt% U–235. The staff

has found this design adequate.

The new fuel storage racks may be

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

enrichment of 5.0 wt% U–235 for the

new fuel assemblies results in a

maximum Keff of less than 0.95 under

conditions of accidental flooding. The

staff has found the design of the

licensee’s new fuel storage racks to be

adequate to store fuel enriched to no

greater than 5.0 wt% U–235.

Nuclear fuel is moved between the

new fuel storage racks, the reactor

vessel, and the spent fuel pool to

accommodate refueling operations. In

addition, fuel is moved into the facility

and within the reactor vessel, or within

the spent fuel pool. Fuel movements are

procedurally controlled and designed to

preclude conditions involving criticality

concerns. Fuel handling procedures and

the design features of the fuel handling

system are discussed in the licensee’s

Final Safety Analysis Report.

Technical Specification Section 3.8

precludes certain movements of heavy

loads over the spent fuel pool to prevent

a fuel handling accident. Previous

accident analyses have demonstrated

that a fuel handling accident (i.e., a

dropped fuel assembly) will not create

conditions which could result in

inadvertent criticality.

Procedures and controls prevent an

inadvertent criticality during fuel

handling; nevertheless the licensee will

provide monitoring in the IP3 Fuel

Storage Building during dry fuel

handling operations. During dry fuel

handling operations, the licensee will

have in operation at least one portable

detector that will meet the detection and

sensitivity criteria of Sections 5.6 and

5.7 of ANSI/ANS 8.3 (1986), ‘‘American

National Standard Criticality Accident

Alarm System.’’ Upon detection, this

instrument shall automatically cause an

immediate alarm audible in all areas

from which evacuation is necessary to

minimize exposure. The staff has

determined that the detection and

sensitivity criteria in the ANSI standard

are as rigorous as those specified in 10

CFR 70.24(a)(1). The staff has also

determined that, because fuel handling

equipment design and procedures make

a criticality unlikely, one detector will

be adequate and that in the case of fuel

handling at IP3 two detectors as

required by 10 CFR 70.24(a)(1) are not

necessary.

The licensee has procedures and

conducts training on dealing with

radiological emergencies consistent

with 10 CFR 50.47 and Part 50,

Appendix E. In addition to this training,

the licensee gives training on

responding to a criticality monitor alarm
Federal Register / Vol. 62, No. 64 / Thursday, April 3, 1997 / Notices 15943

to radiation workers accessing the fuel

handling building. This training will be

provided as necessary until dry fuel

handling in 1997 is complete and the

subject material has been incorporated

into general employee training. The staff

has determined that the licensee’s

procedures and training meet the intent

of 10 CFR 70.24(a)(3); therefore,

adherence to the specific requirements

of this section is not necessary to serve

the underlying purpose of the rule.

Because inadvertent criticality is

precluded by both design and

procedure, because adequate radiation

monitoring is present, and because the

licensee maintains emergency

procedures for the areas in which fuel

is handled, the staff has concluded that

there is reasonable assurance that

irradiated and unirradiated fuel will

remain subcritical; furthermore, there is

reasonable assurance that, should an

inadvertent criticality occur, the

licensee will detect such a criticality

and workers will respond properly. The

combination of plant design features,

fuel handling procedures, the use of a

portable criticality monitor, radiological

emergency procedures and radiation

worker training constitute good cause

for granting an exemption to the

requirements of 10 CFR 70.24.

IV

Accordingly, 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 the following exemption:

The Power Authority of the State of New

York is exempt from the requirements of 10

CFR 70.24(a), 10 CFR 70.24(a)(1), and 10 CFR

70.24(a)(3) for Indian Point Nuclear

Generating Unit No. 3. This exemption is

contingent on the facility’s maintaining the

hardware, procedure, and training described

in Section III above.

Pursuant to 10 CFR 51.32, the

Commission has determined that the

granting of this exemption will have no

significant impact on the quality of the

human environment (62 FR 14705).

This exemption is effective upon

issuance.

Dated at Rockville, MD, this 27th day of

March 1997.

For the Nuclear Regulatory Commission,

Frank J. Miraglia, Jr.,

Acting Director, Office of Nuclear Reactor

Regulation.

[FR Doc. 97–8545 Filed 4–2–97; 8:45 am]

BILLING CODE 7590–01–P

[Docket 70–7001]

Notice of Amendment to Certificate of

Compliance GDP–1 for the U.S.

Enrichment Corporation, Paducah

Gaseous Diffusion Plant, Paducah,

Kentucky

The Director, Office of Nuclear

Material Safety and Safeguards, has

made a determination that the following

amendment request is not significant in

accordance with 10 CFR 76.45. In

making that determination the staff

concluded that (1) there is no change in

the types or significant increase in the

amounts of any effluents that may be

released offsite; (2) there is no

significant increase in individual or

cumulative occupational radiation

exposure; (3) there is no significant

construction impact; (4) there is no

significant increase in the potential for,

or radiological or chemical

consequences from, previously analyzed

accidents; (5) the proposed changes do

not result in the possibility of a new or

different kind of accident; (6) there is no

significant reduction in any margin of

safety; and (7) the proposed changes

will not result in an overall decrease in

the effectiveness of the plant’s safety,

safeguards or security programs. The

basis for this determination for the

amendment request is shown below.

The NRC staff has reviewed the

certificate amendment application and

concluded that it provides reasonable

assurance of adequate safety, safeguards,

and security, and compliance with NRC

requirements. Therefore, the Director,

Office of Nuclear Material Safety and

Safeguards, is prepared to issue an

amendment to the Certificate of

Compliance for the Paducah Gaseous

Diffusion Plant. The staff has prepared

a Compliance Evaluation Report which

provides details of the staff’s evaluation.

The NRC staff has determined that

this amendment satisfies the criteria for

a categorical exclusion in accordance

with 10 CFR 51.22. Therefore, pursuant

to 10 CFR 51.22(b), no environmental

impact statement or environmental

assessment need be prepared for this

amendment.

USEC or any person whose interest

may be affected may file a petition, not

exceeding 30 pages, requesting review

of the Director’s Decision. The petition

must be filed with the Commission not

later than 15 days after publication of

this Federal Register Notice. A petition

for review of the Director’s Decision

shall set forth with particularity the

interest of the petitioner and how that

interest may be affected by the results of

the decision. The petition should

specifically explain the reasons why

review of the Decision should be

permitted with particular reference to

the following factors: (1) The interest of

the petitioner; (2) how that interest may

be affected by the Decision, including

the reasons why the petitioner should

be permitted a review of the Decision;

and (3) the petitioner’s areas of concern

about the activity that is the subject

matter of the Decision. Any person

described in this paragraph (USEC or

any person who filed a petition) may

file a response to any petition for

review, not to exceed 30 pages, within

10 days after filing of the petition. If no

petition is received within the

designated 15-day period, the Director

will issue the final amendment to the

Certificate of Compliance without

further delay. If a petition for review is

received, the decision on the

amendment application will become

final in 60 days, unless the Commission

grants the petition for review or

otherwise acts within 60 days after

publication of this Federal Register

Notice.

A petition for review 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.

For further details with respect to the

action see (1) the application for

amendment and (2) the Commission’s

Compliance Evaluation Report. These

items are 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.

Date of amendment request: February

14, 1997, revised March 10, 1997.

Brief description of amendment: The

amendment revises the Technical Safety

Requirement for the design features for

the cranes in the feed facilities and

reflects the associated changes to the

Safety Analysis Report.

Basis for finding of no significance:

1. The proposed amendment will not

result in a change in the types or

significant increase in the amounts of

any effluents that may be released

offsite.

The proposed change to TSR 2.2.5.2

involves a change to the design features

of the hoist brakes for the feed facility

cranes. These changes have no impact

on plant effluents and will not result in

any impact to the environment.

2. The proposed amendment will not

result in a significant increase in

individual or cumulative occupational

radiation exposure.