17886 Federal Register / Vol. 62, No. 70 / Friday, April 11, 1997 / Notices

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.

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 John F.

Stolz, Director, Project Directorate I–2:

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–

0001, and to Jay Silberg, Esquire, Shaw,

Pittman, Potts and 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).

If a request for a hearing is received,

the Commission’s staff may issue the

amendment after it completes its

technical review and prior to the

completion of any required hearing if it

publishes a further notice for public

comment of its proposed finding of no

significant hazards consideration in

accordance with 10 CFR 50.91 and

50.92.

For further details with respect to this

action, see the applications for

amendment dated November 27, 1996,

as supplemented February 12, 1997,

which 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

located at the Osterhout Free Library,

Reference Department, 71 South

Franklin Street, Wilkes-Barre, PA 18701.

Dated at Rockville, Maryland, this 4th day

of April 1997.

For the Nuclear Regulatory Commission.

David H. Jaffe,

Senior Project Manager, Project Directorate

I–2, Division of Reactor Projects—I/II, Office

of Nuclear Reactor Regulation.

[FR Doc. 97–9395 Filed 4–10–97; 8:45 am]

BILLING CODE 7590–01–P

NUCLEAR REGULATORY

COMMISSION

[Docket Nos. 50–387 and 50–388]

Pennsylvania Power and Light

Company Susquehanna Steam Electric

Station, Units 1 and 2 Environmental

Assessment and Finding of No

Significant Impact

The U.S. Nuclear Regulatory

Commission (the Commission) is

considering issuance of an amendment

to Facility Operating License Nos. DPR–

14 and DPR–22, issued to Pennsylvania

Power and Light Company (PP&L) (the

licensee), for the operation of

Susquehanna Steam Electric Station

(SSES), Units 1 and 2, located at the

licensee’s site in Luzerne County,

Pennsylvania.

Environmental Assessment

Identification of the Proposed Action

The proposed amendment will add to

the current SSES Technical

Specifications (TSs) (Special Test

Exception Section 3.10.7 and 3.10.8),

the Improved Technical Specifications

Sections (ITS) 3.10.3 and 3.10.4 in a

modified format and with applicable

cross references.

The proposed action is in accordance

with the licensee’s amendment request

dated February 11, 1997.

The Need for the Proposed Action

It has been recognized that nuclear

safety in all plants would benefit from

improvement and standardization of

TSs. The ‘‘NRC Interim Policy

Statement on Technical Specification

Improvements for Nuclear Power

Reactors,’’ (52 FR 3788, February 6,

1987) and later the Final Policy

Statement (58 FR 39132, July 22, 1993),

formalized this need. To facilitate the

development of individual ITS, each

reactor vendor owners group (OG) and

the NRC staff developed standard TS

(STS). For General Electric (GE) plants,

the STS are NUREG–1433 for BWR/4

reactor facilities and NUREG–1434 for

BWR/6 facilities. NUREG–1433 formed

the basis of the SSES ITS. The NRC

Committee to Review Generic

Requirements (CRGR) reviewed the STS

and made note of the safety merits of the

STS and indicated its support of

conversion to the STS by operating

plants.

Description of the Proposed Change

The February 11, 1997 submittal

requested that two sections be approved

prior to the staff approval of the entire

ITS to adopt Sections 3.10.3 and 3.10.4

of the ITS into the current TS Special

Test Exception Sections 3.10.3 and

3.10.4. This change will permit control

rod testing during refueling outages. The

only creditable accident associated with

control rod testing during the refuel

outage is the ‘‘Rod Withdrawal Error—

Low Power’’ and is addressed in Section

15.4.1 of SSES Updated Final Safety

Analysis Report (UFSAR).

The February 11, 1997 request is part

of a larger amendment request

submitted on August 1, 1996. The

requests are based on NUREG–1433 and

on guidance provided in the abovereferenced Policy Statement. If granted,

the amendments would completely

rewrite, reformat, and streamline the

existing TSs. Emphasis is placed on

human factors principles to improve

clarity and understanding. The Bases

section would be significantly expanded

to clarify and better explain the purpose
Federal Register / Vol. 62, No. 70 / Friday, April 11, 1997 / Notices 17887

and foundation of each specification. In

addition to NUREG–1433, portions of

the existing TSs were also used as the

basis for the ITS. Plant-specific issues

(unique design features, requirements,

and operating practices) were discussed

at length with the licensee, and generic

matters with the OGs.

Environmental Impacts of the Proposed

Action

As stated above, the only plausible

consequence of the proposed action is a

rod withdrawal error during low power.

The effects of such an error were

analyzed in ‘‘Rod Withdrawal Error-Low

Power,’’ Section 15.4.1 of the UFSAR.

This analysis indicates that withdrawal

of a single rod during refueling is

insufficient to cause criticality and thus

no radioactive materials would be

released. The proposed change to the

TSs does not change this conclusion.

Additionally, the proposed revision to

the TS was found to provide control of

plant operations, specifically control of

rod movement during Conditions 3 and

4. Thus, reasonable assurance will be

provided that the health and safety of

the public will be adequately protected.

These TS changes will not increase

the probability or consequences of

accidents, no changes are being made in

the types of any effluent that may be

released offsite, and there is no

significant increase in the allowable

individual or cumulative occupational

radiation exposure. Therefore, the

Commission concludes that there are no

significant radiological environmental

impacts associated with the proposed

TS amendment.

With regard to potential nonradiological impacts, the proposed

amendment involves features located

entirely within the restricted areas as

defined in 10 CFR Part 20. It does not

affect non-radiological plant effluents

and has no other environmental impact.

Therefore, the Commission concludes

that there are no significant nonradiological impacts associated with the

proposed amendment.

Alternatives to the Proposed Action

The Commission has concluded there

are no significant environmental

impacts associated with the proposed

amendment. Any alternatives with

equal or greater environmental impact

need not be evaluated. As an alternative

to the proposed action, the staff

considered denial of the proposed

action. Denial of the application would

result in no change in current

environmental impacts. The

environmental impacts of the proposed

action and the no-action alternative

action are similar.

Alternative Use of Resources

This action does not involve the use

of any resources not considered

previously in the Final Environmental

Statement for the Susquehanna Steam

Electric Station, Units 1 and 2, dated

June 1981.

Agencies and Persons Consulted

In accordance with its stated policy,

on March 27, 1997, the staff consulted

with the Pennsylvania State official, Mr.

David Ney of the Pennsylvania

Department of Environmental

Resources, Bureau of Radiation

Protection, regarding the environmental

impact of the proposed action. The State

official had no comments.

Finding of No Significant Impact

Based upon the environmental

assessment, the Commission concludes

that the proposed action will not have

a significant effect on the quality of the

human environment. Accordingly, the

Commission has determined not to

prepare an environmental impact

statement for the proposed action.

For further details with respect to this

proposed action, see the licensee’s letter

dated February 11, 1997. The letter is

available for public inspection at the

Commission’s Public Document Room,

the Gelman Building, 2120 L Street,

NW, Washington, DC 20555, and at the

local public document room located at

the Osterhout Free Library, Reference

Department, 71 South Franklin Street,

Wilkes-Barre, PA 18701.

Dated at Rockville, Maryland, this 4th day

April of 1997.

For the Nuclear Regulatory Commission.

John F. Stolz,

Director, Project Directorate I–2, Division of

Reactor Projects—I/II, Office of Nuclear

Reactor Regulation.

[FR Doc. 97–9393 Filed 4–10–97; 8:45 am]

BILLING CODE 7590–01–P

NUCLEAR REGULATORY

COMMISSION

Degradation of Control Rod Drive

Mechanism Nozzle and Other Vessel

Closure Head Penetrations; Issued

AGENCY: Nuclear Regulatory

Commission.

ACTION: Notice of issuance.

SUMMARY: The Nuclear Regulatory

Commission (NRC) has issued Generic

Letter 97–01 to notify all holders of

operating licenses for pressurized water

reactors (PWRs), except those who have

permanently ceased operations and

have certified that fuel has been

permanently removed from the reactor

vessel, of the need for information

concerning their programs for ensuring

the timely inspection of control rod

drive mechanism (CRDM) and other

vessel closure head penetrations. The

information requested is needed by the

NRC staff to verify compliance with 10

CFR 50.55a and 10 CFR Part 50,

Appendix A, GDC 14, and to determine

whether an augmented inspection

program, pursuant to 10 CFR

50.55a(g)(6)(ii), is required.

The proposed generic letter is a ‘‘rule’’

for purposes of the Small Business

Regulatory Enforcement Fairness Act (5

U.S.C., Chapter 8). The staff has

received confirmation from the Office of

Management and Budget that the

generic letter is a non-major rule.

This generic letter is available in the

NRC Public Document Room under

accession number 9703260336.

DATES: The generic letter was issued on

April 1, 1997.

ADDRESSES: Not applicable.

FOR FURTHER INFORMATION CONTACT: C. E.

Carpenter, Jr. at (301) 415–2169.

SUPPLEMENTARY INFORMATION: The NRC

staff has concluded that vessel closure

head penetration (VHP) cracking does

not pose an immediate or near term

safety concern. In the long term,

however, the degradation of CRDM

nozzles and other VHPs is an important

safety consideration that warrants

further evaluation. The vessel closure

head provides the vital function of

maintaining reactor pressure boundary.

Cracking in the VHPs has occurred and

is expected to continue to occur as

plants age. The NRC staff considers

cracking of VHPs to be a safety concern

for the long term based on the

possibility of (1) Exceeding the

American Society of Mechanical

Engineers (ASME) Code for margins if

the cracks are sufficiently deep and

continue to propagate during

subsequent operating cycles, and (2)

eliminating a layer of defense in depth

for plant safety. Therefore, to verify that

the margins required by the ASME

Code, as specified in 10 CFR 50.55a are

met, that the guidance of General Design

Criterion 14 of Appendix A to 10 CFR

Part 50 is continued to be satisfied, and

to ensure that the safety significance of

VHP cracking remains low, the NRC

staff believes that an integrated, longterm program, which includes periodic

inspections and monitoring of VHPs, is

necessary. In addition, the NRC staff

finds that the requested information is

also needed to determine if the

imposition of an augmented inspection

program, pursuant to 10 CFR

50.55a(g)(6)(ii), is required to maintain

public health and safety. The staff is not