Federal Register / Vol. 66, No. 115 / Thursday, June 14, 2001 / Notices 32397

1The scope of (a)(1) are those SSCs necessary to

assure (i) the integrity of the reactor coolant

pressure boundary, (ii) the capability to shut down

the reactor and maintain it in a safe condition, or

continued (iii) the capability to prevent or mitigate

the consequences of accidents which could result

in potential offsite exposures comparable to the

guideline exposures of part 100. The scope of (a)(2)

are those SSCs necessary for continued operation

without undue risk to the health and safety of the

public.

(HNP), Units 1 and 2, for an additional

20 years of operation. The HNP units are

operated by the Southern Nuclear

Operating Company, Inc. (SNC). HNP is

located in Appling County, Georgia.

Possible alternatives to the proposed

action (license renewal) include no

action and reasonable alternative

methods of power generation.

In Section 9.3 of the report:

The staff recommends that the Commission

determine that the adverse environmental

impacts of license renewal for HNP, Units 1

and 2 are not so great that preserving the

option of license renewal for energy planning

decisionmakers would be unreasonable. This

recommendation is based on (1) the analysis

and findings in the Generic Environmental

Impact Statement for License Renewal of

Nuclear Power Plants, NUREG–1437; (2) the

ER [Environmental Report] submitted by

SNC; (3) consultation with other Federal,

State, and local agencies; (4) the staff’s own

independent review; and (5) the staff’s

consideration of public comments.

The final Supplement 4 to the GEIS is

available electronically for public

inspection in the NRC Public Document

Room located at One White Flint North,

11555 Rockville Pike (first floor),

Rockville, Maryland, or from the

Publicly Available Records (PARS)

component of NRC’s document

management system (ADAMS). ADAMS

is accessible from the NRC Web site at

http://www.nrc.gov/NRC/ADAMS/

index.html (the Public Electronic

Reading Room).

FOR FURTHER INFORMATION CONTACT: Mr.

Andrew J. Kugler, Generic Issues,

Environmental, Financial, and

Rulemaking Branch, Division of

Regulatory Improvement Programs, U.S.

Nuclear Regulatory Commission,

Washington, DC 20555. Mr. Kugler may

be contacted at (301) 415–2828 or by

writing to: Andrew J. Kugler, U.S.

Nuclear Regulatory Commission, MS 0–

11 F1, Washington, DC 20555.

Dated at Rockville, Maryland, this 31 day

of May, 2001.

For the Nuclear Regulatory Commission.

David B. Matthews,

Director, Division of Regulatory Improvement

Programs, Office of Nuclear Reactor

Regulation.

[FR Doc. 01–14975 Filed 6–13–01; 8:45 am]

BILLING CODE 7590–01–P

NUCLEAR REGULATORY

COMMISSION

[Docket Nos. 50–498 and 50–499]

STP Nuclear Operating Company;

South Texas Project Electric

Generating Station, Units 1 and 2;

Environmental Assessment and

Finding of No Significant Impact

The U.S. Nuclear Regulatory

Commission (NRC) is considering

issuance of exemptions from certain

regulations found in 10 CFR parts 21,

50, and 100 for Facility Operating

License Nos. NPF–76 and NPF–80,

issued to STP Nuclear Operating

Company (STPNOC or the licensee) for

operation of the South Texas Project

Generating Station, Units 1 and 2, (STP)

located in Matagorda County, Texas.

Environmental Assessment

Identification of the Proposed Action

The proposed action would grant the

licensee relief from certain special

treatment requirements found in Title

10 of the Code of Federal Regulations,

Parts 21, 50, and 100 (10 CFR parts 21,

50 and 100) for certain structures,

systems, and components (SSCs). The

licensee has used a risk-informed

process to categorize SSCs as low safety

significant (LSS) or non-risk significant

(NRS); and other SSCs as medium safety

significant (MSS) or high safety

significant (HSS). The purpose of this

categorization process is to identify

those SSCs for which the special

treatment requirements may be relaxed.

Currently, LSS and NRS SSCs, which

are not as risk significant as MSS and

HSS SSCs, are treated with the same

level of protection. The licensee is

seeking limited exemptions from the

following regulations for just those SSCs

that have been categorized as LSS or

NRS:

1. Requirements for quality assurance

(QA) found in:

a. 10 CFR part 50, Appendix B,

‘‘Quality Assurance Criteria for Nuclear

Power Plants and Fuel Reprocessing

Plants,’’ for QA requirements on SSCs

that are safety-related (with the

exception of the Criterion III, ‘‘Design

Control,’’ Criterion XV,

‘‘Nonconforming Materials, Parts, or

Components,’’ and Criterion XVI,

‘‘Corrective Action’’);

b. 10 CFR part 50, Appendix A,

General Design Criteria (GDC) 1,

‘‘Quality Standards and Records,’’ for

SSCs important to safety that contains

quality assurance program and record

keeping requirements;

c. 10 CFR 50.34(b)(6)(ii) that requires

the licensee to describe in the Final

Safety Analysis Report how 10 CFR part

50, Appendix B, requirements are being

satisfied;

d. 10 CFR 50.54(a)(3) regarding NRC

review and approval of changes to the

QA program that result in a reduction in

commitments in the program

description as accepted by the NRC for

LSS and NRS SSC program descriptions;

and,

e. 10 CFR 21.3 defining the term

‘‘basic component’’ that includes safetyrelated LSS and NRS SSCs and impose

10 CFR part 21 requirements for

procurement, dedication, and reporting.

2. Requirements for environmental

qualification (EQ) found in:

a. 10 CFR 50.49(b) that defines the

scope of electric components important

to safety subject to the EQ program

requirements of 10 CFR 50.49;

b. 10 CFR part 50, Appendix A, GDC

2, ‘‘Design Bases for Protection Against

Natural Phenomena,’’ for tests and

inspections to demonstrate that SSCs

important to safety are designed to

withstand the effects of natural

phenomena without loss of capability to

perform their safety functions;

c. 10 CFR part 50, Appendix A, GDC

4, ‘‘Environmental and Dynamic Effects

Design Bases,’’ for tests and inspections

to demonstrate that SSCs important to

safety are able to withstand

environmental conditions of normal

operation, maintenance, testing, and

postulated accidents; and,

d. 10 CFR part 100, Appendix A,

Sections VI(a)(1) and (a)(2) for testing

and inspection to demonstrate that SSCs

within the scope of these regulations1

are designed to remain functional

during a safe-shutdown earthquake and

operating-basis earthquake, respectively,

and 10 CFR 50.34(b)(10) and 10 CFR

50.34(b)(11) to the extent that they

reference the 10 CFR part 100,

Appendix A, criteria, discussed above.

3. Requirements for testing and

inspection found in:

a. 10 CFR part 50, Appendix A, GDC

18, ‘‘Inspection and Testing of Electric

Power Systems,’’ that requires SSCs

important to safety be designed to

permit inspection and testing; and

b. 10 CFR part 50, Appendix J, Option

B, section III.B, ‘‘Type B and C Tests,’’

that requires Type C containment

32398 Federal Register / Vol. 66, No. 115 / Thursday, June 14, 2001 / Notices

2There are no NRS SSCs and limited LSS SSCs

modeled in the plant’s PRA due to a negligible

impact on risk or due to implicit modeling.

isolation valve leak rate tests for safetyrelated SSCs.

4. Requirements for monitoring the

effectiveness of maintenance under 10

CFR 50.65 for safety-related SSCs and

nonsafety-related SSCs that are relied

upon to mitigate accidents or transients

or are used in plant emergency

operating procedures, or whose failure

could prevent safety-related SSCs from

fulfilling their safety-related function, or

whose failure could cause a reactor

scram or actuation of a safety-related

system. The licensee is requesting an

exemption to exclude the LSS and NRS

SSCs from the scope of the maintenance

rule but would still conduct monitoring

at the plant, system, or train level.

Failure of an LSS or NRS SSC would not

count as a Maintenance Rule Functional

Failure unless the failure caused a

failure of a high or medium safety

significant function.

5. Industry code standards found in:

a. 10 CFR 50.55a(f) and (g) that

require repair and replacement,

inservice testing (IST), and inservice

inspection (ISI), under Section XI of the

American Society of Mechanical

Engineers Boiler and Pressure Vessel

Code; and,

b. 10 CFR 50.55a(h) that imposes the

quality and qualification requirements

of Sections 4.3 and 4.4 of Institute of

Electrical and Electronics Engineers

(IEEE) 279, ‘‘Criteria for Protection

Systems for Nuclear Power Plant

Generating Stations,’’ for electric SSCs

important to safety.

6. 10 CFR 50.59 to the extent that this

regulation requires a written evaluation

and prior NRC review and approval of

changes in special treatment

requirements for LSS and NRS SSCs.

The proposed action is in accordance

with the licensee’s application for

exemption dated July 13, 1999, as

supplemented on October 14 and 22,

1999; January 26 and August 31, 2000;

and January 15, 18, and 23, March 19,

and May 8 and 21, 2001 (hereinafter, the

submittal).

The Need for the Proposed Action

The exemptions are necessary to

provide the licensee relief from

regulatory requirements found in 10

CFR parts 21, 50, and 100 for LSS and

NRS components currently within the

scope of these regulations. In

accordance with 10 CFR 21.7 and 10

CFR 50.12, the Commission may grant

exemptions from the requirements of 10

CFR parts 21 and 50, respectively, under

certain circumstances. Further, the NRC

staff has determined that the requested

exemptions from 10 CFR part 100,

Appendix A, sections VI(a)(1) and (a)(2)

may be granted in accordance with the

requirements of 10 CFR 50.12. The NRC

staff has approved a Graded Quality

Assurance Program for STP. Exemptions

from certain special treatment

requirements are necessary to realize the

full benefit of the Graded Quality

Assurance Program.

The exemption is also necessary to

reduce occupational radiation exposures

and costs that would be expended in

providing qualifications, quality

assurance controls, maintenance,

monitoring requirements, testing, and

inspections for the LSS and NRS

components that may not be necessary

to maintain safety.

Environmental Impacts of the Proposed

Action

The NRC staff has completed its

evaluation of the proposed action and

concludes that many of the exemption

requests are not necessary for, or are not

consistent with the objective of,

maintaining design and functionality of

an SSC and will not be granted. The

NRC staff has determined that some of

the exemption requests that would

remove LSS and NRS SSCs from the

scope of the regulations, if granted,

would not present an undue risk to the

public health and safety. The

regulations for which exemptions are to

be granted are listed below and are

referred to as the proposed action in the

following sections.

a. 10 CFR 21.3—definition of basic

component;

b. 10 CFR 50.34(b)(10) and 10 CFR

50.34(b)(11), impose the requirements of

10 CFR part 100, Appendix A, section

VI(a)(1) and (2);

c. 10 CFR 50.49(b), scope of electrical

equipment subject to environmental

qualification requirements [design

aspects of 10 CFR 50.49(e)(1) through (7)

continue to apply];

d. 10 CFR 50.55a(f)—IST

requirements;

e. 10 CFR 50.55a(g)—repair and

replacement, and ISI requirements;

f. 10 CFR 50.55a(h), quality and

qualification requirements of sections

4.3 and 4.4 of IEEE 279;

g. 10 CFR 50.59—written evaluations

and prior NRC review and approval for

changes to special treatment

requirements;

h. 10 CFR 50.65(b)—scope of

maintenance rule [the requirements of

10 CFR 50.65(a)(4) continue to apply];

i. 10 CFR part 50, App. B—quality

assurance requirements (the

requirements of Criteria III, ‘‘Design

Control,’’ XV, ‘‘Nonconforming

Materials, Parts, or Components,’’ and

XVI, ‘‘Corrective Action,’’ continue to

apply);

j. 10 CFR part 50, Appendix J, Option

B, section III.B, Type C containment

isolation valve leak rate tests only;

k. 10 CFR part 100, Appendix A,

sections VI(a)(1) and (2), seismic

requirements for safe shutdown and

operating basis earthquakes.

The regulations, listed above, apply to

SSCs that are located entirely within the

restricted area and, if the exemptions

are granted, would not result in off-site

impacts due to normal operation. The

NRC staff evaluated the licensee’s

probabilistic risk analysis (PRA)

sensitivity study that addressed the

overall impact of reduced treatment for

LSS and NRS SSCs on plant risk for

those LSS SSCs that are modeled in the

STP PRA.2 Since the impact on failure

rates for these SSCs resulting from a

reduction in special treatment

requirements is not known, a factor of

10 increase in the failure rates of all LSS

SSCs modeled in the STP PRA was

used. The results of the sensitivity

analysis showed that the overall plant

risk for a core damage event increased

by 2.7 percent. The large early release

frequency increased by about 1.2

percent. The NRC staff finds the

sensitivity study to be an acceptable

method of ensuring that the cumulative

risk is only slightly impacted when

predicting significant changes in the

SSC failure rates, which may not occur.

Therefore, the NRC staff finds that the

postulated change in failure rates of the

LSS SSCs that are modeled in the PRA

would be expected to have a low overall

impact on plant risk.

On the other hand, the proposed

exemptions may have a beneficial

impact on occupational exposure, since

the additional requirements for QA, EQ,

monitoring, testing, and inspection for

certain LSS and NRS components

would not be necessary. The magnitude

of this benefit has not been quantified.

The proposed action will not

significantly increase the probability or

consequences of accidents, no changes

are being made in the types of any

effluents that may be released off site,

and there is no significant increase in

occupational or public radiation

exposure. Therefore, there are no

significant radiological environmental

impacts associated with the proposed

action.

With regard to potential

nonradiological impacts, the proposed

action does not involve any historic

sites. It does not affect nonradiological

plant effluents and has no other

environmental impact. Therefore, there

Federal Register / Vol. 66, No. 115 / Thursday, June 14, 2001 / Notices 32399

are no significant nonradiological

environmental impacts associated with

the proposed action.

Accordingly, the NRC staff concludes

that there are no significant

environmental impacts associated with

the proposed action.

Alternatives to the Proposed Action

As an alternative to the exemption

requests listed above, the NRC staff

considered denial of the proposed

action (i.e., the ‘‘no-action’’ alternative).

Denial of the application would result

in no significant change in current

environmental impacts.

Another alternative is to await

applicable regulations that are the result

of a future rulemaking under Option 2

of the Commission’s alternatives to risk

inform 10 CFR part 50 of the NRC’s

regulations discussed in SECY–98–300,

‘‘Options for Risk-Informed Revisions to

10 CFR part 50, ‘‘Domestic Licensing of

Production and Utilization Facilities’.’’

The exemptions requested by the

licensee are a proof-of-concept for this

broader rulemaking effort. The

Commission plans to use the STPNOC

exemption request and other industry

pilot programs to assist with the

development of the revised riskinformed 10 CFR part 50. The only

adverse environmental impact

associated with this proposed action

would be a slight increase in the risk of

an accident, but this impact would not

be significantly changed with the

alternative of awaiting a rulemaking.

Therefore, any relief granted under a

subset of a larger set of risk-informed

regulations under Option 2 in lieu of the

exemption requests would not provide a

significant benefit to public health or

safety, or the environment. The

environmental impacts associated with

granting the exemptions found to be

acceptable by the NRC staff and the

alternatives listed above are similar.

Alternative Use of Resources

This action does not involve the use

of any resources not previously

considered in the Final Environmental

Statement (NUREG–1171) for the South

Texas Project, Units 1 and 2, dated

August 1986.

Agencies and Persons Consulted

In accordance with its stated policy,

on June 1, 2001, the NRC staff consulted

with the Texas State official, Arthur C.

Tate, of the Division of Compliance and

Inspection, Bureau of Radiation Control,

Texas Department of Health, regarding

the environmental impact of the

proposed action. The State official had

no comments.

Finding of No Significant Impact

On the basis of the environmental

assessment, the NRC staff concludes

that the proposed action will not have

a significant effect on the quality of the

human environment. Accordingly, the

NRC staff has determined not to prepare

an environmental impact statement for

the proposed action.

For further details with respect to the

proposed action, see the licensee’s letter

dated July 13, 1999, as supplemented on

October 14 and 22, 1999; January 26 and

August 31, 2000; and January 15, 18,

and 23, March 19, and May 8 and 21,

2001. Documents may be examined and/

or copied for a fee, at the NRC’s Public

Document Room, located at One White

Flint North, 11555 Rockville Pike (first

floor), Rockville, Maryland. Publicly

available records will be accessible

electronically from the ADAMS Public

Library component of the NRC web site

http://www.nrc.gov (Electronic Reading

Room).

Dated at Rockville, Maryland, this 8th day

of June, 2001.

For the Nuclear Regulatory Commission.

Cynthia A. Carpenter,

Acting Director, Division of Licensing Project

Management, Office of Nuclear Reactor

Regulation.

[FR Doc. 01–14976 Filed 6–13–01; 8:45 am]

BILLING CODE 7590–01–P

NUCLEAR REGULATORY

COMMISSION

[Docket No. 50–271]

Vermont Yankee Nuclear Power

Corporation; Vermont Yankee Nuclear

Power Station; Environmental

Assessment and Finding of No

Significant Impact

The U.S. Nuclear Regulatory

Commission (NRC) is considering

amending a previously granted approval

to dispose of slightly contaminated soil

under 10 CFR 20.2002 by expanding the

allowable waste stream to include low

levels of radioactively contaminated soil

generated as a residual byproduct of

other types of on-site construction

activities. This approval is requested by

Vermont Yankee Nuclear Power

Corporation (the licensee), for operation

of the Vermont Yankee Nuclear Power

Station (Vermont Yankee), located in

Windham County, Vermont.

Environmental Assessment

Identification of the Proposed Action

The proposed action would amend

the previously granted approvals to

dispose of slightly contaminated septic

waste, cooling tower silt, soil/sand from

roadways and walkways, to include low

levels of radioactively contaminated

construction soil generated as a residual

byproduct of on-site construction

activities such as design change

implementation and land maintenance.

The proposed action is in accordance

with the licensee’s request dated

September 11, 2000.

The Need for the Proposed Action

The proposed action is needed to

dispose of slightly contaminated soil onsite. In accordance with 10 CFR

20.2002, which requires that a licensee

apply to the Commission for approval of

proposed procedures, not otherwise

authorized in the regulations, to dispose

of licensed material generated by the

licensee’s activities. The licensee

identified 28.3 cubic meters of approved

materials (i.e., soil/sand from roadways

and walkways, and soil from on-site

construction-related activities

including, but not limited to, design

change implementation and land

maintenance) to be disposed of on-site

on an annual basis until the expiration

of the plant’s operating license in 2013.

Since the previous approval did not

include disposal of soil generated as a

result of certain construction-related

activities, the licensee is requesting

approval to amend the previously

granted application pursuant to 10 CFR

20.2002, dated June 15, 2000.

Environmental Impacts of the Proposed

Action

The NRC has completed its evaluation

of the proposed action and concludes

that the proposed action will be bound

by the conditions for the on-site

disposals previously reviewed and

approved by the NRC. The licensee will

continue to use the designated and

approved areas on their property

(approximately 1.9 acres) and use

approximately 10 acres which have not

been previously used for disposal. The

amount of soil and soil/sand materials

that will be disposed has not increased,

and will remain at 28.3 cubic meters.

Determination of the radiological dose

impact of the new material has been

made based on the same dose

assessment models and pathway

assumptions used in the previous

submittals. The licensee’s proposal was

evaluated against the NRC staff’s

guidelines for on-site disposal and

found not to be a significant radiological

environmental impact. The bounding

dose conditions for the previously

approved materials will not be

exceeded. The potential exposure to

members of the general public from the

radionuclides in material was