25242 Federal Register / Vol. 63, No. 88 /Thursday, May 7, 1998 / Notices

affirmation, specifically admit or deny

each allegation or charge made in this

Order and shall set forth the matters of

fact and law on which Mr. Nardslico or

other person adversely affected relies

and the reasons as to why the Order

should not have been issued. Any

answer or request for a hearing shall be

submitted to the Secretary, U.S. Nuclear

Regulatory Commission, Attn: Chief,

Rulemakings and Adjudications Staff,

Washington, DC 20555. Copies also

shall be sent to the Director, Office of

Enforcement, U.S. Nuclear Regulatory

Commission, Washington, DC 20555, to

the Assistant General Counsel for

Hearings and Enforcement at the same

address, to the Regional Administrator,

NRC Region I, U.S. Nuclear Regulatory,

475 Allendale Road, King of Prussia,

Pennsylvania 19406, and to Mr.

Nardslico if the answer or hearing

request is by a person other than Mr.

Nardslico. If a person other than Mr.

Nardslico requests a hearing, that

person shall set forth with particularity

the manner in which that person’s

interest is adversely affected by this

Order and shall address the criteria set

forth in 10 CFR 2.714(d).

If a hearing is requested by Mr.

Nardslico or a person whose interest is

adversely affected, the Commission will

issue an Order designating the time and

place of any hearing. If a hearing is held,

the issue to be considered at such

hearing shall be whether this Order

should be sustained.

Pursuant to 10 CFR 2.202(c)(2)(i), Mr.

Nardslico may, in addition to

demanding a hearing, at the time the

answer is filed or sooner, move the

presiding officer to set aside the

immediate effectiveness of the Order on

the ground that the Order, including the

need for immediate effectiveness, is not

based on adequate evidence but on mere

suspicion, unfounded allegations, or

error.

In the absence of any request for

hearing, or written approval of an

extension of time in which to request a

hearing, the provisions specified in

Section IV above shall be final 20 days

from the date of this Order without

further order or proceedings. If an

extension of time for requesting a

hearing has been approved, the

provisions specified in Section IV shall

be final when the extension expires if a

hearing request has not been received.

An answer or a request for hearing shall

not stay the immediate effectiveness of

this order.

Dated at Rockville, Maryland this 28th day

of April 1998.

For the Nuclear Regulatory Commission.

James Lieberman,

Director, Office of Enforcement.

[FR Doc. 98–12181 Filed 5–6–98; 8:45 am]

BILLING CODE 7590–01–P

NUCLEAR REGULATORY

COMMISSION

[Docket Nos. 50–282, 50–306]

Northern States Power Company

(Prairie Island Nuclear Generating

Plant, Units 1 and 2); Exemption

I

Northern States Power Company

(NSP, the licensee) is the holder of

Facility Operating License Nos. DPR–42

and DPR–60, which authorize operation

of Prairie Island Nuclear Generating

Plant, Units 1 and 2, respectively. The

licenses provide, among other things,

that the licensee is subject to all rules,

regulations, and orders of the

Commission now or hereafter in effect.

The facility consists of two

pressurized-water reactors located at the

licensee’s site in Goodhue County,

Minnesota.

II

In its letter dated March 6, 1998, the

licensee requested an exemption from

specific requirements of Title 10 of the

Code of Federal Regulations Part 50,

Section 60, and Appendix G.

Specifically, NSP proposed to use

American Society of Mechanical

Engineers (ASME) Code Case N–514 to

permit setting the pressure setpoint of

each unit’s overpressure protection

system (OPPS) so that the pressuretemperature (P-T) limits required by 10

CFR Part 50, Appendix G, could be

exceeded by 10 percent during a low

temperature pressure transient.

The NRC has established

requirements in 10 CFR Part 50 to

protect the integrity of the reactor

coolant system pressure boundary. As a

part of these, 10 CFR Part 50, Appendix

G, requires that P-T limits be established

for reactor pressure vessels during

normal operation, including anticipated

operational occurrences and vessel

hydrostatic testing and as stated in

Appendix G, ‘‘The appropriate

requirements on * * * the pressuretemperature limits * * * must be met

for all conditions.’’ In order to ensure

these P-T limit curves are not exceeded

and provide pressure relief during low

temperature overpressurization events,

pressurized-water reactor licensees have

installed protection systems (OPPS) as

part of the reactor coolant system

pressure boundary. NSP is required as

part of the Prairie Island Units 1 and 2

Technical Specifications to develop,

update, and submit reactor vessel P–T

limits and OPPS setpoints for NRC

review and approval.

By letter dated March 6, 1998, NSP

submitted an exemption request to

enable the use of ASME Code Case N–

514 as an alternative method for

determining the OPPS pressure setpoint.

NSP determined that the exemption

request from the provisions of 10 CFR

50.60 and Appendix G was necessary

since these regulations require, as noted

above, that the reactor vessel conditions

not exceed the P-T limits established by

Appendix G. In referring to 10 CFR

50.12 on specific exemptions, NSP cited

special circumstances as stated in 10

CFR 50.12(a)(2)(ii) on achieving the

underlying purpose of the regulations as

its basis for requesting this exemption.

III

Pursuant to 10 CFR 50.12, the

Commission may, upon application by

any interested person or upon its own

initiative, grant exemptions from the

requirements of 10 CFR Part 50 when (1)

the exemptions are authorized by law,

will not present an undue risk to public

health or safety, and are consistent with

the common defense and security, and

(2) when special circumstances are

present. Special circumstances are

present whenever, according to 10 CFR

50.12(a)(2)(ii), ‘‘Application of the

regulation in the particular

circumstances would not serve the

underlying purpose of the rule or is not

necessary to achieve the underlying

purpose of the rule.’’

The underlying purpose of 10 CFR

Part 50, Appendix G, is to establish

fracture toughness requirements for the

RCS pressure boundary to provide

adequate margins of safety during any

condition of normal operation. NSP

stated that the OPPS provides a physical

means of protecting the vessel by not

exceeding the limits. NSP proposed that

establishing the OPPS pressure setpoint

per the N–514 provisions such that the

vessel pressure would not exceed 110

percent of the P-T limit allowables

would still provide an acceptable level

of safety and mitigate the potential for

an inadvertent actuation of the OPPS.

The finding of an ‘‘acceptable level of

safety’’ while using N–514 was made

based on the conservatisms that have

been explicitly incorporated into the

procedure for developing the P-T limit

curves. This procedure, referenced from

Appendix G to Section XI of the ASME

Code, includes the following

conservatisms: (1) A safety factor of 2 on

the pressure stresses, (2) a margin factor

applied to the determination of RTNDT
Federal Register / Vol. 63, No. 88 /Thursday, May 7, 1998 / Notices 25243

[reference temperature nil ductility temperature]

(using Regulatory Guide 1.99 ‘‘Radiation

Embrittlement of Reactor Vessel

Materials,’’ Revision 2), and (3) a

limiting material toughness curve based

on bounding dynamic crack initiation

and crack arrest data.

In addition, NSP explained that plant

operators must operate the plant

between the minimum pressure

required to preserve reactor coolant

pump seals and a maximum pressure

that does not challenge the poweroperated relief valve setpoint. Without

the application of ASME Code Case N–

514, Prairie Island would have an

operating window that is too narrow to

permit reasonable system makeup and

pressure control. NSP continued by

stating that further reduction of the

OPPS pressure setpoint below 500 psig

would increase the probability that the

reactor coolant pump’s no. 1 seal will

fail as a result of OPPS operation, and

that such a seal failure could produce a

breach in the reactor coolant system

boundary that could not be isolated.

Therefore, inadvertent OPPS actuation

could lead to a small break loss-ofcoolant accident and the unnecessary

release of reactor coolant inside

containment.

IV

For the foregoing reasons, the NRC

staff has concluded that the licensee’s

proposed use of the alternate

methodology in determining the

acceptable setpoint for OPPS events will

not present an undue risk to public

health and safety and is consistent with

the common defense and security. The

NRC staff has determined that there are

special circumstances present, as

specified in 10 CFR 50.12(a)(2)(ii), in

that the application of 10 CFR 50.60 is

not necessary in order to achieve the

underlying purpose of this regulation.

The NRC staff agreed with NSP’s

determination that an exemption would

be required to approve the use of Code

Case N–514. The NRC staff examined

NSP’s rationale to support the

exemption request and concluded that

the use of Code Case N–514 would also

meet the underlying intent of the

regulations. Based upon a consideration

of the conservatisms that are explicitly

defined in the Appendix G methodology

(as listed in Section III above), the staff

concluded that permitting the OPPS

setpoint to be established such that the

vessel pressure would not exceed 110

percent of the limit defined by the P-T

limit curves would provide an adequate

margin of safety against brittle failure of

the reactor vessel. This is also consistent

with the determination that the staff has

reached for other licensees under

similar conditions based on the same

considerations. Therefore, requesting

the exemption under the special

circumstances of 10 CFR 50.12(a)(2)(ii)

was found to be appropriate. The staff

also agrees that limiting the potential for

inadvertent OPPS actuation (and

limiting the potential for reactor coolant

pump seal damage) may improve plant

safety.

Accordingly, the Commission has

determined that, pursuant to 10 CFR

50.12(a), an exemption is authorized by

law, will not endanger life or property

or common defense and security, and is

otherwise in the public interest.

Therefore, the Commission hereby

grants an exemption from the

requirements of 10 CFR 50.60 and

Appendix G to allow NSP to apply the

methods in ASME Code Case N–514 for

the determination of the Prairie Island

Nuclear Generating Plant Units 1 and 2

pressure setpoints.

Pursuant to 10 CFR 51.32, the

Commission has determined that the

granting of this exemption will not have

a significant effect on the quality of the

human environment (63 FR 23477).

This exemption is effective upon

issuance.

Dated at Rockville, Maryland, this 30th day

of April 1998.

For the Nuclear Regulatory Commission.

Samuel J. Collins,

Director, Office of Nuclear Reactor

Regulation.

[FR Doc. 98–12183 Filed 5–6–98; 8:45 am]

BILLING CODE 7590–01–P

NUCLEAR REGULATORY

COMMISSION

[Docket No. 50–259; License No. DPR–33]

Tennessee Valley Authority; Receipt of

Petition for Director’s Decision Under

10 CFR 2.206

Notice is hereby given that by petition

dated April 5, 1998, the Union of

Concerned Scientists, (or Petitioner),

has requested that the U.S. Nuclear

Regulatory Commission (NRC) take

action with regard to Browns Ferry

Nuclear Plant, Unit No. 1. Petitioner

requests (1) that the operating license

for Browns Ferry Unit 1 be revoked and

(2) that the NRC require the Tennessee

Valley Authority (TVA) to submit either

a decommissioning plan or a lay-up

plan for Browns Ferry Unit 1. Petitioner

further requests a hearing on this

petition to present new information on

Browns Ferry Unit 1 that would include

a discussion of the licensing basis

reconstitution that would be required to

support restart, and certain financial

aspects that might be a consideration for

the TVA’s decision for retaining the

Browns Ferry Unit 1 operating license.

As the basis for this request, the

Petitioner asserts that revocation of the

operating license and requiring

relicensing if TVA later decides to

restart Unit 1 is a better, safer process

than is the current Inspection Manual

Chapter 0350 restart process. Further,

the petition asserts that requiring a

decommissioning plan would provide

assurance that the irradiated fuel is

stored safely and that Units 2 and 3 are

sufficiently independent of Unit 1 for

safe operation.

The petition is being treated pursuant

to 10 CFR 2.206 of the Commission’s

regulations and has been referred to the

Director of the Office of Nuclear Reactor

Regulation. As provided by Section

2.206, appropriate action will be taken

on this petition within a reasonable

time.

By letter dated April 29, 1998, the

Director acknowledged receipt of the

petition and denied Petitioner’s request

for a public hearing to present new

information.

A copy of the petition is available for

inspection at the Commission’s Public

Document Room at 2120 L Street, NW.,

Washington, D.C. 20555.

Dated at Rockville, Maryland, this 29th day

of April 1998.

For the Nuclear Regulatory Commission.

Samuel J. Collins,

Director, Office of Nuclear Reactor

Regulation.

[FR Doc. 98–12178 Filed 5–6–98; 8:45 am]

BILLING CODE 7590–01–P

NUCLEAR REGULATORY

COMMISSION

[Docket No. 50–390]

Tennessee Valley Authority; 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. NPF–

90, issued to the Tennessee Valley

Authority (TVA or the licensee) for

operation of the Watts Bar Nuclear Plant

(WBN), Unit 1 located in Rhea County,

Tennessee.

WBN currently has two containment

hydrogen ignitors that are inoperable

due to an apparent fault in the common

circuit supplying these ignitors. This

condition renders Train A of the WBN