38198 Federal Register / Vol. 63, No. 135 /Wednesday, July 15, 1998 / Notices

NUCLEAR REGULATORY

COMMISSION

Biweekly Notice; Applications and

Amendments to Facility Operating

Licenses Involving No Significant

Hazards Considerations

I. Background

Pursuant to Pub. L. 97–415, the U.S.

Nuclear Regulatory Commission (the

Commission or NRC staff) is publishing

this regular biweekly notice. Pub. L. 97–

415 revised section 189 of the Atomic

Energy Act of 1954, as amended (the

Act), to require the Commission to

publish notice of any amendments

issued, or proposed to be issued, under

a new provision of section 189 of the

Act. This provision grants the

Commission the authority to issue and

make immediately effective any

amendment to an operating license

upon a determination by the

Commission that such amendment

involves no significant hazards

consideration, notwithstanding the

pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all

notices of amendments issued, or

proposed to be issued from June 22,

1998, through July 2, 1998. The last

biweekly notice was published on July

1, 1998 (63 FR 35986).

Notice of Consideration of Issuance of

Amendments to Facility Operating

Licenses, Proposed No Significant

Hazards Consideration Determination

and Opportunity for a Hearing

The Commission has made a

proposed determination that the

following amendment requests involve

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. The basis for this

proposed determination for each

amendment request is shown below.

The Commission is seeking public

comments on this proposed

determination. Any comments received

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 before

action is taken. Should the Commission

take this action, it will publish in the

Federal Register a notice of issuance

and provide for opportunity for a

hearing after 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 and

Directives Branch, Division of

Administration 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 a hearing and

petitions for leave to intervene is

discussed below.

By August 14, 1998, 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 for the particular

facility involved. 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 a 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.
Federal Register / Vol. 63, No. 135 /Wednesday, July 15, 1998 / Notices 38199

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, or

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 the

attorney for the licensee.

Nontimely filings of petitions for

leave to intervene, amended petitions,

supplemental petitions and/or requests

for a hearing will not be entertained

absent a determination by the

Commission, the presiding officer or the

Atomic Safety and Licensing Board that

the petition and/or request should be

granted based upon a balancing of

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 which is 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 for the particular

facility involved.

Commonwealth Edison Company,

Docket Nos. STN 50–454 and STN 50–

455, Byron Station, Unit Nos. 1 and 2,

Ogle County, Illinois

Docket Nos. STN 50–456 and STN 50–

457, Braidwood Station, Unit Nos. 1 and

2, Will County, Illinois

Date of amendment request: May 29,

1998.

Description of amendment request:

The proposed amendment would revise

the technical specifications to credit the

automatic function of the pressurizer

power operated relief valves (PORVs) to

provide mitigation for inadvertent safety

injection at power accident. The

limiting condition for operation and

surveillance requirements for the

PORVs would also be revised.

Basis for proposed no significant

hazards consideration determination:

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. The change does not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

The changes to the Technical

Specification (TS) Limiting Condition

for Operation (LCO), Surveillance

Requirements, and Bases do not involve

an increase in the probability or

consequences of the Inadvertent

Operation of Emergency Core Cooling

System (Spurious SI) at Power transient.

Crediting the PORVs in the maximum

pressurizer overfill case for this

transient does not increase the

probability of the occurrence of the

transient since the automatic function of

the PORVs for Reactor Coolant System

(RCS) pressure control is not an initiator

for the Spurious SI at Power transient.

This change allows for the NRC

Standard Review Plan (NUREG–0800)

acceptance criteria to be met for the

Spurious SI at Power transient, ensuring

that the consequences of this transient

remain within acceptable levels.

As documented in various Safety

Evaluation Reports (SERs) from the

NRC, the overpressure protection

function of the PORVs was not

originally considered to be a safety

related function. In response to Generic

Issue 70, the NRC performed a

regulatory analysis related to PORV and

block valve reliability in Pressurized

Water Reactor (PWR) plants. This

regulatory analysis is documented in

NUREG–1316, ‘‘Technical Findings and

Regulatory Analysis Related to Generic

Issue 70, Evaluation of Power-Operated

Relief Valve and Block Valve Reliability

in PWR Nuclear Power Plants,’’ where

the NRC staff concluded that it was not

cost effective to backfit non-safety

related PORVs to upgrade them to safety

related status to perform safety related

functions. The safety related functions

were those detailed in Section 2.1 of

NUREG–1316 and any other safety

related function identified in the future.

As an example, the PORVs are credited

for the cold overpressure protection

function of the reactor pressure vessel

during low temperature operations. The

analysis documented in this License

Amendment request demonstrates that

the PORVs provide an acceptable level

of quality and performance to allow

them to be credited to mitigate the

consequences of the Spurious SI at

Power transient documented in Byron

and Braidwood Updated Final Safety

Analysis Report (UFSAR) Section

15.5.1. The PORVs are equipped with

safety related actuators and safety

related accumulator tanks which

maintain valve function during a loss of

instrument air. The position indication

and control switches in the Main

Control Room (MCR) are safety related.

All pressurizer PORV open/close

functions and circuitry are supplied

with uninterruptible Class 1E power

supplies. The automatic portion of the

PORV circuitry which processes the

high pressurizer and high RCS pressure

at low temperature is designated nonsafety related and is isolated from the

safety related portions of the circuitry

by safety related interposing relays

which actuate on a faulted condition.

However, both Byron and Braidwood

Stations have implemented

modifications for both Units 1 and 2,

which ensure that automatic control of

both PORVs is available during loss of

offsite power conditions. In addition,

the PORV function is monitored within

the scope of the Maintenance Rule

Program and the postulated failure of

the PORV automatic function does not

result in unacceptable risk.

The probability of a Spurious SI at

Power transient is not affected by this

proposed change and the above analysis

demonstrates that the PORVs will

adequately function in automatic mode

to mitigate the consequences of the

transient. As such, there are no changes

in the type or amount of any effluent

released offsite as a result of this

change. Therefore, based on this

evaluation, this proposed amendment

does not involve a significant increase

in the probability or consequences of an

accident previously evaluated.

2. The change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.
38200 Federal Register / Vol. 63, No. 135 /Wednesday, July 15, 1998 / Notices

This proposed change does not create

the possibility of a new or different

accident from any accident previously

evaluated. This change would

specifically allow for the PORV

automatic function to be credited in

Modes 1, 2, and 3 for the Spurious SI

at Power transient only. This change

allows for added assurance that the

acceptance criteria as documented in

the NRC Standard Review Plan

(NUREG–0800) for ANS Condition II

transients will be met. The acceptance

criteria of concern is that a Condition II

transient must not lead to an event

(Condition III or IV) of more significant

consequences without additional

failures occurring. The PORV automatic

function is to be credited with

mitigating the maximum pressurizer

overfill case for the Spurious SI at

Power transient. This case has the

acceptance criteria that the pressurizer

must not go water solid prior to RCS

pressure reaching the setpoint of the

pressurizer safety relief valves (PSRVs).

This conservative acceptance criteria is

based on the fact that the PSRVs are not

qualified to pass subcooled water and

reseat, thereby creating a concern for an

uncontrolled release path from the RCS.

This proposed change helps ensure that

the acceptance criteria for this accident

are met. There is a small probability that

the PORV function, either automatic or

manual, would not successfully mitigate

this transient due to the failure of one

or both PORVs. However, the low

likelihood of a total failure of the PORV

function during the Spurious SI at

Power transient does not create a new

accident because a similar scenario is

already addressed by UFSAR Section

15.6.1, ‘‘Inadvertent Opening of a

Pressurizer Safety or Relief Valve.’’ The

UFSAR analysis for the Section 15.6.1

ANS Condition II transient indicates

that the radiological consequences of

this transient are significantly less than

that of a LOCA and are therefore,

acceptable. The same arguments for

radiological consequences apply to the

Spurious SI at Power transient in the

event the PORV automatic function fails

and water relief occurs through the

PSRVs.

The proposed change to the LCO

requirements in TS Section 3/4.4.4

would allow for the PORV block valve

to be closed but remain energized in the

event a PORV was considered

inoperable due to the automatic

actuation circuitry. Currently, the PORV

block valve is closed but remains

energized only if a PORV is considered

inoperable due to excessive seat leakage.

The proposed change would extend the

allowance to include the circumstance

where the PORV was inoperable due to

the automatic actuation circuitry. This

allows a PORV to remain functional in

the manual mode for other safety related

functions consistent with the discussion

contained in NRC NUREG–1316.

However, this revised LCO requirement

would not represent a new failure mode

or accident over what has been

previously evaluated.

In summary, the proposed changes

documented in this TS amendment to

credit the automatic PORV function and

to revise the TS LCO requirements for

PORV inoperability do not create the

potential for any new or different

accidents from what was previously

evaluated.

3. The change does not involve a

significant reduction in a margin of

safety.

The current TS bases do not credit the

function of the pressurizer PORVs for

any Mode 1, 2, or 3 transients. This

change would allow for the PORV

automatic function to be credited for the

Spurious SI at Power transient only.

This does not represent a significant

reduction in the margin of safety. This

change would allow for the conservative

acceptance criteria for the current

UFSAR design analysis to be met. The

PORVs are reliable and are maintained

in a manner consistent with their

proposed safety related function to

mitigate the Spurious SI at Power

transient. This proposed change would

not result in a significant increase in

risk or consequences, and therefore,

does not involve a significant reduction

in the margin of safety.

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

requested amendments involve no

significant hazards consideration.

Local Public Document Room

location: For Byron, the Byron Public

Library District, 109 N. Franklin, P.O.

Box 434, Byron, Illinois 61010; for

Braidwood, the Wilmington Public

Library, 201 S. Kankakee Street,

Wilmington, Illinois 60481.

Attorney for licensee: Michael I.

Miller, Esquire; Sidley and Austin, One

First National Plaza, Chicago, Illinois

60603.

NRC Project Director: Stuart A.

Richards.

Florida Power Corporation, et al.,

Docket No. 50–302, Crystal River

Nuclear Generating Plant, Unit No. 3,

Citrus County, Florida

Date of amendment request: May

28,1998.

Description of amendment request:

The proposed amendment proposes

changes to the Final Safety Analysis

Report (FSAR) to include a description

of the use of Generic Letter (GL) 87–11,

‘‘Relaxation in Arbitrary Intermediate

Pipe Rupture Requirements,’’ and

NUREG/CR–2913, ‘‘Two-Phase Jet

Loads,’’ as a part of the approved

licensing basis and design basis for

Crystal River Unit 3. GL 87–11 will be

used to determine where high energy

line breaks (HELB) are postulated to

occur for high energy lines located

inside the Reactor Building (RB) and

analyzed in accordance with the

guidelines described in USAS B31.1.0–

1967, ‘‘USA Standard Code for Pressure

Piping, Power Piping.’’ NUREG/CR–

2913 will be used to determine the

effects of the resultant jet impingement

from postulated Reactor Coolant System

(RCS) piping ruptures on safety-related

systems, structures, and components

(SSCs).

Basis for proposed no significant

hazards consideration determination:

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. Involve a significant increase in the

probability or consequences of an

accident previously evaluated?

The use of new design methodologies

for determining postulated break

locations of RCS piping and other high

energy lines located inside containment,

and the dynamic effects of postulated

ruptures of RCS piping on SSCs

required for safe shutdown or accident

mitigation, does not impact the design

of these high energy lines such that

previously analyzed ruptures would

now be more likely to occur. The

approval of the license amendment will

not result in an actual modification to

RCS piping or other high energy lines

which would reduce their design

capabilities to maintain pressure

boundary integrity during normal

operating and accident conditions. By

using these new design methodologies,

protection of SSCs required for accident

mitigation is assured. Protection of SSCs

required for accident mitigation will

continue to be assured by use of these

well-defined design methodologies if

modifications to those SSCs are

implemented in the future. Therefore,

there will be no reduction in the

capability of those SSCs in limiting the

consequences of previously evaluated

accidents, and the proposed amendment

does not significantly increase the

probability or consequence of an

accident previously evaluated.
Federal Register / Vol. 63, No. 135 /Wednesday, July 15, 1998 / Notices 38201

2. Create the possibility of a new or

different kind of accident from

previously evaluated accidents?

The use of new design methodologies

for determining postulated break

locations of RCS piping and other high

energy lines located inside containment,

and the dynamic effects of postulated

ruptures of RCS piping on SSCs

required for safe shutdown or accident

mitigation, does not impact the design

of these high energy lines such that

previously unanalyzed ruptures would

now occur. The approval of the license

amendment will not result in an actual

modification to RCS piping or other

high energy lines which would reduce

their design capabilities to maintain

pressure boundary integrity during

normal operating and accident

conditions. By using these new design

methodologies, the current design of

RCS piping and other high energy lines

located inside containment can be

shown to include sufficient design

margin to prevent unanalyzed ruptures

from occurring. Therefore, use of these

design methodologies instead of the

previous licensing basis requirements

cannot create the possibility of a new or

different kind of accident.

3. Involve a significant reduction in a

margin of safety?

The use of new design methodologies

for determining postulated break

locations of RCS piping and other high

energy lines located inside containment,

and the dynamic effects of postulated

ruptures of RCS piping on SSCs

required for safe shutdown or accident

mitigation, does not impact the design

of these high energy lines such that

unanalyzed ruptures would now occur,

and cannot create a reduction in the

margin of safety for those ruptures of

high energy lines previously analyzed.

The approval of the license amendment

will not result in an actual modification

to RCS piping or other high energy lines

which would reduce their design

capabilities to maintain pressure

boundary integrity during normal

operating and accident conditions. By

using these new design methodologies,

protection of SSCs required for accident

mitigation is assured. Protection of SSCs

required for accident mitigation will

continue to be assured by use of these

well-defined design methodologies if

modifications to those SSCs are

implemented in the future. Therefore,

the capability of those SSCs to limit the

consequences of previously evaluated

accidents at levels below the approved

acceptance limits will continue to be

assured. As a result, use of these design

methodologies instead of the previous

licensing basis and design basis

requirements cannot significantly

reduce the existing margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Local Public Document Room

location: Coastal Region Library, 8619

W. Crystal Street, Crystal River, Florida

34428.

Attorney for licensee: R. Alexander

Glenn, General Counsel, Florida Power

Corporation, MAC—A5A, P. O. Box

14042, St. Petersburg, Florida 33733–

4042.

NRC Project Director: Frederick J.

Hebdon.

GPU Nuclear, Inc. et al., Docket No. 50–

219, Oyster Creek Nuclear Generating

Station, Ocean County, New Jersey

Date of amendment request: May 28,

1998.

Description of amendment request:

Revision of Technical Specification (TS)

4.5.A.1 such that the first Type A test

required by the primary containment

leakage rate testing program be

performed during refueling outage 18

rather than refueling outage 17.

Basis for proposed no significant

hazards consideration determination:

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. The proposed TS change does not

involve a significant increase in the

probability or consequences of an

accident previously evaluated.

The proposed changes does not alter

the design, function or manner of

operation of any structures, systems or

components. As a result, the proposed

change does not affect any of the

parameters or conditions that could

contribute to initiation of any accidents.

NUREG–1493 found that the effect of

containment leakage on overall accident

risk is small since risk is dominated by

accident sequences that result in failure

or bypass of the containment. The major

contributor to the total identified

leakage from Primary Containment

comes from Type B and C tested

components. Only a small portion of the

total leakage is detectable soley through

Type A testing. The leaks that have been

found by Type A tests have been only

marginally above existing requirements.

In addition, Oyster Creek has two means

(monitoring nitrogen use and

performing torus to drywell vacuum

breaker leak tests) of detecting gross

containment leakage. The proposed

change does not alter the requirements

to perform Type B and C testing in

accordance with the Primary

Containment Leakage Rate Testing

Program and does not affect the ability

of the facility to mitigate the

consequences of an accident.

Therefore, the proposed TS change

does not involve an significant increase

in the probability or consequences of an

accident previously evaluated.

2. The proposed TS change does not

create the possibility of a new or

different kind of accident from any

previously evaluated.

Deferring the Type A test for an

operating cycle does not alter the

design, function or manner of operation

of any structures, systems or

components. The proposed change does

not affect any of the parameters or

conditions that could contribute to

initiation of any accidents nor does it

introduce any new mechanisms which

could contribute to the creation of a new

or different kind of accident than

previously evaluated.

3. The proposed TS changes do not

involve a significant reduction in a

margin of safety.

The proposed change does not alter

the design, function or manner of

operation of any structures, systems or

components. The proposed change does

not impact the primary containment

system’s ability to provide a barrier

against the uncontrolled release of

fission products in the event of a break

in the reactor coolant system nor does

the proposed change impact the primary

containment accident leak rate. In

addition, NUREG–1493’s Summary of

Technical Findings states ‘‘Reducing the

frequency of Type A tests (ILRTs) from

the current three per 10 years to one per

20 years was found to lead to an

imperceptible increase in risk. The

estimated increase in risk is very small

because ILRTs identify only a few

potential containment leakage paths that

cannot be identified by Type B and C

testing, and the leaks that have been

found by Type A tests have been only

marginally above existing

requirements.’’ Therefore, the proposed

TS change does not involve a significant

reduction in a margin of safety.

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.

Local Public Document Room

location: Ocean County Library,
38202 Federal Register / Vol. 63, No. 135 /Wednesday, July 15, 1998 / Notices

Reference Department, 101 Washington

Street, Toms River, NJ 08753.

Attorney for licensee: Ernest L. Blake,

Jr., Esquire. Shaw, Pitman, Poets &

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037.

NRC Project Director: Cecil O.

Thomas.

Northeast Nuclear Energy Company

(NNECO), et al., Docket No. 50–423,

Millstone Nuclear Power Station, Unit

No. 3, New London County, Connecticut

Date of amendment request: June 10,

1998.

Description of amendment request:

The proposed revision to the Millstone

Unit 3 licensing basis would address

post-accident mitigation activities, vital

area access travel routes, and time.

NNECO determined that the Final

Safety Analysis Report (FSAR)

description of post-accident vital area

routing was out of date because the

radiological control area boundary fence

created an access problem on the

designated routes to the hydrogen

recombiner and fuel building. The

FSAR change would revise the routes to

accommodate the fence location and

allow for the time to unlock gates.

Basis for proposed no significant

hazards consideration determination:

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:

NNECO has reviewed the proposed

revision in accordance with 10 CFR

50.92 and has concluded that the

revision does not involve a significant

hazards consideration (SHC). The basis

for this conclusion is that the three

criteria of 10 CFR 50.92(c) are not

satisfied. The proposed revision does

not involve an SHC because the revision

would not:

1. Involve a significant increase in the

probability or consequence of an

accident previously evaluated.

Final Safety Analysis Report (FSAR)

Section 12.3.1.3.2, Post-accident access

to vital areas, and its associated Figures

and Tables are being updated. The

current FSAR descriptions are out of

date and as such do not include all

required post-accident actions.

Therefore, this FSAR change adds

actions to those listed in the FSAR as

well as incorporating the recalculation

of the doses associated with the

required post-accident actions. The dose

calculations utilize the appropriate postaccident source terms, area access

requirements and stay times, including

the appropriate routes to the areas. The

calculations show that for all design

basis required post-accident actions the

calculated dose to the Operators/

Emergency workers performing those

actions remains below the 5 rem

criterion of General Design Criteria

(GDC) 19. The revision to the FSAR

provides the required post-accident

required operator actions. Changing the

FSAR to include the current postaccident vital access requirements and

associated information for the

supporting dose calculations [cannot]

cause an accident. In addition, the

calculated dose to the Operators/

Emergency workers for all design basis

required actions is below the GDC 19

limit of 5 rem.

Therefore, the proposed revision does

not involve a significant increase in the

probability or consequence of an

accident previously evaluated.

2. Create the possibility of a new or

different kind of accident from any

accident previously evaluated.

The change is to the calculated postaccident vital access dose analyses and

the FSAR description of that analyses.

No new procedural Operator/Emergency

worker actions are associated with the

change.

However, since the information in the

FSAR was outdated, there are Operator/

Emergency actions being added to the

FSAR. Dose calculations associated with

those actions have been performed

utilizing the appropriate assumptions

with respect to source terms, vital area

access travel routes and stay times, and

times when the post-accident actions

would be performed. The analyses

confirmed that the calculated doses

associated with all required postaccident actions are less than the 5 rem

limit of GDC 19. There are no changes

to the Emergency Operating Procedures

associated with this change.

Therefore, the proposed revision does

not create the possibility of a new or

different kind of accident from any

accident previously evaluated.

3. Involve a significant reduction in a

margin of safety.

The dose calculations confirm that the

calculated dose associated with all

design basis post-accident Operator/

Emergency worker actions is below the

limit of 5 rem of GDC 19. There is one

action, initiation of hydrogen purge, for

which the calculated dose to the

Operator/Emergency worker exceeds 5

rem. This action is a backup means of

limiting the hydrogen concentration

inside containment post-accident. This

action would only be performed for

multiple failures which would disable

both trains of the safety-grade hydrogen

recombiner system. As such this action

is not a required design basis action and

does not need to meet the 5 rem limit.

The calculated dose for this action is

below the 25 rem limit that is specified

in the Station Emergency Plan for severe

accident mitigation actions.

Therefore, the proposed revision does

not involve a significant reduction in a

margin of safety.

In conclusion, based on the

information provided, it is determined

that the proposed revision does not

involve an SHC.

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.

Local Public Document Room

location: Learning Resources Center,

Three Rivers Community-Technical

College, 574 New London Turnpike,

Norwich, Connecticut, and the

Waterford Library, ATTN: Vince

Juliano, 49 Rope Ferry Road, Waterford,

Connecticut.

Attorney for licensee: Lillian M.

Cuoco, Esq., Senior Nuclear Counsel,

Northeast Utilities Service Company,

P.O. Box 270, Hartford, Connecticut.

NRC Deputy Director: Phillip F.

McKee.

PECO Energy Company, Public Service

Electric and Gas Company, Demarva

Power and Light Company, and Atlantic

City Electric Company, Dockets Nos. 50–

277 and 50–278, Peach Bottom Atomic

Power Station, Units Nos. 2 and 3, York

County, Pennsylvania

Date of application for amendments:

February 4, 1998.

Description of amendment request:

The proposed amendment would revise

the Technical Specifications (TS)

Surveillance Requirement (SR)

concerning Secondary Containment

doors at Peach Bottom Atomic Power

Station, Units 2 and 3.

Basis for proposed no significant

hazards consideration determination:

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. The proposed TS changes do not

involve a significant increase in the

probability or consequences of an

accident previously evaluated.

TS SR 3.6.4.1.2 will be revised to

require either all inner or outer

secondary containment access doors to

be closed in each air lock. This revision

will not adversely affect the ability of

the Secondary Containment to mitigate

the radiological consequences of a Lossof-Coolant Accident or fuel handling

accident, and does not involve a
Federal Register / Vol. 63, No. 135 /Wednesday, July 15, 1998 / Notices 38203

significant increase in the probability or

consequences of an accident previously

evaluated. During those times that one

or more inner (or outer) doors are open,

the closed outer (or inner) doors will

serve as the Secondary Containment

boundary.

Allowing certain inner or outer

Secondary Containment access doors in

an air lock to be open does not

compromise the design of the Secondary

Containment. No commitment is made

in the UFSAR to consider the single

failure of passive structural components

such as Secondary Containment doors.

As discussed in Section 1.5 of the

UFSAR, ‘‘* * * Essential safety actions

shall be carried out by equipment of

sufficient redundance and

independence that no single failure of

active components can prevent the

required actions’’. The same UFSAR

section goes on to state that, ‘‘For

systems or components to which IEEE–

279 (1968) is applicable, single failures

of passive electrical components are

considered, as well as single failures of

active components, in recognition of the

higher anticipated failure rates of

passive electrical components relative to

passive mechanical components.’’

Therefore, based on this UFSAR

discussion, it is concluded that failure

of outer (inner) secondary containment

doors need not be postulated with the

inner (outer) door being open.

The performance of the Secondary

Containment and the Standby Gas

Treatment System is unaffected by this

activity. Surveillance testing will prove

the capability to maintain Secondary

Containment with only inner or only

outer doors closed. This change will not

result in greater or more frequent

loading of Secondary Containment

doors, and does not result in changes

that impact the reliability of the

Secondary Containment and the

Standby Gas Treatment System.

2. The proposed TS changes do not

create the possibility of a new or

different kind of accident from any

accident previously evaluated.

The Secondary Containment, in

conjunction with the Standby Gas

Treatment System, provides the means

for mitigating the radiological

consequences of an accident. The

configuration of the Secondary

Containment has no effect on accident

initiators which lead to a new or

different kind of accident. This change

will not involve any changes to plant

systems, structures, or components

which could act as new accident

initiators. The design, function, and

reliability of Secondary Containment

and the Standby Gas Treatment System

are also not impacted by this change.

Therefore, this change will not create

the possibility of a new or different kind

of accident from any previously

evaluated.

3. The proposed TS changes do not

involve a significant reduction in a

margin of safety.

No margins of safety are reduced as a

result of this change to the TS. No safety

limits will be changed as a result of this

TS change. The Secondary Containment

will continue to perform its intended

safety function of limiting the ground

level release of airborne radioactive

materials and to provide a means for

controlled elevated release of the

building atmosphere so that off-site

doses from the postulated design basis

accidents are below the limits of 10 CFR

100. The design and reliability of the

Secondary Containment are also not

impacted as a result of this change.

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.

Local Public Document Room

location: Government Publications

Section, State Library of Pennsylvania,

(REGIONAL DEPOSITORY) Education

Building, Walnut Street and

Commonwealth Avenue, Box 1601,

Harrisburg, PA 17105.

Attorney for Licensee: J. W. Durham,

Sr., Esquire, Sr. V.P. and General

Counsel, PECO Energy Company, 2301

Market Street, Philadelphia, PA 19101.

NRC Project Director: Robert A. Capra.

Power Authority of the State of New

York, Docket No. 50–286, Indian Point

Nuclear Generating Unit No. 3,

Westchester County, New York

Date of amendment request:

November 13, 1997.

Description of amendment request:

The proposed amendment will reduce

the maximum test interval from 1 year

to 6 months for the test frequency of the

main turbine stop and control valves

(TS & CVs) in Table 4.1–3 and add a

footnote.

Basis for proposed no significant

hazards consideration determination:

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 the proposed license

amendment involve a significant

increase in the probability or

consequences of an accident previously

evaluated?

Response

The proposed license amendment

does not involve a significant increase

in the probability or consequences of an

accident previously evaluated. The

proposed change increases the

frequency of testing of the TS & CVs by

reducing the maximum allowable test

interval. The maximum test interval is

reduced from one year to six months.

Thus, the proposed change will make

the maximum test interval more

conservative. Therefore, the proposed

change does not involve a significant

increase in the probability or

consequences of an accident previously

evaluated.

(2) Does the proposed license

amendment create the possibility of a

new or different kind of accident from

any accident previously evaluated?

Response

The proposed license amendment

does not create the possibility of a new

or different kind of accident from any

accident previously evaluated. The

proposed change does not involve the

addition of any new or different type of

equipment, nor does it involve the

operation of equipment required for safe

operation of the facility in a manner

different from those addressed in the

Final Safety Analysis Report.

(3) Does the proposed license

amendment involve a significant

reduction in a margin of safety?

Response

The proposed license amendment

does not involve a significant reduction

in a margin of safety. The proposed

change does not adversely affect

performance of any safety related

system or component, instrument

operation, or safety system setpoints

and does not result in increased severity

of any accidents considered in the safety

analysis. The proposed change does not

reduce the frequency of testing of these

valves but updates the methodology for

determination of the test frequency and

reduces the maximum test interval from

one year to six months. It establishes a

more conservative acceptance criteria of

5.0 × 10¥6 per year than the NRC

acceptance criteria of 1.0 × 10¥5 for a

turbine missile event. Therefore, the

proposed change does not involve a

significant reduction in the margin of

safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.
38204 Federal Register / Vol. 63, No. 135 /Wednesday, July 15, 1998 / Notices

Local Public Document Room

location: White Plains Public Library,

100 Martine Avenue, White Plains, New

York 10601.

Attorney for licensee: Mr. David E.

Blabey, 10 Columbus Circle, New York,

New York 10019.

NRC Project Director: S. Singh Bajwa,

Director.

Power Authority of the State of New

York, Docket No. 50–333, James A.

FitzPatrick Nuclear Power Plant,

Oswego County, New York

Date of amendment request: June 16,

1998.

Description of amendment request:

The proposed amendment would

relocate the Safety Review Committee

review, audit and related record keeping

requirements from the Technical

Specifications (TSs) to Chapter 17 of the

Final Safety Analysis Report (FSAR)

(i.e., Quality Assurance Program).

Basis for proposed no significant

hazards consideration determination:

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 the proposed license

amendment involve a significant

increase in the probability or

consequences of an accident previously

analyzed?

Response

This amendment application does not

involve a significant increase in the

probability or consequences of an

accident previously analyzed. The

relocation of the SRC [Safety Review

Committee] review, audit, and related

record keeping requirements from the

TS to the FSAR does not alter the

performance or frequency of these

activities. Future changes to the QA

[Qualify Assurance] program, located in

Chapter 17 of the FSAR, which

constitute a reduction in commitments,

are governed by 10 CFR 50.54(a).

Therefore, sufficient controls for these

requirements exist and these changes do

not involve a significant increase in the

probability or consequences of an

accident previously analyzed.

(2) Does the proposed license

amendment create the possibility of a

new or different kind of accident from

any accident previously evaluated?

Response

This amendment application does not

create the possibility of a new or

different kind of accident from any

accident previously evaluated. The

proposed changes involve the relocation

of SRC requirements from the TS to the

FSAR. Relocation of these requirements

does not affect plant equipment or the

way the plant operates. The reviews,

audits, and record keeping will continue

to be performed in the identical manner

as they are currently being performed.

Therefore, the proposed revisions

cannot create a new or different kind of

accident.

(3) Does the proposed amendment

involve a significant reduction in a

margin of safety?

Response

This amendment application does not

involve a significant reduction in a

margin of safety. The requested

Technical Specification revisions

relocate SRC review, audit and related

record keeping requirements from the

TS to the FSAR. These requirements are

not being altered by this relocation. The

reviews, audits, and record keeping will

continue to be performed in the

identical manner as they are currently

being performed. Any changes to these

requirements which constitute a

reduction in commitments will be

processed in accordance with 10 CFR

50.54(a). Therefore, the proposed

changes do not involve a significant

reduction in a margin of safety.

Local Public Document Room

location: Reference and Documents

Department, Penfield Library, State

University of New York, Oswego, New

York 13126.

Attorney for licensee: Mr. David

Blabey, 10 Columbus Circle, New York,

New York 10019.

NRC Project Director: S. Singh Bajwa.

Tennessee Valley Authority, Docket

Nos. 50–327 and 50–328, Sequoyah

Nuclear Plant, Units 1 and 2, Hamilton

County, Tennessee

Date of application for amendments:

April 30, 1998 (TS 98–01).

Brief description of amendments: The

amendments would change the

Sequoyah (SQN) Technical

Specifications (TSs) to allow

surveillance testing of the reactor

coolant system (RCS) pressurizer poweroperated relief valves (PORVs) in Modes

3, 4, and 5.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a),

Tennessee Valley Authority (TVA), the

licensee, has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

TVA has concluded that operation of

SQN Units 1 and 2, in accordance with

the proposed change to the TSs, does

not involve a significant hazards

consideration. TVA’s conclusion is

based on its evaluation, in accordance

with 10 CFR 50.91(a)(1), of the three

standards set forth in 10 CFR 50.92(c).

A. The proposed amendment does not

involve a significant increase in the

probability or consequences of an

accident previously evaluated.

The possibility of occurrence or the

consequences for an accident or

malfunction of equipment is not

increased as the test conditions for the

PORVs in Mode 5 are representative

conditions based on a steam bubble

being present, and testing in this mode

is more conservative, if RCS pressure is

less, since there is less fluid force to aid

the solenoid force in opening the valve.

Testing in Modes 3 and 4 was the initial

request of GL [Generic Letter] 90–06. No

changes are proposed to operation of the

PORV block valves. Offsite dose

consequences are unchanged by this

request.

B. The proposed amendment does not

create the possibility of a new or

different kind of accident from any

accident previously evaluated.

A possibility for an accident or

malfunction of a different type than any

evaluated previously in SQN’s Final

Safety Analysis Report is not created;

nor is the possibility for an accident or

malfunction of a different type. A new

test method is not required. No new

failure modes are introduced.

C. The proposed amendment does not

involve a significant reduction in a

margin of safety.

The margin of safety has not been

reduced for testing in Mode 5 since the

proposed test conditions are equal to or

more conservative, if RCS pressure is

less, than those currently in use with

existing SRs [surveillance

requirements]. Testing in Modes 3 and

4 was the initial request of GL 90–06.

The results of the accident analysis

remain unchanged by this request.

The NRC 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.

Local Public Document Room

location: Chattanooga-Hamilton County

Library, 1001 Broad Street, Chattanooga,

Tennessee 37402.

Attorney for licensee: General

Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET 10H,

Knoxville, Tennessee 37902.

NRC Project Director: Frederick J.

Hebdon.
Federal Register / Vol. 63, No. 135 /Wednesday, July 15, 1998 / Notices 38205

Tennessee Valley Authority, Docket

Nos. 50–327 and 50–328, Sequoyah

Nuclear Plant, Units 1 and 2, Hamilton

County, Tennessee

Date of application for amendments:

June 26, 1998 (TS 98–02).

Brief description of amendments: The

amendments would change the

Sequoyah Nuclear Plant (SQN)

Technical Specifications (TS) and their

Bases to lower the specific activity of

the primary coolant from 1.0

microcurie/gram dose equivalent

iodine-131 to 0.35 microcurie/gram, as

provided for in NRC Generic Letter 95–

05, ‘‘Voltage-Based Repair Criteria for

Westinghouse Steam Generator Tubes

Affected by Outside Diameter Stress

Corrosion Cracking.’’ This change

allows a proportional increase in main

steam line break induced primary-tosecondary leakage when implementing

the alternate steam generator tube repair

criteria, which the NRC has already

approved for Units 1 and 2.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

Tennessee Valley Authority (TVA), the

licensee, has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

TVA has concluded that operation of

SQN Units 1 and 2, in accordance with

the proposed change to the TS [or

operating license(s)], does not involve a

significant hazards consideration. TVA’s

conclusion is based on its evaluation, in

accordance with 10 CFR 50.91(a)(1), of

the three standards set forth in 10 CFR

50.92(c).

A. The proposed amendment does not

involve a significant increase in the

probability or consequences of an

accident previously evaluated.

The proposed TS change lowers the

[maximum allowable] reactor coolant

specific activity, which allows an

increase in the leakage quantity that

would be postulated to occur during a

MSLB accident. This in turn allows a

larger quantity of tubes with axial

ODSCC to remain in service. The

methodology for identifying and

defining the ODSCC and for developing

the leakage quantity remains

unchanged. Therefore, the proposed

change does not result in a significant

increase in the probability of an

accident.

An increase in the consequences of an

accident would not occur because the

proportional decrease in reactor coolant

specific activity, while proportionally

increasing the primary-to-secondary

leakage during a postulated MSLB

accident, has been evaluated to confirm

the amount of activity released to the

environment remains unchanged. The

evaluation uses the same methodology

used to establish the original primaryto-secondary leak limits in

[Westinghouse Topical Report] WCAP–

13990.

The control room dose, the low

population zone dose, and the dose at

the exclusion area boundary remains

bounded by the acceptance criteria of

NUREG–0800 and continue to satisfy an

appropriate fraction of the 10 CFR 100

dose limits and GDC [General Design

Criterion] 19. Therefore, the proposed

TS change does not result in a

significant increase in the consequences

of an accident previously analyzed.

B. The proposed amendment does not

create the possibility of a new or

different kind of accident from any

accident previously evaluated.

The proposed TS change does not

alter the configuration of the plant. The

changes do not directly affect plant

operation. The change will not result in

the installation of any new equipment

or systems or the modification of any

existing equipment or systems. No new

operating procedures, conditions or

modes will be created by this proposed

change. SG [steam generator] tube

structural integrity, as defined in draft

Regulatory Guide 1.121, remains

unchanged.

Therefore the possibility of a new or

different kind of accident from any

accident previously evaluated is not

created.

C. The proposed amendment does not

involve a significant reduction in a

margin of safety.

Lowering the reactor coolant specific

activity, while allowing the proportional

increase in the primary-to-secondary

leakage during a postulated MSLB

accident, keeps the amount—of activity

released to the environment unchanged.

Design basis and offsite dose calculation

assumptions remain satisfied. Therefore,

the proposed change does not result in

a significant reduction in the margin of

safety.

The NRC 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.

Local Public Document Room

location: Chattanooga-Hamilton County

Library, 1001 Broad Street, Chattanooga,

Tennessee 37402.

Attorney for licensee: General

Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET 10H,

Knoxville, Tennessee 37902.

NRC Project Director: Frederick J.

Hebdon.

TU Electric Company, Docket Nos. 50–

445 and 50–446, Comanche Peak Steam

Electric Station, Units 1 and 2,

Somervell County, Texas

Date of amendment request: February

27, 1998 (TXX–98033), June 10, 1998

(TXX 98145).

Brief description of amendments: The

proposed amendment would increase

the RWST Low-Low level setpoint from

‘‘greater than or equal to 40%’’ to

‘‘greater than or equal to 45%’’ of span

for CPSES, Units 1 and 2. The change

raises the RWST Low-Low level setpoint

in order to increase the volume

available to complete containment spray

switchover without turning off the

containment spray pumps.

Basis for proposed no significant

hazards consideration determination:

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. Do the proposed changes involve a

significant increase in the probability or

consequences of an accident previously

evaluated?

The changes in the License

Amendment Request proposes more

restrictive setpoint Allowable Values for

the RWST Low-Low setpoint. This more

restrictive value assures that all

applicable safety analysis limits are

being met. Changing an RWST Low-Low

setpoint from greater than or equal to

40% to greater than or equal to 45% in

the Technical Specifications has no

impact on the probability of occurrence

of any accident previously evaluated.

None of the accident analyses were

affected, therefore, the consequences of

all previously evaluated accidents

remain unchanged.

2. Do the proposed changes create the

possibility of a new or different kind of

accident from any accident previously

evaluated?

The proposed changes involve the use

of a more conservative value for the

RWST Low-Low setpoint. As such, none

of the changes effect plant hardware or

the operation of plant systems in a way

that could initiate an accident.

Therefore, the proposed changes do not

create the possibility of a new or

different kind of accident from any

accident previously evaluated.

3. Do the proposed changes involve a

significant reduction in a margin of

safety?

There were no changes made to any

of the accident analyses or safety

analysis limits as a result of this

proposed change. Further, the proposed
38206 Federal Register / Vol. 63, No. 135 /Wednesday, July 15, 1998 / Notices

change does not affect the acceptance

criteria for any analyzed event. ECCS,

Containment spray, and the RWST will

remain capable of performing their

safety function, and the new

requirement will continue to provide

adequate assurance of that capability.

Raising the RWST Low-Low setpoint

from 40% to 45% has no impact on the

assumptions used in the safety analysis

as discussed in Chapter 15 of the FSAR.

The margin of safety established by the

Limiting Conditions for Operation also

remains unchanged. Thus there is no

effect on the margin of safety.

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 are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Local Public Document Room

location: University of Texas at

Arlington Library, Government

Publications/Maps, 702 College, P.O.

Box 19497, Arlington, TX 76019.

Attorney for licensee: George L. Edgar,

Esq., Morgan, Lewis and Bockius, 1800

M Street, N.W., Washington, DC 20036.

NRC Project Director: John N.

Hannon.

Virginia Electric and Power Company,

Docket Nos. 50–280 and 50–281, Surry

Power Station, Unit Nos. 1 and 2, Surry

County, Virginia

Date of amendment request: June 19,

1998. This amendment request

supersedes the November 5, 1997,

submittal in its entirety (63 FR 19981).

Description of amendment request:

The proposed Operating License change

and changes to the technical

specifications (TS) would permit the use

of a temporary alternate supply line

(jumper) to provide service water (SW)

to the component cooling heat

exchangers. The temporary jumper will

permit maintenance to be performed on

the existing supply line.

Basis for proposed no significant

hazards consideration determination:

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:

Virginia Electric and Power Company

has reviewed the proposed changes

against the criteria of 10 CFR 50.92 and

has concluded that the changes do not

pose a significant safety hazards

consideration as defined therein. The

proposed Operating License and

Technical Specifications and Bases

changes are necessary to allow the use

of a temporary, seismic, non-missile

protected jumper to provide service

water (SW) to the Component Cooling

Heat Exchangers (CCHXs) while

maintenance work is performed on the

existing SW supply line to the CCHXs.

Since there is only one SW supply line

to the CCHXs, an alternate SW supply

must be provided whenever the line is

removed from service. The temporary

jumper provides this function. The

jumper will only be used for a 35-day

period during each of two Unit 1

refueling outages.

The use of the temporary jumper has

been thoroughly evaluated, and

appropriate constraints and

compensatory measures (including a

Contingency Action Plan) have been

developed to ensure that the temporary

jumper is reliable, safe, and suitable for

its intended purpose. A complete and

immediate loss of SW supply to the

operating CCHXs is not considered

credible, given the project constraints

and the unlikely probability of a

generated missile or heavy load drop.

Existing station abnormal procedures

already address a loss of component

cooling, and the use of alternate cooling

for a loss of decay heat removal, in the

unlikely event that they are required.

Furthermore, appropriate mitigative

measures have been identified to

address potential flooding concerns.

The minor administrative changes

merely correct a table format

inconsistency and update Basis section

references.

Consequently, the operation of Surry

Power Station with the proposed

amendment and license condition will

not:

1. Involve a significant increase in the

probability or consequences of an

accident previously evaluated.

The SW and CC Systems will function

as designed under the Unit operating

constraints specified by this project (i.e.,

Unit 2 in operation and Unit 1 in a

refueling outage), and the potential for

a loss of component cooling is already

addressed by Station Abnormal

Procedures. Therefore, there is no

increase in the probability of an

accident previously evaluated. The

possibility of flooding due to failure of

the temporary SW supply jumper in the

Turbine Building basement has been

evaluated and dispositioned by the

implementation of appropriate

precautions and compensatory measures

to preclude damage to the temporary

jumper and to respond to a postulated

flooding event. A flood watch will be

present around-the-clock with authority

and procedural guidance to isolate the

jumper, if required. Furthermore, the

CCHXs serve no design basis accident

mitigating function. Therefore, the

consequences of an accident previously

evaluated are not increased.

2. Create the possibility of a new or

different kind of accident from any

accident previously evaluated.

The SW and CC Systems’ design

functions and basic configurations are

not being altered as a result of using a

temporary SW supply jumper. The

temporary jumper is designed to be

safety-related and seismic with all of the

design attributes of the normal SW

supply line, except for the automatic

isolation function and complete missile

and heavy load drop protection. The

design functions of the SW and CC

systems are unchanged as a result of the

proposed changes due to (1) required

plant conditions, (2) compensatory

measures, (3) a Contingency Action Plan

for restoration of the normal SW supply

if required, and (4) strict administrative

control of the temporary SW isolation

valve to preclude flooding or to isolate

non-essential SW within the design

basis assumed time limits. Unit 1 will

be in a plant condition which will

provide adequate time to restore the

normal SW supply, if required.

Therefore, since the SW and CC systems

will basically function as designed and

will be operated in their basic

configuration, the possibility of a new or

different type of accident than

previously evaluated in the UFSAR

[Updated Final Safety Analysis Report]

is not created.

3. Involve a significant reduction in a

margin of safety.

The margin of safety as defined in the

Technical Specifications is not reduced

since an operable SW flowpath to the

required number of CCHXs is provided,

and Unit operating constraints,

compensatory measures and

contingencies will be implemented as

required to ensure the integrity and the

capability of the SW flowpath. The use

of the temporary jumper will be limited

to the time period when missile

producing weather is not expected, and

Unit 1 meets specified unit conditions.

Therefore, the temporary SW jumper,

under the imposed project constraints

and compensatory measures, provides

the same reliability as the normal SW

supply line. Furthermore, the

Probabilistic Safety Assessment for

Surry Power Station has been reviewed

relative to the use of the temporary SW

jumper. It has been determined that due

to the SW restoration project’s

compensatory and contingency

measures, as well as the configuration

restrictions that will be imposed by the

Maintenance Rule online risk matrix,

the impact on core damage frequency is

negligible.
Federal Register / Vol. 63, No. 135 /Wednesday, July 15, 1998 / Notices 38207

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment request

involves no significant hazards

consideration.

Local Public Document Room

location: Swem Library, College of

William and Mary, Williamsburg,

Virginia 23185.

Attorney for licensee: Michael W.

Maupin, Esq., Hunton and Williams,

Riverfront Plaza, East Tower, 951 E.

Byrd Street, Richmond, Virginia 23219.

NRC Project Director: P. T. Kuo,

Acting.

Westinghouse Electric Corporation

(Licensee), Westinghouse Test Reactor,

Waltz Mill Site, Westmoreland,

Pennsylvania, Docket No. 50–22,

License No. TR–2

Date of amendment request:

December 22, 1997, supplemented on

June 15, 1998.

Description of amendment request: In

1959, the Westinghouse Electric

Corporation was granted a license for

the Westinghouse Test Reactor (WTR) at

Waltz Mill. On December 22, 1997, the

licensee informed the Nuclear

Regulatory Commission it had changed

its name to CBS Corporation, and

requested the license to be amended to

reflect the name change.

On June 15, 1998, the CBS

Corporation agreed that the name of the

WTR licensee, as reflected on the

license, can be revised to ‘‘CBS

Corporation acting through its

Westinghouse Electric Company

Division.’’ Therefore, the purpose of this

amendment is to change the name of the

licensee as indicated on the WTR

license from Westinghouse Electric

Corporation to CBS Corporation acting

through its Westinghouse Electric

Company Division.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration. The NRC staff has

reviewed the licensee’s analysis against

the standards of 10 CFR 50.92(c). A

proposed amendment to an operating

license for a facility involves no

significant hazards consideration if

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.

The staff agrees with the licensee’s no

significant hazards consideration

determination submitted on June 15,

1998 for the following reason.

This corporate name change does not

involve any change in the management,

organization, location, facilities

equipment, or procedures related to or

personnel responsible for the licensed

activities of the WTR license. All

existing commitments, obligations and

representations remain in effect.

Based on a review of the licensee’s

analysis, and on the staff’s analysis

detailed above, 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.

Attorney for license: Lisa A.

Campagna, Assistant General Counsel,

Law Department, CBS Corporation, P.O.

Box 355, Pittsburgh, Pennsylvania

15230.

NRC Project Director: Seymour H.

Weiss.

Wisconsin Electric Power Company,

Docket Nos. 50–266 and 50–301, Point

Beach Nuclear Plant, Units 1 and 2,

Town of Two Creeks, Manitowoc

County, Wisconsin

Date of amendment request: February

26, 1998 (TSCR 204).

Description of amendment request:

The proposed changes would modify

Technical Specifications (TS) and bases

to reflect a lower containment leakage

limit, a revised program for control of

primary coolant sources outside

containment, a revised control room

emergency filtration design, and the

addition of the primary auxiliary

building exhaust filtration system.

Basis for proposed no significant

hazards consideration determination:

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. Operation of this facility under the

proposed Technical Specifications will

not create a significant increase in the

probability or consequences of an

accident previously evaluated.

The probabilities of accidents

previously evaluated are based on the

probability of initiating events for these

accidents. Initiating events for accidents

previously evaluated for Point Beach

include: Control rod withdrawal and

drop, CVCS [chemical volume control

system] malfunction (Boron Dilution),

startup of an inactive reactor coolant

loop, reduction in feedwater enthalpy,

excessive load increase, losses of reactor

coolant flow, loss of external electrical

load, loss of normal feedwater, loss of

all AC [alternating current ] power to

the auxiliaries, turbine overspeed, fuel

handling accidents, accidental releases

of waste liquid or gas, steam generator

tube rupture, steam pipe rupture,

control rod ejection, and primary

coolant system ruptures.

This license amendment request

proposes to change the limiting

conditions for operation, action

statements, allowable outage times, and

surveillance requirements for the Point

Beach Nuclear Plant [PBNP] Technical

Specifications associated with the

maximum permissible containment leak

rate, control room emergency filtration,

primary auxiliary building exhaust

filtration, and primary coolant sources

outside containment. These proposed

changes do not cause an increase in the

probabilities of any accidents previously

evaluated because these changes will

not cause an increase in the probability

of any initiating events for accidents

previously evaluated. In particular,

these changes affect accident mitigation

systems and equipment which do not

cause accidents.

The consequences of the accidents

previously evaluated in the PBNP FSAR

[Final Safety Analysis Report] are

determined by the results of analyses

that are based on initial conditions of

the plant, the type of accident, transient

response of the plant, and the operation

and failure of equipment and systems.

The changes proposed in this license

amendment request provide appropriate

limiting conditions for operation, action

statements, allowable outage times, and

surveillance requirements for maximum

permissible containment leak rate,

control room emergency filtration,

primary auxiliary building exhaust

filtration, and primary coolant sources

outside containment.

The proposed changes affect

components that are required to ensure

the proper operation of accident

mitigation systems and equipment. The

proposed changes do not increase the

probability of failure of this equipment

or its ability to operate as required for

the accidents previously evaluated in

the PBNP FSAR.

Therefore, this proposed license

amendment does not affect the

consequences of any accident

previously evaluated in the Point Beach

Nuclear Plant FSAR, because the factors

that are used to determine the

consequences of accidents are not being

changed.

2. Operation of this facility under the

proposed Technical Specifications

change will not create the possibility of
38208 Federal Register / Vol. 63, No. 135 /Wednesday, July 15, 1998 / Notices

a new or different kind of accident from

any accident previously evaluated.

New or different kinds of accidents

can only be created by new or different

accident initiators or sequences. New

and different types of accidents

(different from those that were

originally analyzed for Point Beach)

have been evaluated and incorporated

into the licensing basis for Point Beach

Nuclear Plant. Examples of different

accidents that have been incorporated

into the Point Beach Licensing basis

include anticipated transients without

scram and station blackout. The changes

proposed by this license amendment

request do not create any new or

different accident initiators or

sequences because these changes to

limiting conditions for operation, action

statements, allowable outage times, and

surveillance requirements for maximum

permissible containment leak rate,

control room emergency filtration,

primary auxiliary building exhaust

filtration, and primary coolant sources

outside containment will not cause

failures of equipment or accident

sequences different than the accidents

previously evaluated. Therefore, these

proposed Technical Specifications

changes do not create the possibility of

an accident of a different type than any

previously evaluated in the Point Beach

FSAR.

3. Operation of this facility under the

proposed Technical Specifications

change will not create a significant

reduction in a margin of safety.

The margins of safety for Point Beach

are based on the design and operation

of the reactor and containment and the

safety systems that provide their

protection.

The changes proposed by this license

amendment request provide the

appropriate limiting conditions for

operation, action statements, allowable

outage times, and surveillance

requirements for maximum permissible

containment leak rate, control room

emergency filtration, primary auxiliary

building exhaust filtration, and primary

coolant sources outside containment.

This ensures that the safety systems that

protect the reactor and containment will

operate as required. The design and

operation of the reactor and

containment are not affected by these

proposed changes. Therefore, the

margins of safety for Point Beach are not

being reduced because the design and

operation of the reactor and

containment are not being changed and

the safety systems and limiting

conditions of operation for these safety

systems that provide their protection

that are being changed will continue to

meet the requirements for accident

mitigation for Point Beach Nuclear

Plant.

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.

Local Public Document Room

location: The Lester Public Library,

1001 Adams Street, Two Rivers,

Wisconsin 54241.

Attorney for licensee: John H. O’Neill,

Jr., Shaw, Pittman, Potts, and

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037.

NRC Project Director: Cynthia A.

Carpenter.

Wisconsin Electric Power Company,

Docket Nos. 50–266 and 50–301, Point

Beach Nuclear Plant, Units 1 and 2,

Town of Two Creeks, Manitowoc

County, Wisconsin

Date of amendment request: May 28,

1998 (TSCR 203).

Description of amendment request:

The proposed amendments would

revise Technical Specifications (TS) to

provide a specific numerical setting for

reactor trip, reactor coolant pump trip,

and auxiliary feedwater initiation on a

loss of power to the 4 kilovolt (kV)

buses. Changes to the bases for the

affected TS sections are also being

made.

Basis for proposed no significant

hazards consideration determination:

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. Operation of the Point Beach

Nuclear Plant [PBNP] in accordance

with the proposed amendments will not

create a significant increase in the

probability or consequences of an

accident previously evaluated.

The probabilities of accidents

previously evaluated are based on the

probability of initiating events for these

accidents. Initiating events for accidents

potentially affected by the proposed

amendments previously evaluated for

Point Beach include losses of reactor

coolant flow, loss of external electrical

load, loss of normal feedwater, and loss

of all AC [alternating current] power to

the auxiliaries.

This license amendment request

proposes to clarify the setting limit for

the undervoltage reactor trip, auxiliary

feedwater initiation and reactor coolant

pump trip by providing an actual

numerical value in place of the word

‘‘Normal’’ thereby eliminating any

confusion as to the actual value used in

the setting limit for this protection

function.

This proposed change does not cause

an increase in the probabilities of any

accidents previously evaluated because

the change will not cause an increase in

the probability of any initiating events

for accidents previously evaluated. In

particular, the proposed change more

clearly defines the actual setting limit

for the 4 KV undervoltage protection

function taking into account the effects

of voltage decay and response times.

This is a protection function for

mitigation of these events. Appropriate

delay times are implemented in this

function to ensure momentary voltage

transients do not initiate these events

while ensuring appropriate protection

for these loss of power events.

Therefore, there is no significant

increase in the probability or

consequences of any event previously

analyzed.

The consequences of the accidents

previously evaluated in the PBNP FSAR

[Final Safety Analysis Report] are

determined by the results of analyses

that are based on initial conditions of

the plant, the type of accident, transient

response of the plant, and the operation

and failure of equipment and systems.

The changes proposed in this license

amendment request provide appropriate

limiting conditions for the setting limits

for the Point Beach Nuclear Plant

Technical Specifications for the 4 KV

undervoltage protection function. Thus

the analyses of the events remain valid

and demonstrate that there are no

radiological consequences from these

events.

Therefore, this proposed license

amendment does not affect the

consequences of any accident

previously evaluated in the Point Beach

Nuclear Plant FSAR, because the factors

that are used to determine the

consequences of accidents are not being

changed.

2. Operation of the Point Beach

Nuclear Plant in accordance with the

proposed amendments will not create

the possibility of a new or different kind

of accident from any accident

previously evaluated.

New or different kinds of accidents

can only be created by new or different

accident initiators or sequences. New

and different types of accidents

(different from those that were

originally analyzed for Point Beach)

have been evaluated and incorporated

into the licensing basis for Point Beach

Nuclear Plant. Examples of different

accidents that have been incorporated

into the Point Beach Licensing basis

include anticipated transients without

scram and station blackout.
Federal Register / Vol. 63, No. 135 /Wednesday, July 15, 1998 / Notices 38209

The change proposed by the

amendments to provide specific

undervoltage setting limits does not

create any new or different accident

initiators or sequences because the

change to the 4 KV undervoltage

protection function will not cause

failures of equipment or accident

sequences different than the accidents

previously evaluated. Therefore, the

proposed Technical Specification

change does not create the possibility of

an accident of a different type than any

previously evaluated in the Point Beach

FSAR.

3. Operation of the Point Beach

Nuclear Plant in accordance with the

proposed amendments [will] not create

a significant reduction in a margin of

safety.

The margins of safety for Point Beach

are based on the design and operation

of the reactor and containment and the

safety systems that provide their

protection.

The change proposed by this license

amendment request provides the

appropriate setting limit for the 4 KV

undervoltage protection function. This

ensures that the safety systems that

protect the reactor and containment will

operate as required. The design and

operation of the reactor and

containment are not affected by these

proposed changes. Therefore, the

margins of safety for Point Beach are not

being reduced because the design and

operation of the reactor and

containment are not being changed.

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.

Local Public Document Room

location: The Lester Public Library,

1001 Adams Street, Two Rivers,

Wisconsin 54241.

Attorney for licensee: John H. O’Neill,

Jr., Shaw, Pittman, Potts, and

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037.

NRC Project Director: Cynthia A.

Carpenter.

Previously Published Notices of

Consideration of Issuance of

Amendments to Facility Operating

Licenses, Proposed No Significant

Hazards Consideration Determination,

and Opportunity for a Hearing

The following notices were previously

published as separate individual

notices. The notice content was the

same as above. They were published as

individual notices either because time

did not allow the Commission to wait

for this biweekly notice or because the

action involved exigent circumstances.

They are repeated here because the

biweekly notice lists all amendments

issued or proposed to be issued

involving no significant hazards

consideration.

For details, see the individual notice

in the Federal Register on the day and

page cited. This notice does not extend

the notice period of the original notice.

Consumers Energy Company, Docket

No. 50–255, Palisades Plant, Van Buren

County, Michigan

Date of amendment request: June 17,

1998, as supplemented June 23, 1998.

Brief description of amendment

request: The proposed amendment

would revise Section 3.1.1c of the

Technical Specifications (TS),

Appendix A of the Operating License

for the Palisades Nuclear Plant, to

change the minimum required primary

coolant system flow. The currently

specified value is 140.7×10 6 lb/hr

[pounds per hour] or greater, when

corrected to 532 °F. The licensee

proposed to revise the TS to specify a

value of greater than or equal to 352,000

gpm [gallons per minute], which is

equivalent to approximately 135×10 6

lb/hr, when corrected to 532 °F.

Date of publication of individual

notice in Federal Register: July 2, 1998

(63 FR 36271)

Expiration date of individual notice:

August 3, 1998.

Local Public Document Room

location: Van Wylen Library, Hope

College, Holland, Michigan 49423–3698.

Detroit Edison Company, Docket No.

50–341, Fermi 2, Monroe County,

Michigan

Date of amendment request: June 26,

1998 (NRC–98–0040).

Brief description of amendment

request: The proposed amendment

would provide a one-time extension of

the interval for a number of technical

specification (TS) surveillance

requirements that will be performed in

the sixth refueling outage. TS 4.0.2 and

Index page xxii would be revised and

TS tables 4.0.2–1 and 4.0.2–2 would be

replaced to reflect the extensions.

Date of publication of individual

notice in Federal Register: July 2, 1998

(63 FR 36273).

Expiration date of individual notice:

August 3, 1998.

Local Public Document Room

location: Monroe County Library

System, Ellis Reference and Information

Center, 3700 South Custer Road,

Monroe, Michigan 48161.

Duquesne Light Company, et al., Docket

Nos. 50–334 and 50–412, Beaver Valley

Power Station, Unit Nos. 1 and 2,

Shippingport, Pennsylvania

Date of amendment request: June 19,

1998.

Brief description of amendment

request: This amendment revises the

Beaver Valley Power Station, Units 1

and 2 (BVPS–1 and BVPS–2), Technical

Specifications (TS) definition of a

channel calibration to add two

sentences stating that (1) the calibration

of instrument channels with resistance

temperature detector or thermocouple

sensors may consist of an inplace

qualitative assessment of sensor

behavior and normal calibration of the

remaining adjustable devices in the

channel and (2) whenever a sensing

element is replaced, the next required

channel calibration shall include an

inplace cross calibration that compares

the other sensing elements with the

recently installed sensing element. This

proposed change would make the

BVPS–1 and BVPS–2 TS definition of

channel calibration consistent with the

definition of a channel calibration

contained in the NRC’s improved

Standard Technical Specifications for

Westinghouse Plants (NUREG–1431,

Revision 1).

Date of publication of individual

notice in Federal Register: June 26,

1998.

Expiration date of individual notice:

July 27, 1998 (63 FR 34939).

Local Public Document Room

location: B. F. Jones Memorial Library,

663 Franklin Avenue, Aliquippa, PA

15001.

Pennsylvania Power and Light

Company, Docket No. 50–388,

Susquehanna Steam Electric Station,

Unit 2, Luzerne County, Pennsylvania

Date of amendment request: June 17,

1998.

Brief description of amendment

request: This amendment revises the

applicability requirement in TS Sections

3.4.2, ‘‘Safety/Relief Valves’’ (Action c);

4.4.2; 3.3.7.5, ‘‘Accident Monitoring

Instrumentation’’ (TS Table 3.3.7.5–1,

Action 80 and 4.3.7.5, ‘‘Surveillance

Requirements,’’ Table 4.3.7.5–1

‘‘Accident Monitoring Instrumentation

Surveillance Requirements’’). The

change to the referenced TSs adds the

following applicability footnote:

Compliance with these requirements for

the ‘‘J’’ SRV acoustic monitor is not required

for the period beginning June 15, 1998, until

the next unit shutdown of sufficient duration

to allow for containment entry, not to exceed

the 9th refueling and inspection outage.
38210 Federal Register / Vol. 63, No. 135 /Wednesday, July 15, 1998 / Notices

Date of publication of individual

notice in Federal Register: June 23, 1998

(63 FR 34200).

Expiration date of individual notice:

July 23, 1998.

Local Public Document Room

location: Osterhout Free Library,

Reference Department, 71 South

Franklin Street, Wilkes-Barre, PA 18701.

Notice of Issuance of Amendments to

Facility Operating Licenses

During the period since publication of

the last biweekly notice, the

Commission has issued the following

amendments. The Commission has

determined for each of these

amendments that the application

complies with the standards and

requirements of the Atomic Energy Act

of 1954, as amended (the Act), and the

Commission’s rules and regulations.

The Commission has made appropriate

findings as required by the Act and the

Commission’s rules and regulations in

10 CFR Chapter I, which are set forth in

the license amendment.

Notice of Consideration of Issuance of

Amendment to Facility Operating

License, Proposed No Significant

Hazards Consideration Determination,

and Opportunity for A Hearing in

connection with these actions was

published in the Federal Register as

indicated.

Unless otherwise indicated, the

Commission has determined that these

amendments satisfy the criteria for

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 these

amendments. If the Commission has

prepared an environmental assessment

under the special circumstances

provision in 10 CFR 51.12(b) and has

made a determination based on that

assessment, it is so indicated.

For further details with respect to the

action see (1) the applications for

amendment, (2) the amendment, and (3)

the Commission’s related letter, Safety

Evaluation and/or Environmental

Assessment as indicated. All of 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 rooms for the

particular facilities involved.

Commonwealth Edison Company,

Docket Nos. 50–373 and 50–374, LaSalle

County Station, Units 1 and 2, LaSalle

County, Illinois

Date of application for amendments:

May 27, 1997, as supplemented on

August 1, 1997, and March 24, 1998.

Brief description of amendments: The

amendments revise Technical

Specification Section 6, ‘‘Administrative

Controls,’’ to incorporate revised

organizational titles and delete Unit 1

Facility Operating License Condition

2.C.(30)(a). In addition, the amendments

change the submittal frequency of the

Radiological Effluent Release Report

from semiannually to annually and

make several administrative and

editorial changes.

Date of issuance: June 26, 1998.

Effective date: Immediately, to be

implemented within 90 days.

Amendment Nos.: 128 and 113.

Facility Operating License Nos. NPF–

11 and NPF–18: The amendments

revised the Unit 1 Facility Operating

License and the Technical

Specifications.

Date of initial notice in Federal

Register: July 30 ,1997. The August 1,

1997, submittal provided clarifying

information that did not change the

initial proposed no significant hazards

consideration determination. The March

24, 1998, submittal changed the scope of

the initial Federal Register notice. The

proposed amendments were renoticed

on May 20, 1998 (63 FR 27759).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated June 26, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Jacobs Memorial Library,

Illinois Valley Community College,

Oglesby, Illinois 61348.

Commonwealth Edison Company,

Docket No. 50–373, LaSalle County

Station, Unit 1, LaSalle County, Illinois

Date of application for amendment:

November 24, 1997, as supplemented

April 16, 1998.

Brief description of amendment: The

amendment revises Technical

Specification 3/4 3.2, ‘‘Isolation

Actuation Instrumentation’’ to add/

revise various isolation setpoints for

leak detection instrumentation. These

changes are necessary due to

modifications to the Reactor Water

Cleanup (RWCU) system to restore

‘‘hot’’ suction to the RWCU pumps and

due to a re-evaluation of the high energy

line break analysis. In addition, the

amendment eliminates isolation

actuation trip functions for the Residual

Heat Removal system steam condensing

mode and shutdown cooling mode.

Date of issuance: July 6, 1998.

Effective date: Immediately, to be

implemented prior to restart from L1F35

Amendment No.: 129.

Facility Operating License No. NPF–

11: The amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: January 14, 1998 (63 FR 2278).

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated July 6, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Jacobs Memorial Library,

Illinois Valley Community College,

Oglesby, Illinois 61348

Connecticut Yankee Atomic Power

Company, Docket No. 50–213, Haddam

Neck Plant, Middlesex County,

Connecticut

Date of application of amendments:

May 30, 1997, as supplemented May 7,

1998 and June 18, 1998.

Brief description of amendments: The

amendments revise the Facility

Operating License and Technical

Specifications to reflect the permanently

shut down and defueled status of the

reactor.

Date of issuance: June 30, 1998.

Effective date: As of the date of

issuance (June 30, 1998) and shall be

implemented within 90 days.

Amendment No.: 193.

Facility Operating License No. DPR–

61: The amendments revised the

Operating License and the Technical

Specifications.

Date of initial notice in Federal

Register: July 16, 1997 (62 FR 38132 and

62 FR 38133). The May 7, 1998,

supplement relocated the provisions of

Technical Specification 3/4.9.15. The

June 18, 1998, supplement consisted of

supporting technical information. The

supplements did not change the staff’s

initial proposed no significant hazards

consideration determination or expand

the scope of the original notice. The

Commission’s related evaluation of the

amendment is contained in a Safety

Evaluation dated June 30, 1998.

No significant hazards consideration

received: No.

Local Public Document Room

location: Russell Library, 123 Broad

Street, Middletown, Connecticut 06457.

Detroit Edison Company, Docket No.

50–16, Enrico Fermi Atomic Power

Plant, Unit 1, Monroe County, Michigan

Date of amendment request: January

27, 1998 (Reference NRC–98–0023).

Brief description of amendment: This

amendment revises the Fermi 1 License

to allow Detroit Edison to receive,

acquire, possess, use, and transfer

byproduct material without restriction

to chemical form for sample analysis,

instrument calibration, or associated
Federal Register / Vol. 63, No. 135 /Wednesday, July 15, 1998 / Notices 38211

with radioactive apparatus, hardware,

tools, and equipment, provided the

cumulative radioactive material

quantity of the byproduct material does

not exceed the criteria contained in

Section 30.72, Schedule C, Quantities of

Radioactive Materials Requiring

Consideration of the Need for an

Emergency Plan for Responding to a

Release.

Date of issuance: June 22, 1998.

Effective date: Within 60 calendar

days from the date of issuance of this

amendment.

Amendment No.: 12.

Facility Operating License No. DPR–9:

Amendment revised License by adding

a subpart 3 to Part 2.B.

Date of initial notice in Federal

Register: April 8, 1998 (63 FR 17223).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated June 22, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Monroe County Library

System, 3700 South Custer Road,

Monroe, Michigan 48161.

Duke Energy Corporation, Docket Nos.

50–269 and 50–287, Oconee Nuclear

Station, Units 1 and 3, Seneca, South

Carolina

Date of application of amendments:

June 4, 1998.

Brief description of amendments: The

amendments revise Technical

Specification 4.17.2 to allow continued

operation with certain steam generator

tubes that exceed their repair limit as a

result of tube end anomalies. This

action temporarily exempts these tubes

from the requirement for sleeving,

rerolling, or removal from service until

they are repaired during or before the

next scheduled refueling outages for the

respective unit. This action supersedes

the Notice of Enforcement Discretion

that was issued by the staff on June 4,

1998.

Date of Issuance: July 1, 1998.

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment Nos.: Unit 1—230; Unit

2—227.

Facility Operating License Nos. DPR–

38 and DPR–55: The amendments

revised the Technical Specifications.

Public comments requested as to

proposed no significant hazards

consideration: Yes. (63 FR 33097 dated

June 17, 1998). The notice provided an

opportunity to submit comments on the

Commission’s proposed no significant

hazards consideration determination.

No comments have been received. The

notice also provided for an opportunity

to request a hearing by July 16, 1998,

but indicated that if the Commission

makes a final no significant hazards

consideration determination, any such

hearing would take place after issuance

of the amendments.

The Commission’s related evaluation

of the amendments, finding of exigent

circumstances, and a final no significant

hazards consideration determination are

contained in a Safety Evaluation dated

July 1, 1998.

Local Public Document Room

location: Oconee County Library, 501

West South Broad Street, Walhalla,

South Carolina.

Entergy Operations, Inc., Docket No. 50–

382, Waterford Steam Electric Station,

Unit 3, St. Charles Parish, Louisiana

Date of amendment request: June 3,

1997, as supplemented by letter dated

May 1, 1998.

Brief description of amendment: The

amendment changed the Appendix A

Technical Specifications (TSs) by

changing the action requirements for TS

3/4.3.2 for the Safety Injection System

Sump Recirculation Actuation Signal

(RAS). It revised the allowed outage

time for a channel of RAS to be in the

tripped condition from ‘‘prior to entry

into the applicable MODE(S) following

the next COLD SHUTDOWN’’ to the

more restrictive time limit of 48 hours,

and added a shutdown requirement.

Additionally, the TS 3.0.4 exemption

was removed from the action for the

tripped condition. A change to TS Bases

Section 3/4.3.2 was also included.

Date of issuance: July 2, 1998.

Effective date: July 2, 1998, to be

implemented within 60 days.

Amendment No.: 143.

Facility Operating License No. NPF–

38: Amendment revised the Technical

Specifications. Date of initial notice in

Federal Register: June 18, 1997 (62 FR

33124).

The additional information contained

in the supplemental letter dated May 1,

1998, was clarifying in nature and thus,

it was within the scope of the initial

notice and did not affect the staff’s

proposed no significant hazards

consideration determination.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated July 2, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: University of New Orleans

Library, Louisiana Collection, Lakefront,

New Orleans, LA 70122.

North Atlantic Energy Service

Corporation, et al., Docket No. 50–443,

Seabrook Station, Unit No. 1,

Rockingham County, New Hampshire

Date of amendment request: March 2,

1998, as supplemented by letter dated

April 21, 1998.

Description of amendment request:

The amendment revised Technical

Specification 4.5.2.b.1 for the

emergency core cooling system

subsystems to delete the requirement to

vent the operating chemical volume and

control system centrifugal pump casing.

Date of issuance: June 24, 1998.

Effective date: As of its date of

issuance, to be implemented within 60

days.

Amendment No.: 58.

Facility Operating License No. NPF–

86. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: April 8, 1998 (63 FR 17225).

The supplemental letter provided

clarifying information that did not

change the staff’s proposed no

significant hazards determination.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated June 24, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Exeter Public Library,

Founders Park, Exeter, NH 03833.

Northeast Nuclear Energy Company, et

al., Docket No. 50–336, Millstone

Nuclear Power Station, Unit No. 2, New

London County, Connecticut

Date of application for amendment:

April 13, 1998.

Brief description of amendment: The

amendment changes the Technical

Specifications (TSs) by adding a new TS

3.5.5, ‘‘Emergency Core Cooling

Systems—Trisodium Phosphate (TSP).’’

The TSP surveillance requirements in

TSs 4.5.2.c.3 and 4.5.2.c.4 are relocated

to new TS 3.5.5 as TS 4.5.5.1 and TS

4.5.5.2, respectively. Also, the amount

of TSP is increased, the surveillance

requirements are modified, a new

limiting condition of operation is

included, and the applicable TS Index

pages and Bases sections are updated to

reflect the changes.

Date of issuance: June 22, 1998.

Effective date: As of the date of

issuance to be implemented within 60

days.

Amendment No.: 217.

Facility Operating License No. DPR–

65: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: May 6, 1998 (63 FR 25114).
38212 Federal Register / Vol. 63, No. 135 /Wednesday, July 15, 1998 / Notices

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated June 22, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Learning Resources Center,

Three Rivers Community-Technical

College, 574 New London Turnpike,

Norwich, Connecticut, and the

Waterford Library, ATTN: Vince

Juliano, 49 Rope Ferry Road, Waterford,

Connecticut.

Northeast Nuclear Energy Company, et

al., Docket No. 50–336, Millstone

Nuclear Power Station, Unit No. 2, New

London County, Connecticut

Date of application for amendment:

November 13, 1997, as supplemented on

December 29, 1997, and April 8, 1998.

Brief description of amendment: The

amendment changes the Technical

Specifications (TSs) by modifying TS

3.1.2.1, ‘‘Flow Paths—Shutdown;’’ TS

3.1.2.2, ‘‘Flow Paths—Operating;’’ TS

3.1.2.3, ‘‘Charging Pump—Shutdown;’’

TS 3.1.2.4, ‘‘Charging Pumps—

Operating;’’ TS 3.1.2.5, ‘‘Boric Acid

Pumps—Shutdown;’’ TS 3.1.2.6, ‘‘Boric

Acid Pumps—Operating;’’ TS 3.1.2.8,

‘‘Borated Water Sources—Operating;’’

TS 3.4.1.3, ‘‘Coolant Loops and Coolant

Circulation—Shutdown;’’ TS 3.4.3,

‘‘Relief Valves;’’ TS 3.4.9.1, ‘‘Reactor

Coolant System;’’ TS 3.4.9.2,

‘‘Pressurizer;’’ TS 3.4.9.3, ‘‘Overpressure

Protection Systems;’’ TS 3.5.3, ‘‘ECCS

Subsystems—Tavg < 300 °F;’’ and TS

3.10.3, ‘‘Pressure/Temperature

Limitation—Reactor Criticality,’’ and

their associated Bases in the areas that

are affected by the modified Low

Temperature Overpressure Protection

system, the updated reactor coolant

system pressure and temperature curves

and heatup and cooldown limits.

Additionally, minor changes are made

to correct various items, such as,

updating of redundant or outdated TSs.

Date of issuance: July 1, 1998.

Effective date: As of the date of

issuance to be implemented within 60

days.

Amendment No.: 218.

Facility Operating License No. DPR–

65: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: January 28, 1998 (63 FR 4315).

The December 29, 1997, and April 8,

1998, letters provided clarifying

information that did not change the

scope of the November 13, 1997,

application and the initial proposed no

significant hazards consideration

determination.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated July 1, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Learning Resources Center,

Three Rivers Community-Technical

College, 574 New London Turnpike,

Norwich, Connecticut, and the

Waterford Library, ATTN: Vince

Juliano, 49 Rope Ferry Road, Waterford,

Connecticut.

Pacific Gas and Electric Company,

Docket Nos. 50–275 and 50–323, Diablo

Canyon Nuclear Power Plant, Unit Nos.

1 and 2, San Luis Obispo County,

California

Date of application for amendments:

February 27, 1997, as supplemented by

letter dated December 4, 1997.

Brief description of amendments: The

amendments revised the combined

Technical Specifications (TS) for the

Diablo Canyon Power Plant, (DCPP)

Unit Nos. 1 and 2 to change Technical

Specification (TS) 3/4.8.1.1, ‘‘A.C.

Sources—Operating,’’ to clarify that

emergency diesel generator (EDG)

testing is initiated from standby

conditions rather than ‘‘ambient’’

conditions. The associated TS Bases

were revised to discuss the temperature

range that satisfies EDG standby

conditions. TS 3/4.3.2,

‘‘Instrumentation—Engineering Safety

Features Actuation System

Instrumentation’’ was also changed.

This revision clarified that when one or

both of the first level load shed relays,

or one or both of the second level

undervoltage relays are inoperable, the

associated EDG for that bus shall be

declared inoperable.

Date of issuance: June 5, 1998.

Effective date: June 5, 1998, to be

implemented within 90 days from the

date of issuance.

Amendment Nos.: Unit 1—127; Unit

2—125

Facility Operating License Nos. DPR–

80 and DPR–82: The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: April 9, 1997 (62 FR 17240).

The December 4, 1997, supplemental

letter provided additional clarifying

information and did not change the

staff’s initial no significant hazards

consideration determination. The

Commission’s related evaluation of the

amendments is contained in a Safety

Evaluation dated June 5, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: California Polytechnic State

University, Robert E. Kennedy Library,

Government Documents and Maps

Department, San Luis Obispo, California

93407.

Pacific Gas and Electric Company,

Docket Nos. 50–275 and 50–323, Diablo

Canyon Nuclear Power Plant, Unit Nos.

1 and 2, San Luis Obispo County,

California

Date of application for amendments:

April 10, 1998, as supplemented by

letter dated May 1, 1998.

Brief description of amendments: The

amendments revised the combined

Technical Specifications (TS) for the

Diablo Canyon Power Plant (DCPP)

Units 1 and 2 to revise TS 6.2.2.g and

TS 6.3 to change the name of the

Operations Manager to Operations

Director, to add the position of

Operations Middle Manager, and to

change the requirement for the

Operations Director to hold a senior

reactor operator (SRO) license.

Date of issuance: June 11, 1998.

Effective date: June 11, 1998.

Amendment Nos.: Unit 1–128; Unit

2–126.

Facility Operating License Nos. DPR–

80 and DPR–82: The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: May 6, 1998 (63 FR 25116).

The May 1, 1998, supplemental letter

provided additional clarifying

information and did not change the

staff’s initial no significant hazards

consideration determination. The

Commission’s related evaluation of the

amendments is contained in a Safety

Evaluation dated June 11, 1998.

No significant hazards consideration

comments received: Yes.

The Commission received one letter

with comments which did not change

its finding and conclusion as discussed

in the safety evaluation.

Local Public Document Room

location: California Polytechnic State

University, Robert E. Kennedy Library,

Government Documents and Maps

Department, San Luis Obispo, California

93407.

Pennsylvania Power and Light

Company, Docket Nos. 50–387 and 50–

388, Susquehanna Steam Electric

Station, Units 1 and 2, Luzerne County,

Pennsylvania

Date of application for amendments:

May 12, 1998

Brief description of amendments:

These amendments relocate certain

requirements related to fire protection

from the TSs to the Updated Final

Safety Analysis Report. The TS sections

to be relocated are: 3/4.3.7.9, Fire

Detection Instrumentation; 3/4.7.6, Fire

Suppression Systems; 3/4.7.7, Fire
Federal Register / Vol. 63, No. 135 /Wednesday, July 15, 1998 / Notices 38213

Rated Assemblies; and 6.2.2e, Fire

Brigade Staffing. The amendments also

replace License Condition 2.C.(6) for

Unit 1 and License Condition 2.C.(3) for

Unit 2. These amendments are

consistent with the guidance of NRC

Generic Letter (GL) 86–10,

‘‘Implementation of Fire Protection

Requirements,’’ and GL 88–12,

‘‘Removal of Fire Protection

Requirements from Technical

Specifications.’’

Date of issuance: June 24, 1998.

Effective date: Both units, as of date

of issuance, to be implemented within

30 days.

Amendment Nos.: 177 and 150.

Facility Operating License Nos. NPF–

14 and NPF–22: The amendments

revised the Technical Specifications and

Licenses.

Date of initial notice in Federal

Register: May 21, 1998 (63 FR 28010).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated June 24, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Osterhout Free Library,

Reference Department, 71 South

Franklin Street, Wilkes-Barre, PA 18701.

Power Authority of the State of New

York, Docket No. 50–333, James A.

FitzPatrick Nuclear Power Plant,

Oswego County, New York

Date of application for amendment:

November 26, 1997, as supplemented

April 17, 1998.

Brief description of amendment: The

amendment relocates snubber

operability, surveillance, and records

requirements from the Technical

Specifications to plant controlled

documents. A condition is added to the

license to require that the relocated

requirements be described in the Final

Safety Analysis Report such that 10 CFR

50.59 will apply to future changes to

those requirements.

Date of issuance: June 30, 1998.

Effective date: As of the date of

issuance to be implemented within 90

days.

Amendment No.: 243.

Facility Operating License No. DPR–

59: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: January 29, 1997 (62 FR 4352).

The April 17, 1998, supplemental

letter provided clarifying information

that did not change the initial proposed

no significant hazards consideration.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated June 30, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Reference and Documents

Department, Penfield Library, State

University of New York, Oswego, New

York 13126.

Southern California Edison Company, et

al., Docket Nos. 50–361 and 50–362,

San Onofre Nuclear Generating Station,

Unit Nos. 2 and 3, San Diego County,

California

Date of application for amendments:

November 6, 1995, as supplemented by

letters dated January 9, 1998, and

February 3, 1998, for the safety injection

tanks (SITs), and November 8, 1995, as

supplemented by letters dated January

9, 1998, and February 3, 1998, for the

low pressure safety injection (LPSI).

Brief description of amendments: The

amendments modify the technical

specifications (TSs) to extend the

allowed outage times (AOTs) for a single

inoperable SIT from one hour to 24

hours, and for a single inoperable SIT

specifically due to malfunctioning SIT

water level or nitrogen cover pressure

instrumentation inoperability from one

hour to 72 hours. In addition, the

amendments extend the AOT for a

single inoperable LPSI train from 72

hours to 7 days. The amendments also

add a Configuration Risk Management

Program to the TSs that puts a

proceduralized probabilistic risk

assessment-informed process in place

that ensures the licensee assesses the

overall impact of plant maintenance on

plant risk.

Date of issuance: June 19, 1998.

Effective date: June 19, 1998, to be

implemented within 30 days from the

date of issuance.

Amendment Nos.: Unit 2—139; Unit

3—131.

Facility Operating License Nos. NPF–

10 and NPF–15: The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: April 10, 1996 (61 FR 15995)

and February 11, 1998 (63 FR 6991).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated June 19, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Main Library, University of

California, P. O. Box 19557, Irvine,

California 92713.

Southern Nuclear Power Company, Inc.,

et al., Docket Nos. 50–424 and 50–425,

Vogtle Electric Generating Plant, Units 1

and 2, Burke County, Georgia

Date of application for amendments:

September 4, 1997, as supplemented by

letters dated November 20, 1997, May

19 and June 12, 1998.

Brief description of amendments: The

changes to the common Technical

Specifications allow an increase in the

Unit 1 spent fuel storage capacity from

288 to 1476 fuel assemblies.

Date of issuance: June 29, 1998.

Effective date: As of the date of

issuance to be implemented on a

schedule consistent with the receipt and

storage of new fuel in the fall of 1998

for the spring 1999 refueling outage of

Unit 1.

Amendment Nos.: Unit 1—102; Unit

2—80.

Facility Operating License Nos. NPF–

68 and NPF–81: Amendments revised

the Technical Specifications, Operating

Licenses, and Appendix D.

Date of initial notice in Federal

Register: December 31, 1997 (62 FR

68317); and renoticed on May 11, 1998

(63 FR 25883).

The supplements dated May 19 and

June 12, 1998, provided clarifying

information that did not change the

scope of the September 4, 1997,

application and the initial proposed no

significant hazards consideration

determination.

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated June 29, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Burke County Library, 412

Fourth Street, Waynesboro, Georgia.

Tennessee Valley Authority, Docket

Nos. 50–327 and 50–328, Sequoyah

Nuclear Plant, Units 1 and 2, Hamilton

County, Tennessee

Date of application for amendments:

June 7, 1996, as supplemented on

September 26, 1997, January 21, 1998,

May 28, 1998, and June 29, 1998 (TS

95–19).

Brief description of amendments: The

amendments change the Technical

Specifications (TS) by relocating

portions of Section 6, ‘‘Administrative

Controls,’’ to the Sequoyah Nuclear

Quality Assurance Plan. This Change is

consistent with NUREG–1431,

‘‘Standard Technical Specifications—

Westinghouse Plants.’’

Date of issuance: July 1, 1998.

Effective date: July 1, 1998.

Amendment Nos.: 233 and 223.

Facility Operating License Nos. DPR–

77 and DPR–79: Amendments revise the

TS.

Date of initial notice in Federal

Register: July 17, 1996 (61 FR 37302).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated July 1, 1998.

No significant hazards consideration

comments received: None.
38214 Federal Register / Vol. 63, No. 135 /Wednesday, July 15, 1998 / Notices

Local Public Document Room

location: Chattanooga-Hamilton County

Library, 1001 Broad Street, Chattanooga,

Tennessee 37402.

Toledo Edison Company, Centerior

Service Company, and The Cleveland

Electric Illuminating Company, Docket

No. 50–346, Davis-Besse Nuclear Power

Station, Unit 1, Ottawa County, Ohio

Date of application for amendment:

December 23, 1997, as supplemented by

letter dated June 11, 1998.

Brief description of amendment: This

amendment revises Technical

Specification (TS) Section 1.0,

‘‘Definitions,’’ to clarify the meaning of

core alteration; relocates TS Section 3/

4.9.5, ‘‘Refueling Operations—

Communications,’’ and the associated

bases to the Technical Requirements

Manual; and adds TS Section 3.0.6 and

associated bases to address the return to

service of inoperable equipment.

Date of issuance: June 30, 1998.

Effective date: June 30, 1998.

Amendment No.: 224.

Facility Operating License No. NPF–3:

Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: January 28, 1998 (63 FR 4327).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated June 30, 1998.

.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: University of Toledo, William

Carlson Library, Government

Documents Collection, 2801 West

Bancroft Avenue, Toledo, OH 43606.

Virginia Electric and Power Company, et

al., Docket Nos. 50–338 and 50–339,

North Anna Power Station, Units No. 1

and No. 2, Louisa County, Virginia

Date of application for amendments:

March 25, 1998.

Brief description of amendments: The

amendments revise the Technical

Specifications (TS) Sections 6.1.1;

6.2.1.b; 6.5.1.1; 6.5.1.6.a,d,h, and m;

6.5.1.7.c; 6.5.1.8; 6.14.1.2; 6.15.b;

6.2.3.5; 6.5.1.2; 6.5.1.7.a for Unit 1 and

6.1.1; 6.2.1.b; 6.5.1.1; 6.5.1.6.a,d,h, and

m; 6.5.1.7.c; 6.5.1.8; 6.13.b; 6.14.b;

6.2.3.5; 6.5.1.2; and 6.5.1.7.a for Unit 2,

changing the title of Station Manager to

Site Vice President, and the titles of the

Assistant Station Managers to ManagerStation Operation and Maintenance and

Manager-Station Safety and Licensing.

Date of issuance: June 23, 1998.

Effective date: June 23, 1998.

Amendment Nos.: 212 and 193.

Facility Operating License Nos. NPF–

4 and NPF–7: Amendments revised the

Technical Specifications.

Date of initial notice in Federal

Register: April 22, 1998 (63 FR 19980).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated June 23, 1998.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: The Alderman Library, Special

Collections Department, University of

Virginia, Charlottesville, Virginia

22903–2498.

Dated at Rockville, Maryland, this 8th day

of July 1998.

For the Nuclear Regulatory Commission.

Elinor G. Adensam,

Acting Director, Division of Reactor Projects—

III/IV, Office of Nuclear Reactor Regulation.

[FR Doc. 98–18684 Filed 7–14–98; 8:45 am]

BILLING CODE 7590–01–P

PENSION BENEFIT GUARANTY

CORPORATION

Interest Assumption for Determining

Variable-Rate Premium; Interest on

Late Premium Payments; Interest on

Underpayments and Overpayments of

Single-Employer Plan Termination

Liability and Multiemployer Withdrawal

Liability; Interest Assumptions for

Multiemployer Plan Valuations

Following Mass Withdrawal

AGENCY: Pension Benefit Guaranty

Corporation.

ACTION: Notice of interest rates and

assumptions.

SUMMARY: This notice informs the public

of the interest rates and assumptions to

be used under certain Pension Benefit

Guaranty Corporation regulations. These

rates and assumptions are published

elsewhere (or are derivable from rates

published elsewhere), but are collected

and published in this notice for the

convenience of the public. Interest rates

are also published on the PBGC’s web

site (http://www.pbgc.gov).

DATES: The interest rate for determining

the variable-rate premium under part

4006 applies to premium payment years

beginning in July 1998. The interest

assumptions for performing

multiemployer plan valuations

following mass withdrawal under part

4281 apply to valuation dates occurring

in August 1998. The interest rates for

late premium payments under part 4007

and for underpayments and

overpayments of single-employer plan

termination liability under part 4062

and multiemployer withdrawal liability

under part 4219 apply to interest

accruing during the third quarter (July

through September) of 1998.

FOR FURTHER INFORMATION CONTACT:

Harold J. Ashner, Assistant General

Counsel, Office of the General Counsel,

Pension Benefit Guaranty Corporation,

1200 K Street, NW., Washington, DC

20005, 202–326–4024. (For TTY/TDD

users, call the Federal relay service tollfree at 1–800–877–8339 and ask to be

connected to 202–326–4024.)

SUPPLEMENTARY INFORMATION:

Variable-Rate Premiums

Section 4006(a)(3)(E)(iii)(II) of the

Employee Retirement Income Security

Act of 1974 (ERISA) and § 4006.4(b)(1)

of the PBGC’s regulation on Premium

Rates (29 CFR part 4006) prescribe use

of an assumed interest rate in

determining a single-employer plan’s

variable-rate premium. The rate is the

‘‘applicable percentage’’ (described in

the statute and the regulation) of the

annual yield on 30-year Treasury

securities for the month preceding the

beginning of the plan year for which

premiums are being paid (the ‘‘premium

payment year’’). The yield figure is

reported in Federal Reserve Statistical

Releases G.13 and H.15.

For plan years beginning before July

1, 1997, the applicable percentage of the

30-year Treasury yield was 80 percent.

The Retirement Protection Act of 1994

(RPA) amended ERISA section

4006(a)(3)(E)(iii)(II) to change the

applicable percentage to 85 percent,

effective for plan years beginning on or

after July 1, 1997. (The amendment also

provides for a further increase in the

applicable percentage ‘‘ to 100 percent

‘‘ when the Internal Revenue Service

adopts new mortality tables for

determining current liability.)

The assumed interest rate to be used

in determining variable-rate premiums

for premium payment years beginning

in July 1998 is 4.85 percent (i.e., 85

percent of the 5.70 percent yield figure

for June 1998).

(Under section 774(c) of the RPA, the

amendment to the applicable percentage

was deferred for certain regulated public

utility (RPU) plans for as long as six

months. The applicable percentage for

RPU plans has therefore remained 80

percent for plan years beginning before

January 1, 1998. For ‘‘partial’’ RPU

plans, the assumed interest rates to be

used in determining variable-rate

premiums can be computed by applying

the rules in § 4006.5(g) of the premium

rates regulation. The PBGC’s 1997

premium payment instruction booklet

also describes these rules and provides

a worksheet for computing the assumed

rate.)

The following table lists the assumed

interest rates to be used in determining

variable-rate premiums for premium