Federal Register / Vol. 62, No. 250 / Wednesday, December 31, 1997 / Notices 68303

Description of Permit Modification

Requested

1. On December 27, 1996, the

National Science Foundation issued a

permit (97WM–4) to Dr. Rennie S. Holt

at the National Oceanic and

Atmospheric Administration’s (NOAA)

Antarctic Marine Living Resources

(AMLR) Program after posting a notice

in the November 21, 1996 Federal

Register. Public comments were not

received. The issued permit was for the

use and release of designated pollutants

associated with the construction and

operation of a research field camp at

Camp Shirreff, Livingston Island,

Antarctica (62°28′S60°47′W). During the

first season at Cape Shirreff, only

limited research activities were

conducted as most of the effort was

focused on camp construction. In the

coming seasons, the AMLR Program

proposes to expand research activities,

providing a more comprehensive

research program. One project of this

expanded program proposes to use the

doubly labeled water (tritiated and

oxygen-18) method to measure the freeranging foraging energetics of Antarctic

fur seals (Arctocephalus gazella). Use of

tritium labeled water was not included

in the original permit request. The scope

of this application for a permit

modification pertains to waste

management issues involved with the

use and handling of the radioactive

isotope tritium. The duration of the

requested modification is coincident

with the current permit which expires

on April 30, 2001.

All radioisotope materials will be

handled only by researchers trained in

their proper handling and use. For each

season it is anticipated that

approximately 55 mCi 3H2O will be

used for research purposes. All wastes

generated from the research activities

will be double bagged, packaged in

appropriate containers lined with

absorbent pads, and will be returned to

the University of California

Environmental Health and Safety Office,

Santa Cruz for disposal. Conditions of

the permit modification would include

an annual report of all activities

involving the tritium and a declaration

by the institutional radiation safety

officer that all materials returned from

the Antarctic have been received.

Joyce A. Jatko,

Acting Permit Officer.

[FR Doc. 97–34038 Filed 12–30–97; 8:45 am]

BILLING CODE 7555–01–M

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 December 6,

1997, through December 18, 1997. The

last biweekly notice was published on

December 17, 1997 (62 FR 66133).

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 Freedom

of Information and Publications

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 January 30, 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
68304 Federal Register / Vol. 62, No. 250 / Wednesday, December 31, 1997 / Notices

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.

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.

Carolina Power & Light Company, et al.

[Docket Nos. 50–325 and 50–324]

Brunswick Steam Electric Plant, Units 1

and 2, Brunswick County, North

Carolina

Date of amendments request:

November 26, 1997.

Description of amendments request:

Carolina Power & Light Company

(CP&L) has proposed amendments to the

Technical Specifications (TS) for the

Brunswick Steam Electric Plant Units 1

and 2 (BSEP 1 & 2) to revise certain

instrumentation allowable values. The

revised values were calculated using a

methodology and format consistent with

that provided in NUREG–1433, Revision

1, ‘‘Standard Technical Specifications

General Electric Plants, BWR/4.’’ The

current TS are based on the uncertainty

associated with the trip unit portion of

the instrumentation circuitry. The

proposed values are based on the

uncertainty associated with the entire

instrumentation loop (sensor and trip

unit). The NRC has previously approved

this methodology for BSEP 1 & 2 as part

of a 5 percent power uprate amendment

dated November 1, 1996.

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 amendments do not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed changes affect accident

mitigation instrumentation allowable values.

The changes will not affect the accident

mitigation instrumentation functions. No

changes will occur in the way in which

equipment is operated. Therefore, the

probability of a previously evaluated

accident can not be affected.

The proposed changes establish the

allowable values for certain functions in

accordance with the CP&L setpoint

methodology, which has been approved, by

the NRC, for use at the BSEP. The proposed

changes do not affect the actual instrument

setpoints. The proposed allowable values

were calculated by applying calibration

based errors to the trip setpoint values;

thereby establishing an operability limit

associated with the entire loop of an

instrumentation function to ensure sufficient

margin to protect analytical limits. The

changes do not affect the analytical limits

associated with the involved instrumentation

functions. The involved instrumentation will

continue to perform its accident mitigation

functions as designed. Therefore, the

consequences of a previously evaluated

accident are not increased.

2. The proposed amendments would not

create the possibility of a new or different
Federal Register / Vol. 62, No. 250 / Wednesday, December 31, 1997 / Notices 68305

kind of accident from any accident

previously evaluated.

The proposed changes do not affect the

actual instrument setpoints nor do they affect

the accident mitigation instrumentation

functions. No changes will occur in the way

in which equipment is operated. The

involved instrumentation will continue to

perform its accident mitigation functions as

designed. Therefore, the proposed license

amendments can not create the possibility of

a new or different kind of accident.

3. The proposed license amendments do

not involve a significant reduction in a

margin of safety.

The proposed changes affect accident

mitigation instrumentation allowable values.

The changes will not affect the accident

mitigation instrumentation functions. No

changes will occur in the way in which

equipment is operated. The proposed

changes establish the allowable values for

certain functions in accordance with the

CP&L setpoint methodology which has been

approved, by the NRC, for use at the BSEP.

The proposed allowable values were

calculated by applying calibration based

errors to the trip setpoint values; thereby

establishing an operability limit associated

with the entire loop of an instrumentation

function to ensure sufficient margin to

protect analytical limits. The changes do not

affect the analytical limits associated with

the involved instrumentation functions. The

involved instrumentation will continue to

perform its accident mitigation functions as

designed. Therefore, the proposed license

amendments do 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

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: University of North Carolina at

Wilmington, William Madison Randall

Library, 601 S. College Road,

Wilmington, North Carolina 28403–

3297.

Attorney for licensee: William D.

Johnson, Vice President and Senior

Counsel, Carolina Power & Light

Company, Post Office Box 1551,

Raleigh, North Carolina 27602.

NRC Project Director: James E. Lyons.

Carolina Power & Light Company, et al.

Docket No. 50–400, Shearon Harris

Nuclear Power Plant, Unit 1, Wake and

Chatham Counties, North Carolina

Date of amendment request: October

29, 1997.

Description of amendment request:

Technical Specifications (TS)

3.8.1.1.a.3, 3.8.1.1.b.4, and 3.8.1.1.d.2

presently require a plant shutdown and

declaring the redundant required feature

inoperable, when the required feature

powered from the operable A.C. source

is inoperable. The proposed change

clarifies the intent of this TS to permit

the applicable redundant required

feature TS to direct a plant shutdown

when required. The proposed

amendment changes the existing TS

3.8.1.1.a.3, 3.8.1.1.b.4, and 3.8.1.1.d.2 to

eliminate the separate requirement for

plant shutdown and instead allows the

applicable required redundant feature

TS to direct the plant shutdown when

required.

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:

This change does not involve a significant

hazards consideration for the following

reasons:

1. The proposed amendment does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed amendment will not

introduce any new equipment or require

existing equipment to function different from

that previously evaluated in the Final Safety

Analysis Report (FSAR) or TS. The changes

are consistent with NUREG–1431 and the

Commission’s Final Policy Statement on

Technical Specification improvements.

Therefore, the proposed change does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. The proposed amendment does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

The proposed amendment will not

introduce any new equipment or require

existing equipment to function different from

that previously evaluated in the Final Safety

Analysis Report (FSAR) or TS. The changes

are consistent with NUREG–1431 and the

Commission’s Final Policy Statement on

Technical Specification improvements. The

proposed amendment will not create any

new accident scenarios, because the change

does not introduce any new single failures,

adverse equipment or material interactions,

or release paths.

Therefore, the proposed change does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. The proposed amendment does not

involve a significant reduction in the margin

of safety.

Margin of safety for acceptable TS action

times have been determined for each TS

related system. The proposed change will not

alter individual system TS action times. HNP

[the Harris Nuclear Plant] proposes to change

the requirement to shutdown after expiration

of the completion time of an inoperable

A.C. source concurrent with an inoperable

required feature. Instead of requiring a

shutdown, the required feature on the

inoperable A.C. source will be declared

inoperable and the individual TS will be

implemented.

In most cases with both redundant features

inoperable, a plant shutdown will be

required by TS 3.0.3. In the few instances

where additional time is allowed by the

individual TS for both redundant required

features being inoperable, then an immediate

plant shutdown would not be required. The

allowed out of service time for loss of

individual safety functions has been

previously analyzed for HNP TS and

NUREG–1431, Revision 1.

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 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: Cameron Village Regional

Library, 1930 Clark Avenue, Raleigh,

North Carolina 27605.

Attorney for licensee: William D.

Johnson, Vice President and Senior

Counsel, Carolina Power & Light

Company, Post Office Box 1551,

Raleigh, North Carolina 27602.

NRC Project Director: James E. Lyons.

Florida Power and Light Company, et

al.

[Docket Nos. 50–335 and 50–389]

St. Lucie Plant, Unit Nos. 1 and 2, St.

Lucie County, Florida

Date of amendment request:

December 1, 1997.

Description of amendment request:

The proposed amendment revises the

Unit 1 and Unit 2 Environmental

Protection Plans (EPP) Section 4,

‘‘Environmental Conditions,’’ and

Section 5, ‘‘Administrative Procedures,’’

to incorporate the proposed terms and

conditions of the Incidental Take

Statement included in the Biological

Opinion issued by the National Marine

Fisheries Service (NMFS) on February 7,

1997. The proposed amendment also

revises the wording in the Unit 1 EPP

to make it consistent with the Unit 2

EPP.

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 facility in accordance

with the proposed amendment would not

involve a significant increase in the
68306 Federal Register / Vol. 62, No. 250 / Wednesday, December 31, 1997 / Notices

probability or consequences of an accident

previously evaluated.

The changes are administrative in nature

and would in no way affect the initial

conditions, assumptions, or conclusions of

the St. Lucie Unit 1 or Unit 2, accident

analyses. In addition, the proposed changes

would not affect the operation or

performance of any equipment assumed in

the accident analyses.

Based on the above information, we

conclude that the proposed changes would

not significantly increase the probability or

consequences of an accident previously

evaluated.

(2) Use of the modified specification would

not create the possibility of a new or different

kind of accident from any previously

evaluated.

The changes are administrative in nature

and would in no way impact or alter the

configuration or operation of the facilities

and would create no new modes of operation.

We conclude that the proposed changes

would not create the possibility of a new or

different kind of accident.

(3) Use of the modified specification would

not involve a significant reduction in a

margin of safety.

As indicated in the discussion of Criterion

1, the changes are administrative in nature

and would in no way affect plant or

equipment operation or the accident analysis.

We conclude that the proposed changes

would not result in 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 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: Indian River Community

College Library, 3209 Virginia Avenue,

Fort Pierce, Florida 34981–5596.

Attorney for licensee: M.S. Ross,

Attorney, Florida Power & Light, P.O.

Box 14000, Juno Beach, Florida 33408–

0420.

NRC Project Director: Frederick J.

Hebdon.

IES Utilities Inc.

[Docket No. 50–331]

Duane Arnold Energy Center, Linn

County, Iowa

Date of amendment request: October

30, 1996.

Description of amendment request:

The proposed amendment, included as

part of the proposed conversion from

current Technical Specifications (CTS)

to improved Technical Specifications

(ITS), would modify the Surveillance

Requirements (SRs) recommended in

NUREG–1433 LOC 3.5.1 by revising the

combinations (Conditions C, D, G, and

I of ITS 3.5.1) of emergency core cooling

systems/subsystems that may be out of

service. The combinations are supported

by the Duane Arnold Energy Center

(DAEC) Loss-of-Coolant Accident

(LOCA) analysis.

Condition C

ITS 3.5.1 Action C establishes

Required Actions and Completion

Times for the situation when one core

spray (CS) subsystem and one or two

residual heat removal (RHR) pump(s)

are inoperable. The proposed

specification is less restrictive than CTS

3.5.A.4, which allows one RHR pump to

be inoperable for 30 days, and CTS

3.5.A.5, which allows two RHR pumps

(i.e., the low pressure coolant injection

(LPCI) subsystem) to be inoperable for

up to 7 days, provided the remaining

RHR (i.e., LPCI) active components,

both CS subsystems, the containment

spray subsystem, and the diesel

generators are verified to be operable.

The CTS does not allow one CS

subsystem and one or two RHR pump(s)

to be inoperable at the same time. The

LOCA analysis presented in NEDC–

31310P, (Duane Arnold Energy Center

SAFER/GESTR–LOCA Loss-of-Coolant

Accident Analysis), indicates that an

adequate level of protection is provided

by the remaining operable ECCS

subsystems. The accident analysis also

demonstrates that in this condition, the

peak clad temperature remains below

the regulatory limit. However, another

single failure may place the plant in a

condition where adequate core cooling

may not be available during a DBA–

LOCA. Therefore, a Completion Time of

72 hours has been proposed to either

restore the inoperable CS subsystem or

the inoperable RHR pump(s).

Condition D

ITS 3.5.1 Action D establishes

Required Actions and Completion

Times for the situation when two CS

subsystems are inoperable. The

proposed specification is less restrictive

than CTS 3.5.A.2, which allows only

one CS subsystem to be inoperable. CTS

3.5.A.6 would require the plant to be in

Hot Shutdown within 12 hours and

Cold Shutdown within the following 24

hours if both CS subsystems were

inoperable. With two CS subsystems

inoperable, the LOCA analysis

presented in NEDC–31310P, (Duane

Arnold Energy Center SAFER/GESTR–

LOCA Loss-of-Coolant Accident

Analysis), indicates that the remaining

operable low pressure ECCS subsystem

consisting of LPCI with four RHR

pumps operable (only 3 pumps

required), provides adequate protection.

However, another single failure may

place the plant in a condition where

adequate core cooling may not be

available during a Design Basis

Accident LOCA. Therefore, a

Completion Time of 72 hours has been

proposed to restore one CS subsystem to

operable status.

Condition G

ITS 3.5.1 Action G establishes

Required Actions and Completion

Times for the situation when HPCI and

one RHR pump are inoperable. The

proposed specification is less restrictive

than CTS 3.5.D.2, which allows

continued operation if HPCI is

inoperable only if both CSs, LPCI, ADS,

and RCIC are verified to be operable.

While the LPCI subsystem is technically

operable with only 3 of 4 RHR pumps

operable, the CTS is currently

interpreted by DAEC to require all 4

RHR pumps to be operable for the

requirements of CTS 3.5.D.2 to be met,

as a single RHR pump has more makeup

capability than the HPCI System. Thus

for mitigating small and intermediate

break LOCAs, one LPCI pump, in

combination with ADS, is more than

adequate core cooling. The condition of

when HPCI and one RHR pump are

inoperable is bounded by the analysis in

NEDC–31310P, Duane Arnold Energy

Center, SAFER/GESTR-LOCA Loss-ofCoolant Accident Analysis. Since the

remaining operable low pressure ECCS

subsystems are more than capable of

performing their intended function, and

RCIC and ADS are Operable, the

proposed Action G maintains LOCA

analysis assumptions for ECCS

Operability. The proposed ITS

condition allows 7 days to restore the

HPCI System or the RHR pump to

operable status. The licensee considers

the 7 day Completion Time reasonable

in that the LOCA analysis demonstrates

that in this condition, the peak clad

temperature remains below the

regulatory limit. The 7 day Completion

Time also provides the benefit of

potentially avoiding an unnecessary

plant shutdown while the safety

functions are still capable of being

performed.

Condition I

ITS 3.5.1 Action I establishes

Required Actions and Completion

Times for the situation when HPCI and

one ADS valve are inoperable. The

proposed Specification is less restrictive

than CTS 3.5.D.2, which allows

continued operation if HPCI is

inoperable only if both CSs, LPCI, ADS,

and RCIC are verified to be operable.

While ADS is capable of performing its

design function with only 3 of 4 valves

operable, per NEDC–31310P, Duane

Arnold Energy Center, SAFER/GESTR-
Federal Register / Vol. 62, No. 250 / Wednesday, December 31, 1997 / Notices 68307

LOCA Loss-of-Coolant Accident

Analysis, the CTS requires all 4 ADS

valves to be operable for the

requirements of CTS 3.5.D.2 to be met.

The proposed specification is less

restrictive than CTS 3.5.F.2, which

allows continued operation when one

ADS valve is inoperable only if HPCI is

verified to be operable. Since all low

pressure ECCS subsystems remain

capable of performing their design

function and ADS is still capable of

performing its design function, ITS 3.5.1

Action I maintains LOCA assumptions

to ensure an adequate level of protection

is maintained. The proposed condition

allows 72 hours to restore the HPCI

system or the ADS valve to operable

status, since another single failure (i.e.,

loss of another ADS valve), may place

the plant in a condition where adequate

core cooling may not be available during

a small or intermediate break LOCA.

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:

For Condition C

1. Does the change involve a significant

increase in the probability or consequences

of an accident previously evaluated?

The proposed change will allow one Core

Spray subsystem and one or two RHR

pump(s) to be inoperable for up to 72 hours.

The ECCS subsystems affected by this change

are not assumed to be initiators of analyzed

events. Therefore, the proposed change does

not increase the probability of any accident.

The role of these ECCS subsystems is in the

mitigation of accident consequences. The

proposed change does not allow unlimited

continuous operation with the plant in a

condition where an additional single failure

could result in a loss of ECCS function. The

proposed change does not increase the

consequences of an accident because

accident analysis presented in NEDC–

31310P, Duane Arnold Energy Center

SAFER/GESTR–LOCA Loss-of-Coolant

Accident Analysis, indicates that an adequate

level of protection is maintained by the ADS

System and the remaining Operable ECCS

subsystems when one Core Spray subsystem

and one or two RHR pump(s) are inoperable.

Therefore, this change will not involve a

significant increase in the consequences of an

accident previously evaluated.

2. Does the change create the possibility of

a new or different kind of accident from any

accident previously evaluated?

The proposed change will not involve any

physical changes to plant systems, structures,

or components (SSCs), or the manner in

which these SSCs are operated, maintained,

modified, tested or inspected. The change

ensures the remaining ECCS capability is

adequate to mitigate the consequences of

accidents. Therefore, this change will not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. Does this change involve a significant

reduction in a margin of safety?

The proposed change does not significantly

reduce the margin of safety because accident

analysis presented in NEDC–31310P, Duane

Arnold Energy Center SAFER/GESTR–LOCA

Loss-of-Coolant Accident Analysis, indicates

that the plant is protected by the ADS System

and the remaining ECCS subsystems when

one Core Spray subsystem and one or two

RHR pump(s) are inoperable. The accident

analysis demonstrates that in this condition,

the peak clad temperature remains below the

regulatory limit. However, with one Core

Spray subsystem and one or two RHR

pump(s) inoperable, another single failure

may place the plant in a condition where

adequate core cooling may not be available

during a DBA–LOCA. Therefore, a

Completion Time of 72 hours has been

assigned to either restore the inoperable Core

Spray subsystem or the RHR pump. In

addition, this change provides the benefit of

potentially avoiding an unnecessary plant

shutdown (due to a Completion Time being

provided for one Core Spray subsystem and

one or two RHR pump(s)) when the

remaining ECCS subsystems and the ADS are

capable of mitigating potential events.

Therefore, this change does not involve a

significant reduction in a martin safety.

For Condition D

1. Does the change involve a significant

increase in the probability or consequences

of an accident previously evaluated?

The proposed change will allow both Core

Spray subsystems to be inoperable for up to

72 hours. The ECCS subsystems affected by

this change are not assumed to be initiators

of analyzed events. Therefore, the proposed

change does not increase the probability of

any accident. The role of these ECCS

subsystems is in the mitigation of accident

consequences. The proposed change does not

allow unlimited continuous operation with

the plant in a condition where an additional

single failure could result in a loss of ECCS

function. The proposed change does not

increase the consequences of an accident

because accident analysis presented in

NEDC–3131OP, Duane Arnold Energy Center

SAFER/GESTR–LOCA Loss-of-Coolant

Accident Analysis, indicates that an adequate

level of protection is maintained by the ADS

System and remaining Operable ECCS

subsystem when two Core Spray subsystems

or inoperable. Therefore, this change will not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. Does the change create the possibility of

a new or different kind of accident from any

accident previously evaluated?

The proposed change will not involve any

physical changes to plant systems, structures,

or components (SSCs), or the manner in

which these SSCs are operated, maintained,

modified, tested, or inspected. The change

ensures the remaining ECCS capability is

adequate to mitigate the consequences of

accidents. Therefore, this change will not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. Does this change involve a significant

reduction in a margin of safety?

The proposed change does not significantly

reduce the margin of safety because accident

analysis presented in NEDC–31310P, Duane

Arnold Energy Center SAFER/GESTR–LOCA

Loss-of-Coolant Accident Analysis, indicates

that the plant is protected by the ADS System

and the remaining ECCS subsystem when

two Core Spray subsystems are inoperable.

The accident analysis demonstrates that in

this condition, the peak clad temperature

remains below the regulatory limit. However,

with both Core Spray subsystems inoperable,

another single failure may place the plant in

a condition where adequate core cooling may

not be available during a DBA–LOCA.

Therefore, a Completion Time of 72 hours

has been assigned to restore one inoperable

Core Spray subsystem. In addition this

change provides the benefit of potentially

avoiding an unnecessary plant shutdown

(due to a Completion Time being provided

for both Core Spray subsystems inoperable)

when the remaining ECCS subsystem and the

ADS are capable of mitigating potential

events. Therefore, this change does not

involve a significant reduction in a margin of

safety.

Condition G

1. Does the change involve a significant

increase in the probability or consequences

or an accident previously evaluated?

The proposed change will allow the HPCI

System and one RHR pump to be inoperable

for up to 7 days. The ECCS subsystems

affected by this change are not assumed to be

initiators of analyzed events. Therefore, the

proposed change does not increase the

probability of any accident. The role of these

ECCS subsystems is in the mitigation of

accident consequences. The proposed change

does not allow unlimited continuous

operation with the plant in a condition where

an additional single failure could result in a

loss of ECCS function. The proposed change

does not increase the consequences of an

accident because accident analysis presented

in NEDC–31310P, Duane Arnold Energy

Center SAFER/GESTRA–LOCA Loss-ofCoolant Accident Analysis, indicated that an

adequate level of protection is maintained by

the ADS System and the remaining Operable

ECCS subsystems when HPCI and one RHR

pump are inoperable. Therefore, this change

will not involve a significant increase in the

consequences of an accident previously

evaluated.

2. Does the change create the possibility of

a new or different kind of accident from any

accident previously evaluated?

The proposed change will not involve any

physical changes to plant systems, structures,

or components (SSCs), or the manner in

which these SSCs are operated, maintained,

modified, tested, or inspected. The change

ensures the remaining ECCS capability is

adequate to mitigate the consequences of

accidents. Therefore, this change will not

create the possibility of a new or different

kind of accident from any accident

previously evaluate.

3. Does this change involve a significant

reduction in a margin of safety?

The proposed change does not significantly

reduce the margin of safety because accident
68308 Federal Register / Vol. 62, No. 250 / Wednesday, December 31, 1997 / Notices

analysis presented in NEDC–31310P, Duane

Arnold Energy Center SAFER/GESTR–LOCA

Loss-of-Coolant Accident Analysis, indicates

that the plant is protected by the ADS System

and the remaining ECCS subsystems when

HPCI and one RHR pump are inoperable. The

accident analysis demonstrates that in this

condition, the peak clad temperature remains

below the regulatory limit. However, with

both HPCI and one RHR pump inoperable,

another single failure may place the plant in

a condition where adequate core cooling may

not be available during an accident.

Therefore, a Completion Time of 7 days has

been assigned to either restore the inoperable

HPCI System or the RHR pump. In addition,

this change provides the benefit of

potentially avoiding an unnecessary plant

shutdown (due to a Completion Time being

provided for the HPCI System and one RHR

pump inoperable) when the remaining ECCS

subsystems and the ADS are capable of

mitigating potential events. Therefore, this

change does not involve a significant

reduction in a margin of safety.

Condtion I

1. Does the change involve a significant

increase in the probability or consequences

of an accident previously evaluated?

The proposed change will allow the HPCI

system and one ADS valve to be inoperable

for up to 72 hours. The ECCS subsystems

affected by this change are not assumed to be

initiators or analyzed events. Therefore, the

proposed change does not increase the

probability of any accident. The role of these

ECCS subsystems is in the mitigation of

accident consequences. The proposed change

does not allow unlimited continuous

operation with the plant in a condition where

an additional single failure could result in a

loss of ECCS function. The proposed change

does not increase the consequences of an

accident because accident analysis presented

in NEDC–31310P, Duane Arnold Energy

Center SAFER/GESTER–LOCA Loss-ofCoolant Accident Analysis, indicates that an

adequate level of protection is maintained by

the remaining ADS valves (the ADS design

function is maintained) in combination with

the remaining Operable ECCS subsystems

when HPCI and one ADS valve are

inoperable. Therefore, this change will not

involve a significant increase in the

consequences of an accident previously

evaluated.

2. Does the change create the possibility of

a new or difference kind of accident form any

accident previously evaluated?

The proposed change will not involve any

physical changes to plant systems, structures,

or components (SSCs) or the manner in

which these SSCs are operated, maintained,

modified, tested, or inspected. The change

ensures the remaining ECCS capability in

adequate to mitigate the consequences of

accidents. Therefore, this change will not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. Does this change involve a significant

reduction in a margin of safety?

The proposed change does not significantly

reduce the margin of safety because accident

analysis presented in NEDC–31310P, Duane

Arnold Energy Center SAFER/GESTR-LOCA

Loss-of-Coolant Accident Analysis, indicates

that the plant is protected by the remaining

ADS valves and the low pressure ECCS

subsystems when HPCI and one ADS valve

are inoperable. The accident analysis

demonstrates that in this condition, the peak

clad temperature remains below the

regulatory limit. However, with both HPCI

and one ADS valve inoperable, another single

failure (i.e., of an ADS valve) may place the

plant in a condition where adequate core

cooling may not be available during a small

or intermediate break LOCA. Therefore, a

Completion Time of 72 hours has been

assigned to either restore the inoperable HPCI

System or the ADS valve. In addition, this

change provides the benefit of potentially

avoiding an unnecessary plant shutdown

(due to a Completion Time being provided

for the HPCI System and one ADS valve

inoperable) when the remaining ECCS

subsystems and ADS valves are capable of

mitigating potential events. Therefore, this

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: Cedar Rapids Public Library,

500 First Street, S.E., Cedar Rapids,

Iowa 52401.

Attorney for licensee: Jack Newman,

Kathleen H. Shea, Morgan, Lewis, &

Bockius, 1800 M Street, NW.,

Washington, DC 20036–5869.

Acting NRC Project Director: Richard

P. Savio.

Indiana Michigan Power Company

[Docket Nos. 50–315 and 50–316]

Donald C. Cook Nuclear Plant, Units 1

and 2, Berrien County, Michigan

Date of amendment requests: August

1, 1997 (AEP:NRC:0906H).

Description of amendment requests:

The proposed amendments would

revise Technical Specification

surveillance 4.7.1.2.b. to delete the

requirement that the test be performed

at a specified secondary steam supply

pressure.

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:

Criterion 1

The proposed changes will not

significantly increase the probability or

consequences of an accident previously

evaluated.

This is an administrative change intended

to clarify the technical specification. There

will be no change to the test procedure as a

result of this clarification. The proposed

change better correlates with the accident

requirements for which TDAFP [turbine

driven auxiliary feed pump] flow is required,

and the change is consistent with the present

requirement of testing the TDAFP at a

secondary side pressure greater than 310

psig.

Criterion 2

The proposed changes will not create the

possibility of a new or different kind of

accident from any accident previously

evaluated. The proposed change does not

physically modify the plant, nor does it

result in the installation of equipment which

could introduce a new failure mechanism.

Criterion 3

The proposed change does not involve a

significant reduction in a margin of safety.

The proposed change does not affect the

performance of the TDAFP. Thus, the TDAFP

remains capable of providing the required

flow under accident conditions, and no

safety margins are reduced.

This is an administrative change intended

to clarify the technical specification. There

will be no change to the test procedure as a

result of this clarification

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 requests involve no

significant hazards consideration.

Local Public Document Room

location: Maud Preston Palenske

Memorial Library, 500 Market Street, St.

Joseph, MI 49085.

Attorney for licensee: Gerald Charnoff,

Esq., Shaw, Pittman, Potts and

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037.

NRC Project Director: Richard P.

Savio, Acting.

Indiana Michigan Power Company

[Docket Nos. 50–315 and 50–316]

Donald C. Cook Nuclear Plant, Units 1

and 2, Berrien County, Michigan

Date of amendment requests: August

11, 1997 (AEP:NRC:1265).

Description of amendment requests:

The proposed amendments would

revise the Technical Specifications (TS)

to allow the filling of the emergency

core cooling system (ECCS)

accumulators without declaring ECCS

equipment inoperable.

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:
Federal Register / Vol. 62, No. 250 / Wednesday, December 31, 1997 / Notices 68309

Criterion 1

This amendment request does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated because the proposed changes to

the T/S represent the possibility of an event

that has such a low probability as to not be

considered credible. A calculation was

performed that demonstrated the CDF

resulting from the accumulator fill line

operation with all of the conditions assumed

above is approximately 3 x 10¥–10 per year.

This is well below the NEI guidelines of 1 x

10¥6 for acceptable risk for a given evolution.

Therefore, based on probabilistic

considerations and the robust design of the

pumps, we conclude the risk associated with

this proposed change will not result in a

significant increase in the probability or

consequences of a previously evaluated

accident.

Criterion 2

The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated. The change does not involve a

physical change to the plant, but does

involve a change in the plant operating

configuration. The possibility of a LBLOCA

[large break loss of coolant accident]

occurring during the accumulation fill

evolution has been evaluated and determined

to not be credible. Westinghouse has

confirmed the accumulator fill line was not

modeled in the accident analyses due to the

extremely short duration of the fill operation

and the extremely small amount of flow that

the fill line is capable of passing. The overall

effect this configuration would have on the

capability of the SI [safety injection] pump to

perform its design function, should a

LBLOCA occur during the extremely brief

window of opportunity, is negligible and

would not create a new type of accident.

Criterion 3

This proposed change does not involve a

significant reduction in a margin of safety, as

the risk from the postulated sequence of

events is insignificant. Additionally,

engineering evaluation has determined that

the real response of an SI pump under the

postulated conditions would not be severe.

The rugged construction of the pumps, and

the design margin built into them, are factors

that support the engineering judgment that

the affected pump would continue to operate

for some time, at some capacity beyond the

manufacturer’s design limit. As a result of

exceeding the limit, the pump may

experience some cavitation and require

additional corrective maintenance, but would

be expected to deliver a significant fraction

of its design flow.

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 requests involve no

significant hazards consideration.

Local Public Document Room

location: Maud Preston Palenske

Memorial Library, 500 Market Street, St.

Joseph, MI 49085.

Attorney for licensee: Gerald Charnoff,

Esq., Shaw, Pittman, Potts and

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037.

NRC Project Director: Richard P.

Savio, Acting.

Niagara Mohawk Power Corporation

[Docket No. 50–410]

Nine Mile Point Nuclear Station, Unit 2,

Oswego County, New York

Date of amendment request: October

7, 1997.

Description of amendment request:

The proposed amendment would revise

Technical Specifications (TS) to change

the setpoints of Surveillance

Requirements (SRs) 4.9.6.a, 4.9.6.f, and

4.9.6.g for the refueling platform main

hoist. Specifically, each refueling

platform crane or hoist used for

handling control rods or fuel assemblies

within the reactor pressure vessel would

be demonstrated operable by:

a. Demonstrating operation of the

overload cutoff on the main hoist when

the load exceeds 1600 +100/¥0 pounds

(rather than 1200 +50/¥50 pounds).

f. Demonstrating operation of the

loaded interlock on the main hoist when

the load exceeds 700 +50/¥0 pounds

(rather than 485 +50/¥50 pounds).

g. Demonstrating operation of the

redundant loaded interlock on the main

hoist when the load exceeds 700 +50/

¥0 pounds (rather than 550 +50/¥50

pounds).

The proposed amendment, in effect,

would authorize replacement of the

existing triangular refueling platform

mast with a round, heavier mast

(General Electric Model NF–500) which

includes an installed camera/TV 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:

The operation of Nine Mile Point Unit 2,

in accordance with the proposed

amendment, will not involve a significant

increase in the probability or consequences

of an accident previously evaluated.

The proposed change revises the setpoints

for three TS SRs based on modifications to

the refueling platform mast. The new mast is

essentially a direct replacement for the

existing mast, with the exception that the

new mast is approximately 400 lbs. heavier,

which directly affects the setpoints. No

change in the frequency or manner in which

the surveillances are performed is proposed.

Refueling interlocks will continue to function

as designed. No changes to the methods in

which plant systems are operated are

required. The same design criteria and

standards were applied to the new mast,

including the seismic capability of the

refueling platform with the heavier mast.

Therefore, none of the precursors of

previously evaluated accidents are affected,

and no new failure modes are introduced.

Based on the additional weight of the new

mast and camera/TV system, the revised

GESTAR [General Electric GESTAR II

document NEDE–24011-P-A–11-U5] criteria

for fuel rod damage (more conservative

threshold level), the use of GE11 [9x9] fuel

for the bundle drop analysis, the number of

damaged fuel rods has increased slightly for

the potential fuel handling accident. The

results of this increase were evaluated and

dispositioned against the bounding

calculation to show that the current USAR

[updated safety analysis report] analysis

bounds the revised radiological

consequences which remain well within the

GDC [General Design Criterion] 19 and

10CFR[part]100 limits. The systems that are

available to mitigate the consequences of any

accident have not been affected and are still

capable of performing their required

functions. Therefore, this change will not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The operation of Nine Mile Point Unit 2,

in accordance with the proposed

amendment, will not create the possibility of

a new or different kind of accident from any

accident previously evaluated.

The proposed change revises the setpoints

for three TS SRs based on installation of a

new refueling platform which is heavier than

the current mast. No change in the frequency

or manner in which the surveillances are

performed has occurred. Refueling interlocks

will continue to function as designed. No

changes to the methods in which plant

systems are operated are required. The same

design criteria and standards were applied to

the new mast, including the seismic

capability of the refueling platform with the

heavier mast. The basic function and

operation of the refueling platform is

unchanged. The uptravel stop and

downtravel mechanical cutoff setpoints are

not being changed and will continue to

ensure that adequate water shielding is

maintained. As such, the change does not

introduce any new failure modes or

conditions that may create a new or different

kind of accident. Therefore, this change will

not create the possibility of a new or different

kind of accident from any accident

previously evaluated.

The operation of Nine Mile Point Unit 2,

in accordance with the proposed

amendment, will not involve a significant

reduction in a margin of safety.

The proposed change revises three TS SR

setpoints based on installation of a new

refueling platform mast. No change in the

frequency or manner in which the

surveillances are performed has occurred.

Refueling interlocks will continue to function

as designed. No changes to the methods in

which plant systems are operated are

required. The same design criteria and

standards were applied to the new mast,

including the seismic capability of the
68310 Federal Register / Vol. 62, No. 250 / Wednesday, December 31, 1997 / Notices

refueling platform with the heavier mast. The

addition of a camera/TV system will provide

enhanced visibility for fuel handling

activities and additional assurance that the

grapple is oriented over the correct fuel

bundle.

The additional weight of the new mast has

been evaluated and the operability

requirements as described in the TS and TS

Bases are unchanged. The modification and

revised setpoints do not change the function

of the refueling platform main hoist. The

revised setpoints will continue to assure the

lifting capacity of the main hoist will not be

sufficient to result in damage to core

internals or the reactor pressure vessel in the

event that they are accidentally engaged.

The necessary systems are still available to

mitigate any potential radiological

consequences of the increased number of

damaged fuel rods. The radiological

consequences remain within the bounds of

the current safety analysis and well below

the GDC 19 and 10CFR[Part]100 limits.

Therefore, the change does not involve any

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 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: Reference and Documents

Department, Penfield Library, State

University of New York, Oswego, New

York 13126.

Attorney for licensee: Mark J.

Wetterhahn, Esquire, Winston & Strawn,

1400 L Street, NW., Washington, DC

20005–3502.

NRC Project Director: S. Singh Bajwa.

Niagara Mohawk Power Corporation

[Docket No. 50–410]

Nine Mile Point Nuclear Station, Unit 2,

Oswego County, New York

Date of amendment request: October

31, 1997.

Description of amendment request:

The proposed amendment would revise

Technical Specifications (TSs) to

support installation of the General

Electric Nuclear Measurement Analysis

and Control (NUMAC) Power Range

Neutron Monitor (PRNM) System. The

TS changes apply to Sections 2.2,

‘‘Limiting Safety System Settings’’; 3/

4.3.1, ‘‘Reactor Protection System

Instrumentation’’ and its corresponding

Bases; and 3/4.3.6, ‘‘Control Rod Block

Instrumentation.’’

Basis for proposed no significant

hazards consideration determination:

The NUMAC–PRNM will monitor

groups of Local Power Range Monitor

(LPRM) signals and, together with the

Oscillation Power Range Monitor

(OPRM), initiate a reactor scram upon

identifying neutron flux oscillations

characteristic of a thermal-hydraulic

instability. The NUMAC–PRNM will

replace the existing Average Power

Range Monitor (APRM) System and will

ultimately support the activation of the

OPRM. The proposed modification is in

response to Generic Letter 94–02,

‘‘Long-Term Solutions and Upgrade of

Interim Operating Recommendations for

Thermal-Hydraulic Instabilities in

Boiling Water Reactor.’’ Except for

minor deviations, the proposed TS

changes are consistent with General

Electric Licensing Topical Report (LTR),

NEDC–32410P–A, ‘‘Nuclear

Measurement Analysis and Control

Power Range Neutron Monitor

(NUMAC–PRNM) Retrofit Plus Option

III Stability Trip Function,’’ which was

approved by the NRC staff September 5,

1995. Changes with respect to response

time testing requirements would be

based on Supplement 1 to NEDC–

32410P–A, approved by the NRC staff

December 26, 1996.

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:

The operation of Nine Mile Point Unit 2,

in accordance with the proposed

amendment, will not involve a significant

increase in the probability or consequences

of an accident previously evaluated.

As discussed in NEDC–3241OP–A, the

NUMAC–PRNM modification and associated

changes to the TS involve systems that are

intended to detect the symptoms of certain

events or accidents mitigating actions. The

worst case failure of the systems involved

would be a failure to initiate mitigative

actions (i.e., scram or rod block), but no

failure can cause an accident and therefore

the probability of precursors of any accidents

previously evaluated is not increased. The

NUMAC–PRNM system performs the same

operations as the existing equipment, reduces

the need for tedious operator action during

normal conditions and allows the operator to

focus more on overall plant conditions.

Automatic self-test and increased operator

information available with the NUMARC–

PRNM system is likely to reduce the burden

during off-normal conditions as well. The

NUMAC–PRNM system is compatible with

the environmental conditions at the

mounting location (e.g., temperature,

humidity, seismic, electromagnetic fields)

such that system performance will not be

degraded when compared to the system being

replaced. Therefore, the proposed change

will not result in a significant increase in the

probability of any accidents previously

evaluated.

The proposed changes to the RPS [reactor

protection system] and Control Rod Block

instrumentation TSs are necessitated by the

NUMAC–PRNM replacement. As discussed

in the evaluation, in the 4 APRM channel

configuration, any two of the four APRM

channels and one 2-out-of-4 voter channel in

each RPS trip system are required to function

for the APRM safety trip function to be

accomplished. Therefore, the proposed TS

change requires that 3 of the 4 APRM

channels be operable. This assures at least

two APRM channels to each of the 2-out-of4 voter channels are available in the event of

a single APRM channel failure and one

APRM is bypassed. Also, the proposed TS

requires a minimum of two 2-out-of-4 voter

channels per RPS trip system (i.e., all four

voter channels). This assures that at least one

voter channel per trip system is available

even in the event of a single voter channel

failure. Surveillance testing requirements

were revised to take advantage of certain

features of the NUMAC–PRNM (digital)

replacement of the existing analog APRM

system. These advantages included improved

accuracy, stability , self-testing, reduced drift,

and constant time for digital processing.

Testing of the RPS and Control Rod Block

instrumentation will continue to be

performed as described in the evaluation to

assure that the reliability and performance of

these systems will not be adversely affected.

The proposed NUMAC–PRNM

replacement system has been specifically

designed to assure that the system response

times meet the current acceptance limits

(worst case). As a result, due to statistical

variations resulting from the sampling and

update cycles, the response time is typically

faster than required in order to assure the

required response time is always met. The

architecture of the NUMAC–PRNM system

has reduced segmentation compared to the

existing PRM system. Examples of the

reduced segmentation are combining

previously separate functions, several input

channels sharing an input board, and a

central loop processor for many channels.

The replacement equipment includes up to 5

LPRM inputs on a single module compared

to one per module on the current system. Up

to 17 LPRM signals are processed through

one preprocessor. The recirculation flow

signals are processed in the same hardware

as the LPRM processing. The net effect of

these architectural aspects is that there are

some single failures that cause a greater loss

of ‘‘sub-functionality’’ than in the current

system. However, other architectural and

functional aspects have an offsetting effect.

Redundant power supplies are used so that

a single failure of AC power has no effect on

the overall NUMAC–PRNM system functions

while still resulting in a half scram, as does

the current system. Continuous automatic

self-test also assures that if a single failure

does occur, it is much more likely to be

detected immediately. The net effect is that

from a total system level, there is no

increased risk of loss of critical functionality

or reduction in safety margins due to the

architecture of the replacement system.

Failure analysis indicates that a software

common cause failure is not a significant

contributor to the unavailability of the

NUMAC–PRNM. However, in spite of that

conclusion, means are provided within the

system to mitigate the effects of such a failure

and alert an operator. Therefore, such a

failure, even if it occurred, will not increase
Federal Register / Vol. 62, No. 250 / Wednesday, December 31, 1997 / Notices 68311

the consequences of a previously evaluated

accident. To reduce the likelihood of

common cause failures of software controlled

functions, thorough and careful verification

and validation (V&V) activities are performed

both for the requirements and the

implementing software design. In addition,

the software is designed to limit the loading

that external systems or equipment can place

on the system, thus significantly reducing the

risk that some abnormal dynamic condition

external to the system can cause an overload.

For conservatism, however, despite, these

V&V activities, common cause failures of

software controlled functions due to residual

software design faults are assumed to occur.

Both the software and hardware are designed

to manage the consequences of such failures.

Safety outputs are designed to be fail safe by

requiring dynamic update of output modules

or data signals, where failure to update the

information is detected by simple receiving

hardware, which in turn, forces a trip. This

aspect covers all but rather complex failures

where the hardware or software executes a

portion of the overall logic but fails to

process some portion of the new information

(inputs ‘‘freeze’’) or some portion of the logic

(outputs ‘‘freeze’’). To help reduce the

likelihood of complex failures, a watchdog

timer is used which is updated by a very

simple software routine that in turn monitors

the operational cycle time of all tasks in the

system. The software design is such that as

long as all tasks are updating at the design

rate, it is likely that software controlled

functions are executing as intended.

Conversely, if any task fails too update at the

design rate, that is a strong indication of at

least some unanticipated condition. If such a

condition occurs, its watchdog timer will not

be updated, the computer will be restarted,

and the outputs will detect an abnormal

condition and provide an alarm.

It is very difficult to quantify a software

common cause failure rate. Analyses for the

current system did consider common cause

failures and assessed them to be at a rate of

about 0.3 times the random failure rate. The

reference analysis uses a field basis for the

random rates. The analysis for the

replacement design uses conservative

estimates for failure rates of equipment that

are actually a little higher than those

assumed for the current equipment. The

methodology being applied concludes that

the common mode failure rate for the

replacement system is somewhat higher than

the current system. However, that is offset by

more frequent surveillance tests performed

by the self-test that result in an estimated

slightly lower unavailability for the

NUMAC–PRNM scram function compared to

the current PRM system. The USAR, in

general, considers the failure rate of the

function, not that of sub-components. On that

basis, there will not be an increase, due to

software common cause failure, in the

probability of a malfunction analyzed in the

USAR.I21The NUMAC–PRNM humanmachine interface design does not introduce

an increased burden or constraints on the

operators’ ability to adequately respond to an

accident such that there would be more

severe consequential effects. The information

available to the operators is the same as with

the current system. No actions are required

by the operator to obtain information

normally used and equivalent to that

available with the current equipment.

However, the replacement system does

provide more direct accessible information

regarding the condition of the equipment,

including automatic self-test, which can aid

the operator in diagnosing unusual situations

beyond those defined in the licensing basis.

The replacement system has a significantly

lower power requirement and is generally

smaller, reducing somewhat the seismic

loading on the panels. The equipment

qualification also includes EMI [electro

magnetic induction] emissions which,

combined with the fact that the replacement

equipment is mounted in its own cabinet

(replaces all of the current equipment),

minimized the likelihood of significant

impact on other existing equipment.

The replacement equipment makes

increased use of qualified optical methods to

provide both safety and functional isolation

between safety-related and nonsafety-related

systems. Where fiber optic methods cannot

be used, the isolation provided is comparable

to or better than that provided in the current

system.

The net electrical and thermal load for the

replacement system is less than that for the

current system. Accordingly, the replacement

system had adequate cabinet cooling and no

forced cooling is required.

The replacement system meets or exceeds

all applicable requirements for separation,

independence and grounding. The use of

fiber optic connections between the APRM

and RBM [rod block monitor] improves the

separation and reduces the dependence of

the system on common grounds. However,

for noise rejection, the equipment design and

manufacturing requirements assure improved

grounding of the actual equipment.

No change in wiring or grounding external

to the panels containing the replacement

equipment is necessary for correct operation

of the replacement equipment.

NEDC–3241OP–A, Section 3.2.3, discusses

different plant configurations for

recirculation flow channels, including the

case where plants currently (before

implementing the NUMAC PRNM system)

have four flow channels. Absence of any

discussion in the LTR related to separation

for plants originally having four flow

channels implies that those plants are

expected to meet full separation

requirements. The LTR includes a further

statement that ‘‘The criterion is to maintain

equal or better protection against single

failures while allowing bypassing of the

APRM channel that processes the flow

signal.’’

The NMPC [Niagara Mohawk Power

Corporation] NUMAC PRNM system has four

recirculation low channels, but the flow

input circuits for two of the four are not

separated from each other outside the PRNM

panel. As a result, a single failure that causes

both of these flow signals to go high could,

depending on the specific value, cause the

APRM flow biased trip setpoint in two

channels to go to the clamped setpoint. If, at

the same time, a third channel is bypassed,

the APRM flow-biased trip setpoint for the

APRM system could be non-conservative.

(NOTE: The flow signals are compared to one

another. Should the flow signals not be

within specified limits, an alarm and a

control rod block would be initiated.)

Despite the fact that two of the four flow

input circuits are not separated from each

other outside the PRNM panel, the

replacement system is judged to be adequate

with the current field routing of flow signals

and meets the LTR criteria. This conclusion

is based on the fact that there is no credible

fault in the circuits within the duct, in which

the flow signals are routed, that can damage

the other circuits. Also, there is no credible

external fault that can damage the circuits

inside the duct. Therefore, it is concluded

that the separation between the two flow

input circuits is adequate to meet the system

single failure requirements in that no

credible single failure will disable the flow

inputs to more than one APRM channel.

Additionally, there are no reload licensing

transient analyses that take credit for the

flow-biased simulated thermal power scram

setpoint.

The replacement design has been

specifically designed to have the same or

more conservative ‘‘fail safe’’ failure modes

as the current system. For example, in the

case of a single power bus failure, the current

system loses about one half of the LPRM

information and an output trip occurs. For

the replacement system, that failure still

results in an output trip, but no LPRM

information is lost. In the current system, a

static failure in several areas in the system

could result in a ‘‘fail-as-is’’ state of the

outputs. In the replacement system, dynamic

coupling starting in the main processor and

going to the final output virtually eliminates

‘‘fail-as-is’’ failure modes and replaces them

with ‘‘fail tripped’’ modes.

The replacement system has the same loss

of power failure mode as the current system

relative to the trip outputs and for loss of AC

[alternating current] power. For loss of DC

[direct current] power, the replacement

system in most cases continues to operate

normally due to redundancy of the power

supplies. Therefore, the consequences are no

different or improved compared to those

considered in the USAR.

Both the current system and the

replacement system automatically startup on

application of power (or re-application).

However, the replacement system may take

slightly longer to reach normal operation due

to initializing activities. However, no USAR

evaluations take credit for rapid start of the

PRM. Therefore, the slightly longer startup

time from point of power application is

bounded by the USAR analysis. Upon

application of power, once the system is set

up for the specific application, it

automatically returns to those settings upon

application of power. All such setup

parameters are stored in non-volatile

memory.

Human-machine interfaces (HMI) failures

in the current system could be related to

misadjusted settings, incorrect reading of

meters, and failure to return the equipment

to the normal operating configuration. There

are comparable failure modes for some of

these in the digital system where an
68312 Federal Register / Vol. 62, No. 250 / Wednesday, December 31, 1997 / Notices

erroneous potentiometer adjustment in the

current system is equivalent to an erroneous

digital entry in the replacement system.

Certain potential ‘‘failure to reconfigure’’

errors in the current system have no

counterpart in the replacement system

because any ‘‘reconfiguration’’ is

automatically returned to normal by the

system. Also, since parameters are available

for review at any time, even if an error such

as a digital entry error occurs, it is more

likely that the error would be almost

immediately detected by recognition that the

displayed value is not the correct one.

Failure analysis of the current system

assumes certain rates of human error. The

rates for the replacement system will be

lower, and hence are bounded by the USAR

analysis. The NUMAC–PRNM system has

been approved as an acceptable neutron

monitoring replacement by the NRC.

Therefore, based on the above discussions,

the proposed change will not result in a

significant increase in the consequences of

any accident previously evaluated.

The operation of Nine Mile Point Unit 2,

in accordance with the proposed

amendment, will not create the possibility of

a new or different kind of accident from any

accident previously evaluated.

NMPC proposes to replace the existing RPS

APRM system with the NUMAC–PRNM

system and make associated changes to the

RPS and Control Rod Block TS

instrumentation sections. As discussed in

NEDC–3241OP–A, no new system level

failure modes are created with the

replacement system. The NUMAC–PRNM

modification and associated changes to the

TSs involve systems that are intended to

detect the symptoms of certain events or

accidents and initiate mitigating actions. The

worst case failure of the systems involved

would be a failure to initiate mitigative

actions (i.e., scram or rod block), but no

failure can cause an accident. This is

unchanged from the current system. The

proposed changes do not modify the basic

functional requirements of the affected

equipment, create any new system interfaces

or interactions nor create any new system

failure modes or sequence of events that

could lead to an accident. The replacement

system is more tolerant of degraded power

than the current system. Software common

cause failures can at most cause the system

to fail to perform its safety function. As with

system level failures, software failures could

fail to initiate actions to mitigate the

consequences of an accident, but would not

cause one. Surveillance testing will continue

to be performed to assure reliability and

maintain current performance levels.

The NUMAC–PRNM system is a digital

system with software (firmware) control. As

such, it has ‘‘central’’ processing points and

software controlled digital processing where

the current system has analog and discrete

component processing. The result is that the

specific failures of hardware and potentially

common cause software are different from

the current system. Also, automatic self-test

results in some cases in a direct trip as a

result of a hardware failure where the current

system may have remained ‘‘as is.’’ However,

when these are evaluated at the system level,

there are no new effects. In general, the

USAR assumes simplistic failure modes

(relays for example) but does not specifically

evaluate effects added by the NUMAC–

PRNM such as self-test detection and

automatic trip or alarm. The effects of

software common cause failures are mitigated

by hardware design and system architecture.

The replacement system is fully qualified to

operate in its installed location and will not

affect other equipment. The NUMAC–PRNM

system has been approved as an acceptable

neutron monitoring replacement by the NRC.

Therefore, the proposed change will not

create the possibility of a new or different

kind of accident from any previously

evaluated.

The operation of Nine Mile Point Unit 2,

in accordance with the proposed

amendment, will not involve a significant

reduction in a margin of safety.

The proposed modification and associated

TS changes will not adversely affect the

performance characteristics of the RPS and

Control Rod Block instrumentation nor will

it affect the ability of the subject

instrumentation to perform its intended

function. As stated in NEDC–3241OP–A, the

replacement system has improved channel

trip accuracy compared to the current system

and meets or exceeds system requirements

assumed in setpoint analysis. Also, the

channel response time is within acceptable

limits, the channel indicated accuracy is

improved over the current system, and the

replacement system does not cause a plant

parameter for any analyzed event to fall

outside of acceptable limits. The surveillance

testing and frequencies proposed will assure

reliability of the RPS and Control Rod Block

instrumentation. In addition, the subject

equipment was qualified, where appropriate,

to assure its intended safety function is

performed. Therefore, the proposed changes

do not involve 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 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: Reference and Documents

Department, Penfield Library, State

University of New York, Oswego, New

York 13126.

Attorney for licensee: Mark J.

Wetterhahn, Esquire, Winston & Strawn,

1400 L Street, NW., Washington, DC

20005–3502.

NRC Project Director: S. Singh Bajwa.

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 amendment requests: July 30,

1997.

Description of amendment requests:

The proposed amendments would

revise the combined Technical

Specifications (TS) for the Diablo

Canyon Power Plant Unit Nos. 1 and 2

to add a limiting condition for operation

and surveillance requirements for a

residual heat removal (RHR) pump trip

on low refueling water storage tank

(RWST) level to TS 3/4.3.2, ‘‘Engineered

Safety Features Actuation System

Instrumentation.’’

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 change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

This change assures the availability of the

refueling water storage tank (RWST) lowlevel trip of the residual heat removal (RHR)

pumps by establishing limits on the time that

a channel can be out of service to 72 hours

and establishing surveillance criteria to

verify the operation of the logic. The RHR

system is used to respond to loss of coolant

accidents (LOCAs) and other (e.g., secondary

side) accidents that could result in initiation

of a safety injection signal, and is not a

precursor to any of these events as evaluated

in safety analyses. Under accident conditions

the RWST serves as the source of water for

the emergency core cooling system (ECCS)

pumps and the containment spray pumps.

The RWST and the RHR pump trip are

accident mitigation components and are not

precursors for any accident evaluated in the

safety analyses.

The existing Technical Specification (TS)

would allow one RWST level indication

channel to be inoperable indefinitely, and

has an allowed outage time (AOT) for two

channels inoperable of up to seven days.

Additionally, the existing TS does not apply

to the RWST low-level RHR pump trip logic.

The new TS provides controls that require

that all three RWST low-level trip channels

be maintained operable while the plant is in

Modes 1 to 4, and provides for an AOT for

one channel inoperable for up to 72 hours,

if the inoperable channel is placed in the cutout mode within 6 hours. By placing the

inoperable channel in the cut-out mode, the

possibility of a channel failure causing an

RHR pump failure to start at the onset of an

accident is precluded even with a single

active failure. This assures that the

consequences of an accident are not

increased.

The change will have no affect on the

probability of a physical failure of an RHR

pump because it only ensures the presence

of a pump trip signal when required.

Therefore, there is no increase in the

probability of failure of an RHR train to

function as designed. This change will have

no affect on the probability of any other

ECCS equipment failure as it only affects the

presence of a trip signal for the RHR pumps.
Federal Register / Vol. 62, No. 250 / Wednesday, December 31, 1997 / Notices 68313

The new TS 3.3.2 item would provide

controls that require that all three RWST

level channels be maintained operable while

the plant is in operating Modes 1 to 4 (power

operation through hot shutdown). By

maintaining the three channels operable, the

RHR pump actuation/trip logic operability is

assured so that the RHR and RWST can in

all cases perform their intended accident

mitigation functions following a design basis

event as evaluated in the safety analyses.

Therefore, the proposed changes do not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The RHR system is used to respond to

LOCAs and other (e.g., secondary side)

accidents that could result in initiation of a

safety injection signal. Under accident

conditions the RWST serves as the initial

source of water for injection by the RHR and

other ECCS pumps, and is the source of water

for the containment spray pumps. This

change does not affect operation of the

systems as it relates to their response to

accident conditions. It provides additional

assurance that the RHR pump trip logic will

operate as designed by establishing

administrative controls on the time the

system is susceptible to a single failure. No

new failure modes have been introduced.

Therefore, the proposed changes do not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. The proposed change does not involve

a significant reduction in a margin of safety.

The relevant margin of safety is based on

the RHR pumps starting and then

automatically stopping at the correct RWST

water level. The new TS 3.3.2 item provides

controls that require all three RWST level

channels be maintained operable while the

plant is in Modes 1 to 4. By maintaining the

three channels operable, the capability of the

RHR pump actuation/trip logic to survive a

single active failure is assured. Therefore, the

trip logic operability is assured and the

margin is preserved. This change also

provides additional assurances that the

remaining water in the RWST at the time of

switchover is consistent with that assumed in

the Final Safety Analysis Report and Safety

Evaluation Reports.

Therefore, the proposed 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 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment requests

involve no significant hazards

consideration.

Local Public Document Room

location: California Polytechnic State

University, Robert E. Kennedy Library,

Government Documents and Maps

Department, San Luis Obispo, California

93407.

Attorney for licensee: Richard F.

Locke, Esq., Pacific Gas and Electric

Company, P.O. Box 7442, San

Francisco, California 94120.

NRC Project Director: William H.

Bateman.

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 amendment request: June 25,

1997.

Description of amendment request:

The amendments would modify the

Susquehanna Steam Electric Station,

Units 1 and 2 Technical Specifications

to reflect an increase in the secondary

containment bypass leakage.

Specifically, Section 3.6.1.2 is changed

to replace the leakage of 1.2 scf per hour

for any one main steam line drain with

25.43 scfh for secondary containment

bypass leakage from all sources; Section

3.6.1.2 is changed to include the Main

Steam Line Drain, high-pressure coolant

injection (HPCI) system drain, and

reactor core isolation cooling (RCIC)

system drain leakages as part of the 300

scfh leakage requirement; and Section 3/

4.6.1.2 is changed to include a

discussion which related the secondary

containment bypass leakage TS to the

radiological dose analyses.

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 change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

Of the potential accidents described in

FSAR [Final Safety Analysis Report]

Chapters 6 and 15, only a ‘‘Decrease in

Reactor Coolant Inventory’’ as described in

FSAR Section 15.6.5 is affected by the

proposed action. The specific accident of

concern is a design basis LOCA [loss-ofcoolant accident] concurrent with a LOOP

[loss-of-offsite power] which results in RPV

[reactor pressure vessel] depressurization and

failure to recover RPV level above the FW

[feedwater] spargers. For this accident, the

current licensing basis offsite and control

room dose analyses assume a secondary

containment bypass leakage rate of 9 scfh and

primary containment water (called ESF

[engineered safety function]) leakage of 5

gpm. The current licensing basis analyses do

not attribute this leakage to any specific

pathway.

The proposed action does not increase the

probability of a previously analyzed accident

in any way. The condition of concern is the

result of an accident and as such does not

contribute to the initiation of an accident as

analyzed in the FSAR.

Of concern is whether or not the proposed

action significantly increases the

consequences of an accident as previously

evaluated. Calculations of off-site dose

assuming SCBL [secondary containment

bypass leakage] of 28 scfh, primary

containment water leakage of 20 gpm, and

crediting suppression pool scrubbing show

decreases in thyroid dose, but slight

increases in whole body dose when

compared with dose calculations performed

to support the removal of the MSIV–LCS

[main steam isolation valve-leakage control

system]. This result is expected because the

effect of suppression pool scrubbing is

factored into the revised licensing basis

analysis. Suppression pool scrubbing is

effective in reducing iodine release but has

no assumed effect on the removal of noble

gases. Since the methodology/assumptions

for scrubbing are acceptable to the NRC

[Nuclear Regulatory Commission] per the

guidance in SRP [Standard Review Plan]

Section 6.5.5 and the values for

decontamination factors are conservative, the

judgment may be made that considerable

margin is preserved within the analysis.

Although the whole body dose with SCBL

of 28 scfh and water leakage of 20 gpm is

increased from the previously approved

MSIV-LCS dose analysis, the increase is

small (about 1 rem at the two hour site

boundary; less than 0.1 rem 30 day LPZ [low

population zone]). The total dose including

the increase is still well below the 10CFR100

whole body regulatory limit of 25 rem to

which SSES [Susquehanna Steam Electric

Station] was licensed. No change in operating

procedures is anticipated. Calculated post

accident control room thyroid dose decreases

as a result of this change, and the increase

in control room whole body dose is less than

0.05 rem, well below the 10CFR50, Appendix

A, GDC [General Design Criterion] 19 dose

limits outlined in NUREG–0800. Thus, no

appreciable effect on operator response will

occur as a result of this change.

The addition of the HPCI and RCIC Steam

Line Drains to the Tech Spec for MSIV

leakage is being performed as a result of the

modification which eliminated the MSIV

Leakage Control System (MSIV LCS). At the

time this modification was performed, these

lines were not identified as potential SCBL

pathways. However, because leakage from

the HPCI and RCIC drain lines are part of the

same pathway to the condenser which is now

used by the main steam line drains (MSLD)

and included in the Technical Specifications,

they must be combined with the MSIV’s and

MSLD to be less than 300 scfh. This change

only affects the accounting of the various

drain leakages in the valve testing program.

The justification for this change is the same

justification provided in the ITS [Improved

Technical Specification] submittal (PLA–

4488, August 1, 1996) which adds the MSLD

to this Technical Specification. The test

pressure change to allow testing at Pa was

previously proposed in PLA–4502,

September 23, 1996. One additional change

to delete a footnote related to the removal of

the MSIV Leakage Control System is
68314 Federal Register / Vol. 62, No. 250 / Wednesday, December 31, 1997 / Notices

included because this system has been

removed from Susquehanna SES.

Since the increase in SCBL and primary

containment water leakage result in only a

small increase in the doses previously

evaluated by the NRC and the other changes

do not affect the dose analyses, the proposed

change does not result in a significant

increase in the consequences of an accident

previously evaluated.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

Because the FSAR analysis already

assumes SCBL and ESF leakage occur and the

other changes do not affect the type of

accident[s] that are postulated to occur, the

proposed change does not present the

possibility of an accident of a different type.

Additionally, the change in dose analysis

methodology does not create an accident or

malfunction of a different type since it only

involves the analysis of the effects of such

accidents or malfunctions.

Therefore, the proposed change does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. The proposed change does not involve

a significant reduction in a margin of safety.

This question addresses changes in system

parameters only. Dose consequences are

addressed in Section 1 above. The only

Technical Specification dealing with SCBL is

T.S. 3.6.1.2 which requires the leakage from

any one Main Steam Line Drain (MSLD)

Valve to be less than or equal to 1.2 scfh

when tested at Pa (45.0 psig). As noted

earlier, the current licensing basis accident

dose analysis assumes a total of 9 scfh for

bypass leakage and 5 gpm for primary

containment water leakage but does not

attribute them to any particular source. The

proposed action increases the assumed SCBL

from 9 to 28 scfh and water leakage from 5

gpm to 20 gpm. These leakage rates are

insignificant in terms of SGTS [standby gas

treatment system] flows or water loss from

ECCS systems. These leakage rates do not

affect building temperatures or pressures so

that they become closer to acceptance limits.

Likewise, no other system parameter values

become closer to limits as a result of these

changes in leakage. Consequently, the

existing margin of safety between the

licensing basis analysis and system

parameter acceptance limits is not reduced.

The changes to the HPCI, RCIC, and main

steam line drain leakage only affect the

accounting for the various leakages in the

leakage testing program. The deletion of the

footnote is administrative because the MSIV

Leakage Control System has been removed

from the Susquehanna SES. The change in

test pressure was previously evaluated in

PLA–4502, September 23, 1996. Thus, no

decrease in margin of safety results.

The NRC staff has reviewed the

licensee’s analysis and notes that a

discussion of the administrative change

to delete a footnote in Section 3.6.1.2 is

in the third section of the no significant

hazards consideration. The staff finds

that this administrative change also

does not involve a significant increase

in the probability or consequences of an

accident previously evaluated and does

not create the possibility of a new or

different kind of accident from any

accident previously evaluated. Based on

this staff 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: Osterhout Free Library,

Reference Department, 71 South

Franklin Street, Wilkes-Barre, PA 18701.

Attorney for licensee: Jay Silberg,

Esquire, Shaw, Pittman, Potts and

Trowbridge, 2300 N Street NW.,

Washington, DC 20037.

NRC Project Director: John F. Stolz.

Pennsylvania Power and Light

Company

[Docket No. 50–387]

Susquehanna Steam Electric Station,

Unit 1, Luzerne County, Pennsylvania

Date of amendment request: August

26, 1997.

Description of amendment request:

The amendment would modify the

Susquehanna Steam Electric Station,

Unit 1 Technical Specifications to

change the definitions in Section 1.0 to

make them applicable to ATRIUM–10

fuel (reflecting the new design), to

include the Unit 1 Cycle 11 flow

dependent minimum critical power

ratio (MCPR) Safety Limits in Sections

2.1.2 and 3.4.1.1.2, to change Section

5.3.1 to reflect the ATRIUM–10 design,

and to include Siemens Power

Corporation methodology topical

reports and references to the

methodology in Section 6.9.3.2.

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 change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The applicable sections of the FSAR [Final

Safety Analysis Report] are Chapters 5, 6.3,

9, and 15 of the FSAR. Chapter 5 discusses

the results of the ASME [American Society of

Mechanical Engineers] overpressure analysis

for the reactor pressure boundary. Chapter

6.3 discusses the LOCA [loss-of-coolant

accident]. Chapter 9 discusses fuel storage

and handling. Chapter 15 describes the

transient and accident analyses, a majority of

which have been dispositioned to be nonlimiting. A discussion of the impact of the

Technical Specification changes is provided

below.

The change to Definitions 1.2 and 1.3

makes the definitions applicable to

ATRIUM TM–10. There are no effects on

safety functions from this change.

A cycle specific MCPR Safety Limit

analysis was performed for PP&L

[Pennsylvania Power and Light Company] by

SPC [Siemien Power Corporation]. This

analysis used NRC [Nuclear Regulatory

Commission] approved methods described in

Technical Specification Reference 13 (ANF–

524(P)(A), Revision 2 and Supplement 1

Revision 2), as modified by EMF–97–010(P),

Rev. 1. The SAFETY LIMIT MCPR

calculation statistically combines

uncertainties on feedwater flow, feedwater

temperature, core flow, core pressure, core

power distribution, and the uncertainty in

the Critical Power Correlation. The SPC

analysis used cycle specific power

distributions and calculated MCPR values

such that at least 99.9% of the fuel rods are

expected to avoid boiling transition during

normal operation or anticipated operational

occurrences. The SAFETY LIMIT MCPRs are

specified as a function of core flow. The

resulting two-loop and single-loop values

(Technical Specification Sections 2.1.2 and

3.4.1.1.2) are included in the proposed

change. Thus, the cladding integrity and its

ability to contain fission products are not

adversely affected.

The MCPR methodology for ATRIUM TM–

10 fuel (SPC report EMF–97–010(P), Rev. 1),

included in the revised Technical

Specifications via reference (Section 6.9.3.2)

and previously approved by the NRC for Unit

2 Cycle 9, describes conservative methods for

developing the MCPR Safety Limits and

Operating Limits for the U1C11 reload of

ATRIUM TM–10 fuel in the Susquehanna

Steam Electric Station. This methodology

conservatively accounts for a flow

dependence in the ATRIUM TM–10 critical

power test data as well as an increased

correlation uncertainty for high local peaking

factor rods. The results of using this

methodology are core flow dependent MCPR

Safety Limits plus conservative MCPR

Operating Limits for Unit 1 Cycle 11. The

resulting MCPR Safety Limits and Operating

Limits will continue to assure that at least

99.9% of the fuel rods are expected to avoid

boiling transition during normal operation or

anticipated operational occurrences. Thus,

the cladding integrity and its ability to

contain fission products are not adversely

affected. The proposed change in MCPR

methodology does not physically affect the

plant or its systems.

Using the approach discussed in EMF–97–

010(P), Rev. 1, analyses of the Pump Seizure

accident with the new MCPR methodology

(SPC report EMF–97–010(P), Rev. 1) will

demonstrate that the NRC acceptance

criterion (i.e., small fraction of 10CFR100

dose limits) is met.

The change to the Design Features (Section

5.3) increases the maximum allowable lattice

average enrichment. Analyses have

demonstrated that the ATRIUM TM–10 fuel

will remain subcritical (k-effective < 0.95) in

both the spent fuel pool and the new fuel

vault. Thus, the change to maximum

allowable lattice average enrichment has no

impact on safety functions. The description
Federal Register / Vol. 62, No. 250 / Wednesday, December 31, 1997 / Notices 68315

of a fuel assembly (Section 5.3) is also

revised to reflect the ATRIUM TM–10 central

water channel, and reference to an active fuel

length of 150 inches was deleted. This

change reflects the physical characteristics of

the ATRIUM TM–10 fuel and has no impact

on the probability or consequences of an

event.

Included in the revised Technical

Specifications via reference (Section 6.9.3.2)

are additional NRC approved methodology

reports. The NRC approved topical reports

contain methodology which is used to assure

safe operation of Unit 1 with ATRIUM TM–10

fuel. These methodologies assure that the

core meets appropriate margins of safety for

all expected plant operational conditions

ranging from refueling and cold shutdown of

the reactor through power operation. Thus,

the results obtained from the analyses will

provide assurance that the reactor will

perform its design safety function during

normal operation and design basis events.

The BASES changes for Section 2.1.1

(THERMAL POWER, Low Pressure or Low

Flow) reflect that the Safety Limit is valid for

both 9x9–2 and ATRIUM TM–10. BASES for

Section 2.1.2 were changed to refer to Section

6.9.3.2 for applicable references.

Therefore, the proposed action does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The changes to the Unit 1 Technical

Specifications (Definitions, MCPR safety

limits, Design Features, and inclusion of

methodology references) to allow use of

ATRIUM TM–10 fuel do not require any

physical plant modifications, physically

affect any plant components, or entail

significant changes in plant operation. Thus,

the proposed change does not create the

possibility of a previously unevaluated

operator error or a new single failure. The

consequences of transients and accidents will

remain within the criteria approved by the

NRC. Therefore, the proposed change does

not create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. The proposed change does not involve

a significant reduction in a margin of safety.

The applicable Technical Specification

Sections include 1.0, 2.0, 3/4.4, 5.3, and

6.9.3.2.

The changes to the Unit 1 Technical

Specifications discussed in Item 1 above do

not require any physical plant modifications,

physically affect any plant components, or

entail significant changes in plant operation.

Therefore, the proposed change will not

jeopardize or degrade the function or

operation of any plant system or component

governed by Technical Specifications. The

consequences of transients and accidents will

remain within the criteria approved by the

NRC. The proposed MCPR Safety Limits and

the NRC approved methods and revised

MCPR methodology detailed in the

references added to Section 6.9.3.2 maintain

an equivalent margin of safety as defined in

the BASES of the applicable Technical

Specification sections.

Therefore, the proposed 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: Osterhout Free Library,

Reference Department, 71 South

Franklin Street, Wilkes-Barre, PA 18701.

Attorney for licensee: Jay Silberg,

Esquire, Shaw, Pittman, Potts and

Trowbridge, 2300 N Street NW.,

Washington, DC 20037.

NRC Project Director: John F. Stolz.

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 amendment requests: June 18,

1997.

Description of amendment requests:

The licensee proposes to revise

Technical Specification (TS) 3.8.1, ‘‘AC

Sources—Operating’’ and applicable

Bases. This change will more clearly

reflect safety analysis and testing

conditions.

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 change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed change would revise

Technical Specification (TS) TS 3.8.1, ‘‘AC

Sources—Operating,’’ Surveillance

Requirement (SRs) 3.8.1.1, 3.8.1.2, 3.8.1.7,

3.8.1.10, 3.8.1.11, 3.8.1.12, 3.8.1.13, 3.8.1.14,

3.8.1.15, 3.8.1.16, 3.8.1.17, 3.8.1.19, and

3.8.1.20 and applicable Bases to more clearly

reflect surveillance test conditions and

system design requirements. Changes to the

SRs include more restrictive voltage and

frequency acceptability limits. The new

requirements reflect the system design

requirements in order to ensure Class 1E

system operability, meet the requirements of

the safety analysis, and to agree with the

existing test surveillances.

In addition, the discussion regarding

design basis reactive power loading is

eliminated since this cannot be readily

controlled during testing.

Operation of the facility would remain

unchanged as a result of the proposed change

and no assumptions or results of any

accident analyses are affected. Therefore, the

proposed change will not involve a

significant increase in the probability or

consequences of any accident previously

evaluated.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed change would revise

Technical Specification (TS) TS 3.8.1, ‘‘AC

Sources—Operating,’’ Surveillance

Requirement (SRs) 3.8.1.1, 3.8.1.2, 3.8.1.7,

3.8.1.10, 3.8.1.11, 3.8.1.12, 3.8.1.13, 3.8.1.14,

3.8.1.15, 3.8.1.16, 3.8.1.17, 3.8.1.19, and

3.8.1.20 and applicable Bases to more clearly

reflect surveillance test conditions and

system design requirements.

Operation of the facility would remain

unchanged as a result of the proposed

change. Therefore, the proposed change will

not create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. The proposed change does not involve

a significant reduction in a margin of safety.

The proposed change would revise

Technical Specification (TS) TS 3.8.1, ‘‘AC

Sources—Operating,’’ Surveillance

Requirement (SRs) 3.8.1.1, 3.8.1.2, 3.8.1.7,

3.8.1.10, 3.8.1.11, 3.8.1.12, 3.8.1.13, 3.8.1.14,

3.8.1.15, 3.8.1.16, 3.8.1.17, 3.8.1.19, and

3.8.1.20 and applicable Bases to more clearly

reflect surveillance test conditions and

system design requirements. Changes to the

SRs include more restrictive voltage and

frequency acceptability limits. The new

requirements reflect the system design

requirements in order to ensure Class 1E

system operability, meet the requirements of

the safety analysis, and to agree with the

existing test surveillances.

Therefore, the proposed change will 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 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment requests

involve no significant hazards

consideration.

Local Public Document Room

location: Main Library, University of

California, P. O. Box 19557, Irvine,

California 92713.

Attorney for licensee: T. E. Oubre,

Esquire, Southern California Edison

Company, P. O. Box 800, Rosemead,

California 91770.

NRC Project Director: William H.

Bateman.

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 amendment requests:

November 14, 1997 (supersedes

February 1, 1994, amendment request).
68316 Federal Register / Vol. 62, No. 250 / Wednesday, December 31, 1997 / Notices

Description of amendment requests:

The licensee proposes to revise the

licensing basis as described in the

Updated Final Safety Analysis Report

Section 3.5, ‘‘Missile Protection,’’ to

allow the use of NUREG–0800,

‘‘Standard Review Plan’’ methodology

in evaluating tornado-generated

missiles. In particular, a probability

based criteria is proposed to evaluate

missile barrier requirements consistent

with Section 3.5.1.4 of NUREG–0800.

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 change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

NUREG–0800, Standard Review Plan (SRP)

Section 3.5.1.4, Revision 0 and Section

3.5.1.5 Revision 1 provide a conservatively

acceptable probability threshold for safety

due to damage caused by postulated missile

strikes. Section 3.5.1.4, Revision 0 uses 10–7

per year for a tornado-generated missile

strike, and Section 3.5.1.5 Revision 1 uses

10–7 per year for exceeding 10 CFR Part 100

limits.

The proposed criteria of probability of

damage to critical exposed equipment (as

defined in San Onofre Updated Final Safety

Analysis Report proposed Table 3.5–13) of

10¥7 per year per unit is consistent with this

guidance.

The probability of damage to exposed

critical components due to a postulated

missile strike of 10¥7 is so small as to be

negligible. Therefore, this change does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

This amendment request establishes a

conservative criteria for tornado-generated

missiles consistent with the SRP guidance

and will not create a new or different kind

of accident from any accident that has been

previously evaluated.

3. The proposed change does not involve

a significant reduction in a margin of safety.

This proposed change is consistent with

the methodology and acceptance criteria of

the SRP, and the SRP criteria ensures that

there will be no undue risk to the health and

safety of the public. Therefore, there will be

no 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 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendment requests

involve no significant hazards

consideration.

Local Public Document Room

location: Main Library, University of

California, P. O. Box 19557, Irvine,

California 92713.

Attorney for licensee: T.E. Oubre,

Esquire, Southern California Edison

Company, P. O. Box 800, Rosemead,

California 91770.

NRC Project Director: William H.

Bateman.

Southern Nuclear Operating Company,

Inc., Georgia Power Company,

Oglethorpe Power Corporation,

Municipal Electric Authority of

Georgia, City of Dalton, Georgia

Vogtle Electric Generating Plant, Units 1

and 2, Burke County, Georgia

[Docket Nos. 50–424 and 50–425]

Date of amendments request: August

8, 1997, as supplemented October 10,

1997. This application and supplement

supersedes the October 4, 1996,

application, noticed in the Federal

Register on November 19, 1996 (61 FR

58903), in its entirety.

Description of amendments request:

The proposed amendments would

change the Technical Specifications to

credit soluble boron in the spent fuel

pool for maintenance of subcriticality

and increase the allowable fuel

enrichment to 5.0 percent U–235 as

follows:

1. Revisions to the Table of Contents

The Table of Contents would be revised to

include two additional Technical

Specifications 3.7.17, ‘‘Fuel Storage Pool

Boron Concentration,’’ and 3.7.18, ‘‘Fuel

Assembly Storage in the Fuel Storage Pool’’

and add Figures 3.7.18–1, 3.7.18–2, and

4.3.1–1 through 4.3.1–9 describing burnup

credit, checkerboard configurations and

interface requirements. These changes would

be added to support crediting soluble boron

in the fuel storage pool criticality analyses.

2. Addition of Technical Specifications

3.7.17 and 3.7.18

Technical Specifications 3.7.17, ‘‘Fuel

Storage Pool Boron Concentration,’’ and

3.7.18, ‘‘Fuel Assembly Storage in the Fuel

Storage Pool,’’ would be added to credit

soluble boron in the fuel storage pool

criticality analyses, and specify acceptable

enrichment-burnup combinations for storage

of fuel in the fuel storage pool.

3. Revision to Technical Specification 4.3.1.1

Design Features Section 4.3.1.1 would be

revised to reflect the increased maximum

enrichment assumed in the fuel storage pool

criticality analyses, add a requirement to

maintain Keff less than 1.0 when fully

flooded with unborated water, change the

0.95 Keff requirement from ‘‘if fully flooded

with unborated water’’ to ‘‘when fully

flooded with water borated to 450 ppm (Unit

1) or 500 ppm (Unit 2),’’ and to add a

reference to Specification 3.7.18 for

allowable enrichment-burnup combinations.

Requirements for fuel that do not meet the

requirements of Specification 3.7.18, would

also be added to Section 4.3.1.1, including

Figures 4.3.1–1 through 4.3.1–9 depicting

acceptable enrichment-burnup requirements

and checkerboard configurations.

4. Revisions to the Table of Contents (Bases)

The Table of Contents would be revised to

include two additional Technical

Specification Bases Sections B 3.7.17 ‘‘Fuel

Storage Pool Boron Concentration’’ and B

3.7.18 ‘‘Fuel Assembly Storage in the Fuel

Storage Pool.’’

5. Addition of Bases for Technical

Specifications 3.7.17 and 3.7.18

Two additional Technical Specification

Bases Sections B 3.7.17, ‘‘Fuel Storage Pool

Boron Concentration’’ and B 3.7.18, ‘‘Fuel

Assembly Storage in the Fuel Storage Pool’’

would be added to credit soluble boron in the

fuel storage pool criticality analyses.

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 change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The radiological consequences of 5.0

weight percent U–235 fuel on accidents

previously evaluated in the Vogtle FSAR

[Final Safety Analysis Report] are not

significant. Increasing the enrichment up to

and including 5.0 weight percent U–235 has

minor effects on the radiological source terms

and subsequently the potential releases both

normal and accidental are not significantly

affected. Evaluations performed (WCAP–

12610–P–A, Reference 5 [of the licensee’s

application]) considered the source term, gap

fraction, and the accident doses for a

maximum fuel enrichment of 5.0 weight

percent U–235. It was concluded that

operating with and storing fuel with 5.0

weight percent U-235 enrichment may result

in minor changes in the normal annual

releases of long half-life fission products that

are not significant. Also, the radiological

consequences of accidents are minimally

affected due to the very small changes in the

core inventory and the fact that the currently

assumed gap fractions remain bounding.

The use of the slightly higher enrichment

for VEGP [Vogtle Electric Generating Plant]

fuel will not result in burnups in excess of

those currently allowed for VEGP. The cycle

design methods and limits will remain the

same as are currently licensed. Therefore, the

use of fuel with the higher enrichment will

not result in conditions outside those

currently allowed for VEGP.

There is no increase in the probability of

a fuel assembly drop accident in the fuel

storage pool when considering the presence

of soluble boron in the pool water for

criticality control. The handling of the fuel

assemblies in the fuel storage pool has

always been performed in borated water.

Fuel assembly placement will be

controlled pursuant to approved fuel
Federal Register / Vol. 62, No. 250 / Wednesday, December 31, 1997 / Notices 68317

handling procedures and will be in

accordance with the spent fuel rack storage

configuration limitations in the Technical

Specifications. The consequences of a

misplaced assembly have been included in

the analysis supporting this revision to the

Technical Specifications.

There is no increase in the consequences

of the accidental misloading of a fuel

assembly into the fuel storage pool racks

because criticality analyses demonstrate that

the pool will remain subcritical following an

accidental misloading of an assembly. There

are no credible dilution events that reduce

the subcriticality margin below the 5%

margin recommended in NRC guidance

(references 1, 2, and 3 [of the licensee’s

application]). Even if the fuel storage pool

were diluted to a boron concentration of 0

ppm the No Soluble Boron 95/95 analysis

demonstrates that the pool will remain

subcritical. The proposed Technical

Specifications limitations will ensure that an

adequate fuel storage pool boron

concentration will be maintained.

There is no increase in the probability of

the loss of normal cooling to the fuel storage

pool water due to the presence of soluble

boron in the pool water for subcriticality

control, because a concentration of soluble

boron similar to the proposed limit has

always been maintained in the fuel storage

pool water.

The loss of normal cooling to the fuel

storage pool will cause an increase in the

temperature of the fuel storage pool water.

This will cause a decrease in water density

which would normally result in an addition

of negative reactivity. However, since

Boraflex is not considered to be present, and

the fuel storage pool water has a high

concentration of boron, a density decrease

causes a positive reactivity addition. The

amount of soluble boron required to offset

this postulated accident was evaluated for

the allowed storage configurations. The

amount of soluble boron necessary to

mitigate these accidents and ensure that the

Keff will be maintained less than or equal to

0.95 has been included in the fuel storage

pool boron concentration. Because adequate

soluble boron will be maintained in the pool

water, the consequences of a loss of normal

cooling to the fuel storage pool will not be

increased.

Therefore, based on the conclusions of the

above analysis, the proposed changes will

not involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

analyzed.

The potential for criticality accidents in the

fuel storage pool are not new or different

types of concerns. The potential criticality

accidents have been reanalyzed in the

Criticality Analysis report (Enclosure 5 [of

the licensee’s application]) to demonstrate

that the pool remains subcritical.

Soluble boron has been maintained in the

fuel storage pool water since its initial

operation. The possibility of a fuel storage

pool dilution is not affected by the proposed

change to the Technical Specifications.

Therefore, the implementation of Technical

Specification controls for the soluble boron

will not create the possibility of a new or

different kind of accidental pool dilution.

With credit for soluble boron now a major

factor in controlling subcriticality, an

evaluation of fuel storage pool dilution

events was completed. The results of the

evaluation concluded that no credible events

would result in a reduction of the criticality

margin below the 5% margin recommended

by the NRC. In addition, the No Soluble

Boron 95/95 criticality analysis assures that

dilution to 0 ppm will not result in

criticality.

Proposed Technical Specifications 3.7.17,

3.7.18 and 4.3.1.1 which ensure the

maintenance of the fuel storage pool boron

concentration and storage configuration, do

not represent new concepts. The actual boron

concentration in the fuel storage pool has

been maintained at a higher value than the

proposed limits for the Unit 1 and 2 fuel

storage pools for refueling purposes. The

criticality analysis (Enclosure 5 [of the

licensee’s application]) determined that a

boron concentration of 450 ppm (Unit 1) and

500 ppm (Unit 2) results in a Keff [less than

or equal to] 0.95.

There is no significant change in plant

configuration, equipment design, or usage of

plant equipment. The safety analysis for

dilution accidents has been expanded;

however, the criticality analyses assure that

the pool will remain subcritical with no

credit for soluble boron. Therefore, the

proposed changes will not create the

possibility of a new or different kind of

accident.

3. The proposed change does not result in

a significant reduction in the margin of

safety.

Proposed Technical Specifications 3.7.17,

3.7.18, and 4.3.1.1 and the associated fuel

storage pool boron concentration and storage

requirements will provide adequate margin

to assure that the fuel storage array will

always remain subcritical by the 5% margin

recommended by the NRC. Those limits are

based on the criticality analysis (Enclosure 5

[of the licensee’s application]) performed in

accordance with the Westinghouse fuel

storage rack criticality analysis methodology

described in Reference 4 [of the licensee’s

application].

While the criticality analysis utilized credit

for soluble boron, the storage configurations

have been defined using Keff calculations to

ensure that the spent fuel rack Keff will be

less than 1.0 with no soluble boron.

Soluble boron credit is used to offset offnormal conditions (such as a misplaced

assembly) and to provide subcritical margin

such that the fuel storage pool Keff is

maintained less than or equal to 0.95.

The combination of the No Soluble Boron

95/95 Keff calculation which shows that the

Keff will remain less than 1.0 when flooded

with unborated water and the unavailability

of the large volumes of water which are

necessary to dilute the fuel storage pool to a

Keff of > 0.95, provide a level of safety

comparable to the conservative criticality

analysis methodology required by References

1, 2, and 3 [of the licensee’s application].

Therefore, the proposed changes in this

license amendment will not result in a

significant reduction in the plant’s 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: Burke County Public Library,

412 Fourth Street, Waynesboro, Georgia.

Attorney for licensee: Mr. Arthur H.

Domby, Troutman Sanders,

NationsBank Plaza, Suite 5200, 600

Peachtree Street, NE., Atlanta, Georgia.

NRC Project Director: Herbert N.

Berkow.

Southern Nuclear Operating Company,

Inc., Georgia Power Company,

Oglethorpe Power Corporation,

Municipal Electric Authority of

Georgia, City of Dalton, Georgia

Vogtle Electric Generating Plant, Units 1

and 2, Burke County, Georgia

[Docket Nos. 50–424 and 50–425]

Date of amendment request:

September 4, 1997, as supplemented

November 20, 1997.

Description of amendment request:

The proposed amendments would

revise the Technical Specifications (TS)

to change the capacity of the Vogtle Unit

1 spent fuel storage pool from 288 to

1476 assemblies, and would revise the

design features description to reflect the

criticality analyses and storage cell

spacing. Specifically, the changes would

be as follows:

1. Figure 3.7.18–1 would be replaced with

a revised figure based on the criticality

analyses for the Unit 1 racks containing

boral.

2. The criticality information for Unit 2

would be placed unchanged into Section

4.3.1.2, and Section 4.3.1.1. would be revised

to address Unit 1.

3. Design Features Section 4.3.1.1.c would

be revised to indicate 600 ppm as the

required amount of soluble born to maintain

Keff less than or equal to 0.95.

4. Design Features Section 4.3.1.1.d would

be revised to include the reference Keff that

is equivalent to the combination of burnup

and initial enrichment defined by Figure

3.7.18–1.

5. Design Features Section 4.3.1.1.e would

be revised to indicate that fuel assemblies

with up to 5 weight percent U–235 may be

stored in 3-out-of-4 checkerboard storage

configurations; delete Figure 4.3.1–1;

eliminate the reference to 2-out-of-4 storage

for the Unit 1 pool and include the reference

K acceptable for all cell storage in the Unit

1 fuel storage racks.

6. Design Features Section 4.3.1.1.f would

be revised to include the pitch of the Unit 1

fuel storage racks.
68318 Federal Register / Vol. 62, No. 250 / Wednesday, December 31, 1997 / Notices

7. Design Features Section 4.3.3 would be

revised to indicate the Unit 1 fuel storage

pool capacity of 1476 fuel assemblies.

8. The titles on Figures 4.3.1–4, 4.3.1–6,

and 4.3.1–7 would be revised to reflect the

elimination of 2-out-of-4 storage

configuration requirements for the Unit 1 fuel

storage pool.

Changes to the TS Bases are also proposed.

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 analyses methodologies are the

same as previously approved for use by the

NRC. The results of the analyses resulted in

fuel pool boron concentrations, and fuel

assembly storage limitations that are similar

to those already submitted to the NRC. The

increased number of fuel assemblies will

remain less than the number previously

accepted by the NRC for storage in VEGP

[Vogtle Electric Generating Plant] Unit 2,

which has a similarly designed and

constructed facility, with the exception of the

number of fuel storage locations.

Therefore, based on the conclusions of the

above analysis, the proposed changes will

not involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. The effects of accidents that could affect

the fuel were analyzed for the fuel storage

racks, however the types of accidents have

not changed. The fuel to be stored in the Unit

1 pool is expected to meet the all cell storage

requirements. The racks will be placed in the

Unit 1 pool without lifting any loads over

spent fuel. After installation of the new racks,

the Unit 1 pool will have 1476 storage

locations which is well within the 2098

locations that the pool and structure is

capable of storing, based on its similarity to

the Unit 2 pool.

Therefore, the proposed changes will not

create the possibility of a new or different

kind of accident.

3. The changes to the technical

specifications are necessary to incorporate

the parameters resulting from the criticality

analyses. The criticality analyses were

performed using methods and criteria

previously accepted by the NRC. The

requirements are similar to the previously

submitted requirements. The margins of

safety provided by the previous technical

specifications are not significantly affected

because the new racks are based on the same

acceptance values. The larger number of fuel

assemblies to be stored in the Unit 1 pool

remains well within the capability of the

pool.

Therefore, the proposed changes in this

license amendment will not result in a

significant reduction in the plant’s 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: Burke County Public Library,

412 Fourth Street, Waynesboro, Georgia.

Attorney for licensee: Mr. Arthur H.

Domby, Troutman Sanders,

NationsBank Plaza, Suite 5200, 600

Peachtree Street, NE., Atlanta, Georgia.

NRC Project Director: Herbert N.

Berkow.

Southern Nuclear Operating Company,

Inc., Georgia Power Company,

Oglethorpe Power Corporation,

Municipal Electric Authority of

Georgia, City of Dalton, Georgia

Vogtle Electric Generating Plant, Units 1

and 2, Burke County, Georgia

[Docket Nos. 50–424 and 50–425]

Date of amendment request:

November 20, 1997.

Description of amendment request:

The proposed amendment would

change the Technical Specifications

(TS) to provide for the following with

regard to the Reactor Trip System (RTS)

and Engineered Safety Feature

Actuation System (ESFAS)

instrumentation trip setpoints:

1. The inequalities as they are applied to

the Trip Setpoint column of Tables 3.3.1–1

and 3.3.2–1 would be deleted, and the

column heading would be changed from

‘‘Trip Setpoint’’ to ‘‘Nominal Trip Setpoint.’’

2. A footnote would be added to the new

‘‘Nominal Trip Setpoint’’ column of Tables

3.3.1–1 and 3.3.2–1 that would allow the trip

setpoints to be set more conservative than the

nominal value as necessary to respond to

plant conditions.

3. The Allowable Value for Table 3.3.1–1,

Function 14.b, Turbine Trip—Turbine Stop

Valve Closure, would be revised from

‘‘[greater than or equal to] 96.7% open’’ to

‘‘[greater than or equal to] 90% open.’’

4. Footnotes l and m of Table 3.3.1–1

would be revised to refer to the ‘‘Nominal

Trip Setpoint’’ and delete the inequalities

applied to the trip setpoints.

5. A superscript ‘‘(a)’’ would be deleted

from the heading of the ‘‘Trip Setpoint’’

column on page 6 of 8 of Table 3.3.1–1.

6. Notes 1 and 2 to Table 3.3.1–1,

Overtemperature ∆T and Overpower ∆T,

respectively, would be revised to refer to the

‘‘Nominal Trip Setpoint.’’ In addition, these

notes will be revised to delete the

inequalities from the values for the constants

K1 through K6 (except for K5 [greater than or

equal to] 0 for decreasing temperature and K6

= 0 for T [less than or equal to] T′′), and for

T′, T′′, and P′.

7. The inequality applied to the ESFAS

Allowable Value for Steam Line Pressure—

Low (Table 3.3.2–1, Function 1.e) would be

changed from ‘‘[less than or equal to]’’ to

‘‘[greater than or equal to].’’

Associated changes to the TS Bases are also

proposed.

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 change involve a

significant increase in the probability or

consequences of an accident previously

evaluated?

No. The proposed changes affect only the

presentation of the trip set points for the RTS

and ESFAS in the VEGP [Vogtle Electric

Generating Plant] Units 1 and 2 TS. The

calibration of the channels whose setpoints

are specified in the TS will continue to be

performed in a manner consistent with the

setpoint methodology described in WCAP–

11269 Rev. 1. There will be no adverse effect

on the ability of those channels to perform

their safety functions as assumed in the

safety analyses. Since there will be no

adverse affect on the trip setpoints or the

instrumentation associated with those trip

setpoints, there will be no increase in the

probability of any accident previously

evaluated. Similarly, since the ability of the

instrumentation to perform its safety function

is not adversely affected, there will [be] no

increase in the consequences of any accident

previously evaluated.

2. Does the proposed change create the

possibility of a new or different kind of

accident from any accident previously

evaluated?

No. The proposed change affects only the

presentation of trip setpoint requirements in

the TS. Plant operation will not be changed,

and the response of safety related equipment

as assumed in the accident analyses will not

be adversely affected. Therefore, the

proposed change does not involve a new or

different kind of accident than any

previously evaluated.

3. Does the proposed change involve a

significant reduction in a margin of safety [?]

No. As described above, the RTS and

ESFAS instrumentation will remain capable

of performing its safety function as assumed

in the accident analyses. The treatment of

trip setpoints as nominal values is consistent

with the methodology used to establish those

setpoints. As such, margin is not affected by

the proposed 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: Burke County Public Library,

412 Fourth Street, Waynesboro, Georgia.

Attorney for licensee: Mr. Arthur H.

Domby, Troutman Sanders,

NationsBank Plaza, Suite 5200, 600

Peachtree Street, NE., Atlanta, Georgia.

NRC Project Director: Herbert N.

Berkow.
Federal Register / Vol. 62, No. 250 / Wednesday, December 31, 1997 / Notices 68319

Vermont Yankee Nuclear Power

Corporation

[Docket No. 50–271]

Vermont Yankee Nuclear Power Station,

Windham County, Vermont

Date of amendment request: October

10, 1997, as supplemented October 31,

1997.

Description of amendment request:

The proposed amendment would revise

the Technical Specifications (TS) to

reflect the installation of a generator noload disconnect to facilitate use of the

main step-up transformer backfeed as

the delayed access offsite power source.

Also, the amendment would revise

existing limiting conditions for

operation and required action

statements for operation with inoperable

ac power sources to be consistent with

current guidance.

Specifically, the changes proposed

are: (1) TS Limiting Conditions for

Operation Section—Normal Operation,

3.10.A.4 (2) TS Limiting Conditions for

Operation Section—Operation with

Inoperable Components, 3.10.B.3, (3) TS

Surveillance Requirements—Normal

Operation, 4.10.A.4, (4) TS Surveillance

Requirements—Operation with

Inoperable Components, Section

4.10.B.3, (5) Bases Section 3.10.A, (6)

Bases Section 3.10.B, (7) Bases Section

4.10.A, and (8) Bases Section 4.10.B

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 amendment will not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed amendment removes credit

for the Vernon Tie, Vermont Yankee’s station

blackout source of power, from the Technical

Specifications and reflects the installation of

the generator no load disconnect as part of

the backfeed. Neither the backfeed through

the main transformers nor the Vernon Tie are

accident initiators; therefore, the change does

not involve a significant increase in the

probability of an accident previously

evaluated. The change does not affect the

capability, availability, maintenance or

operation of the Vernon Tie. Installation of

the generator no load disconnect switch is

being implemented by a design change in

order to enhance plant safety by reducing

time necessary to establish the backfeed

through the main transformer. A separate 10

CFR 50.59 evaluation is being prepared to

document that the modification does not

create an unreviewed safety question.

The proposed amendment also clarifies the

allowable out of service times, and required

actions; and updates surveillance

requirements for the immediate and delayed

access offsite power sources. These changes

do not involve a significant increase in the

probability or consequences of an accident

previously evaluated. Modification of a

technical specification out of service time

and required action cannot affect the

probability or consequences of an accident.

Enhancing surveillance requirements to

provide assurance that the backfeed can be

achieved when required and to provide

assurance that remaining power sources are

available when an offsite source is

unavailable improves plant safety and does

not increase the probability or consequences

of an accident.

Therefore, the change does not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

(2) The proposed amendment will not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

The proposed amendment removes the

Vernon Tie, Vermont Yankee’s station

blackout source of power, as a delayed access

source from the Technical Specifications and

reflects the improvements to the main

transformer backfeed delayed access source

because of installation of the generator no

load disconnect. Neither the removal of the

Vernon Tie from Technical Specifications

nor the improvements to the delayed access

power source (backfeed) can create the

possibility of a new or different kind of

accident from any previously evaluated.

The proposed amendment also clarifies the

allowed outage times, and action statements;

and updates surveillance requirements for

the immediate and delayed access offsite

power sources. A clarification of a technical

specification out of service time and required

action cannot create a new or different kind

of accident from any accident previously

evaluated. Enhancing surveillance

requirements to provide assurance that the

backfeed can be achieved when required and

to provide assurance that remaining power

sources are available when an offsite source

is unavailable improves plant safety and

cannot create a new or different kind of

accident from any accident previously

evaluated.

Therefore, this change would not create the

possibility of a different type of accident than

previously evaluated.

(3) The proposed amendment will not

involve a significant reduction in a margin of

safety.

The proposed amendment removes the

Vernon Tie, Vermont Yankee’s station

blackout source of power, as a delayed source

of offsite power from the Technical

Specifications and reflects the improvements

to the main transformer backfeed delayed

access source because of installation of the

generator no load disconnect. No existing

safety margins are adversely affected. The

backfeed is modified so that it may be

established in sufficient time to ‘‘assure that

specified acceptable fuel design limits and

design conditions of the reactor coolant

pressure boundary are not exceeded’’.

Vernon Tie will not be affected by the

modification and remain available as a

station blackout source; thus this change

does not involve a significant reduction in

the margin of safety.

The proposed amendment also clarifies the

allowed out of service times, and required

actions; and updates surveillance

requirements for the immediate and delayed

access offsite power sources. A clarification

of a technical specification out of service

time and required action does not involve a

significant reduction in the margin of safety

in the Technical Specifications. Enhancing

surveillance requirements to provide

assurance that the backfeed can be achieved

when required and to provide assurance that

remaining power sources are available when

an offsite source is unavailable improves

plant safety and 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

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Brooks Memorial Library, 224

Main Street, Brattleboro, VT 05301.

Attorney for licensee: Mr. David R.

Lewis, Shaw, Pittman, Potts &

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037–1128.

NRC Project Director: Ronald Eaton,

Acting Director.

Vermont Yankee Nuclear Power

Corporation

[Docket No. 50–271]

Vermont Yankee Nuclear Power Station,

Windham County, Vermont

Date of amendment request:

November 20, 1997.

Description of amendment request:

The proposed amendment would revise

the existing requirements for the

Auxiliary Electrical Power Systems as

identified in Technical Specifications

(TSs) 3/4.10.A and TS 3.10.A.2.b. The

specific changes are:

(1) The requirements in TS 3.10.A.2.b.

are revised to omit the allowance for

Spare Charger AB to substitute for either

Charger A or B.

(2) The Bases in TS 3.10.A. are

revised to omit the statements that

justify Spare Charger AB to substitute

for either Charger A or Charger B.

The proposed changes provide more

limiting requirements for operation with

the standby battery charger in service.

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 facility in accordance

with the proposed amendment would not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.
68320 Federal Register / Vol. 62, No. 250 / Wednesday, December 31, 1997 / Notices

Neither batteries, nor their chargers, are

considered to be an initiator of any

previously analyzed accident. Therefore, this

change will not significantly increase the

probability of any previously analyzed

accident.

At least one Battery System is required to

be available to mitigate the consequences of

a Design Basis Accident. This change

removes an allowance which places the unit

in a more vulnerable condition through the

unrestricted use of the spare battery charger.

Since this change limits such a condition, it

maintains the assumptions of the safety

analysis, and therefore, will not significantly

increase the consequences of any previously

analyzed accident.

(2) Operation of the facility in accordance

with the proposed amendment would not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

The proposed change does not necessitate

a physical alteration of the plant (no new or

different type of equipment will be installed)

nor is operation of the currently installed

equipment changed. The change will,

however, limit a currently allowed

configuration with the spare charger and is

more conservative. Thus, this change will not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

(3) Operation of the facility in accordance

with the proposed amendment would not

involve a significant reduction in the margin

of safety.

The proposed change continues to provide

the previous margin of safety regarding the

capability to withstand a single failure. At

least one Battery System will continue to be

available to provide the required safety

function. The change will limit a currently

allowed configuration with the spare charger

and is thus more conservative. Therefore, this

change will not significantly reduce 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: Brooks Memorial Library, 224

Main Street, Brattleboro, VT 05301.

Attorney for licensee: Mr. David R.

Lewis, Shaw, Pittman, Potts &

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037–1128.

NRC Project Director: Ronald Eaton,

Acting Director.

Vermont 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:

November 5, 1997.

Description of amendment request:

The proposed change to Technical

Specifications 5.3 and 5.4 would reflect

an increase in the maximum permitted

fuel enrichment to 4.3 weight percent

U235 from the current 4.1 weight percent

U235. Fuel burnup limits and reactor

operating power level would remain

unchanged.

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 Technical Specifications

changes for Surry Units 1 and 2 against the

criteria of 10 CFR 50.92. It has been

concluded that use of fuel with the slightly

higher initial enrichment does not involve a

significant hazards consideration as defined

in 10 CFR 50.92. An increase in the

maximum initial fuel enrichment from 4.1 to

4.3 weight percent U235 will not:

1. Involve a significant increase in the

probability of occurrence or the

consequences of an accident previously

evaluated. The only accidents for which the

probability of occurrence is potentially

affected by the fuel enrichment involve

criticality events during handling and

storage. Analyses have demonstrated that the

K-effective will be low enough to ensure

subcriticality during both normal operation

and under postulated accident conditions

during the handling and storage of both new

and spent fuel. Therefore, the probability of

occurrence of criticality during fuel handling

or storage is not increased. Safety analyses of

record are based on inputs which bound the

proposed increase in fuel enrichment. Since

no changes to the fuel burnup limits are

requested, the radiological consequences of

previously evaluated accident scenarios will

not be increased. Therefore, neither the

probability of occurrence nor the

consequences of any accident previously

evaluated is significantly increased.

2. Create the possibility for a new or

different type of accident from any accident

previously evaluated. Fuel with the higher

initial enrichment will meet all applicable

design criteria and will operate within

existing Technical Specifications limits.

Adherence to these standards and criteria

precludes new challenges to components and

systems that could introduce a new type of

accident. All design and performance criteria

will continue to be met. In addition, the use

of a slightly higher initial fuel enrichment

does not involve any alteration to plant

equipment or procedures which would

introduce any new or unique operational

modes or accident precursors. Therefore, the

possibility for a new or different kind of

accident from any accident previously

evaluated is not created.

3. Involve a significant reduction in the

margin of safety. Surry Units 1 and 2 will

continue to operate in compliance with the

Technical Specifications, ensuring that the

plants continue to provide acceptable levels

of protection for the health and safety of the

public. The Technical Specifications are

based upon assumption[s] made in the safety

and accident analyses, including those

relating to the fuel enrichment and the design

of the fuel storage areas. Analyses have

demonstrated that subcriticality will be

ensured during fuel storage and handling

accident scenarios for both new and spent

fuel. Additionally, safety analyses of record

for core operation will remain applicable for

Surry Unit 1 and 2 cores which use fuel with

the slightly higher U235 enrichment.

Therefore, the regulated margin of safety as

defined in the Bases to the Surry Technical

Specifications is not reduced.

Based on the preceding information, it has

been determined that the use of fuel with an

initial enrichment of up to 4.3 weight percent

U235 satisfies the no significant hazards

consideration criteria of 10 CFR 50.92.

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: Swern 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: James E. Lyons.

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
Federal Register / Vol. 62, No. 250 / Wednesday, December 31, 1997 / Notices 68321

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. 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 application for amendments:

June 30, 1997, as supplemented on

September 25, 1997.

Brief description of amendments: The

amendments grant partial credit for

boron in the spent fuel pools to

maintain the subcriticality.

Date of issuance: December 4, 1997.

Effective date: Immediately, to be

implemented within 30 days.

Amendment Nos.: 94, 94, 86 and 86.

Facility Operating License Nos. NPF–

37, NPF–66, NPF–72 and NPF–77: The

amendments revised the Technical

Specifications.

Date of initial notice in Federal

Register: October 22, 1997 (62 FR

54868).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated December 4,

1997.

No significant hazards consideration

comments received: No.

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.

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 application for amendments:

March 14, 1997, as supplemented. July

29, 1997, and August 13, 1997. The July

29, 1997, and August 13, 1997, letters

provided clarifying information that did

not change the initial proposed no

significant hazards consideration

determination or expand the

amendment request beyond the scope of

the May 7, 1997, Federal Register

notice.

Brief description of amendments:

These amendments relocate certain

administrative control Technical

Specifications (TSs) from the Beaver

Valley Power Station, Unit Nos. 1 and

2 (BVPS–1 and BVPS–2), TSs to the

licensee’s operational quality assurance

program description, which is presented

in Section 17.2 of the BVPS–2 Updated

Final Safety Analysis Report (UFSAR).

Section 17.2 of the BVPS–2 UFSAR

contains the quality assurance program

description for both BVPS–1 and BVPS–

2. The following TSs are being relocated

to the quality assurance program

description.

BVPS–2 TS 6.2.3 (Independent Safety

Evaluation Group)

BVPS–1 and BVPS–2 TS 6.5.1 (Onsite

Safety Committee)

BVPS–1 and BVPS–2 TS 6.5.2 (Offsite

Review Committee)

BVPS–1 and BVPS–2 TS 6.8.2

(Procedures, Review and)

BVPS–1 and BVPS–2 TS 6.8.3

(Temporary Procedure Changes,

Review and Approval)

BVPS–1 and BVPS–2 TS 6.10.1 (Records

Retention, At least 5 Years)

BVPS–1 and BVPS–2 TS 6.10.2 (Records

Retention, Duration of Operating

License)

Date of issuance: December 10, 1997.

Effective date: Both units, as of date

of issuance, to be implemented within

60 days.

Amendment Nos.: 209 and 87.

Facility Operating License Nos. DPR–

66 and NPF–73: Amendments revised

the Technical Specifications, and

Appendix C of the License.

Date of initial notice in Federal

Register: May 7, 1997 (62 FR 24986).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated December 10,

1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: B. F. Jones Memorial Library,

663 Franklin Avenue, Aliquippa, PA

15001.

Entergy Operations, Inc.

[Docket No. 50–382]

Waterford Steam Electric Station, Unit

3, St. Charles Parish, Louisiana

Date of amendment request: July 17,

1996, as supplemented by letters dated

June 3, and July 7, 1997. Also,

application dated April 11, 1997.

Brief description of amendment: The

amendment changes the Appendix A

Technical Specification (TS) 3.7.1.3 by

increasing the minimum required

contained water volume in Condensate

Storage Pool from 82 percent to 91

percent indicated level. In addition, this

amendment expands the applicability of

TS 3.7.1.3 to include Mode 4

operational requirements. The

amendment also deletes Action (b) in

TS 3.7.1.3 and its associated

surveillance requirement in Waterford 3

TSs.

Date of issuance: December 18, 1997.

Effective date: December 18, 1997, to

be implemented within 60 days.

Amendment No.: 137.

Facility Operating License No. NPF–

38: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: March 26, 1997 (62 FR 14461),

July 30, 1997 (62 FR 40849) and April

22, 1997 (62 FR 19624).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated December 18,

1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: University of New Orleans

Library, Louisiana Collection, Lakefront,

New Orleans, LA 70122.

Florida Power Corporation, et al.

[Docket No. 50–302]

Crystal River Unit No. 3 Nuclear

Generating Plant, Citrus County, Florida

Date of application for amendment:

August 26, 1997.

Brief description of amendment: The

amendment involves a revision to the

design basis of the Emergency Diesel

Generator (EDG) Air Handling System at

Crystal River 3 resulting from the EDG

upgrade modification which increased

the 200-hour and 2000-hour service

ratings for each EDG.

Date of issuance: December 12, 1997.

Effective date: December 12, 1997.

Amendment No.: 160.

Facility Operating License No. DPR–

31: Amendment revises the Final Safety

Analysis Report.

Date of initial notice in Federal

Register: September 24, 1997 (62 FR

50004).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated December 12,

1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Coastal Region Library, 8619

W. Crystal River, Florida 34428
68322 Federal Register / Vol. 62, No. 250 / Wednesday, December 31, 1997 / Notices

Indiana Michigan Power Company

[Docket Nos. 50–315 and 50–316]

Donald C. Cook Nuclear Plant, Unit

Nos. 1 and 2, Berrien County, Michigan

Date of application for amendments:

September 19, 1997 (AEP:NRC:1278).

Brief description of amendments: The

amendments modify Technical

Specification 4.5.2.d.1 to delete the

interlock that would close the Residual

Heat Removal (RHR) suction valves if

the Reactor Coolant System (RCS)

pressure were to increase to 600 psig

while retaining the interlock that would

prevent the suction valves from opening

while the RCS pressure is above the

RHR system design pressure. This

change maintains the open interlock

function and allows continued

deactivation of the isolation valves to

assure RHR availability and provide low

temperature overpressure protection.

Date of issuance: December 10, 1997.

Effective date: December 10, 1997,

with full implementation within 45

days.

Amendment Nos.: 219 and 203.

Facility Operating License Nos. DPR–

58 and DPR–74: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: October 22, 1997 (62 FR

54861).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated December 10,

1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Maud Preston Palenske

Memorial Library, 500 Market Street, St.

Joseph, Michigan 49085.

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:

October 7, 1996, as supplemented by

letter dated May 9, 1997.

Brief description of amendments:

These amendments modify

Susquehanna Steam Electric Station,

Units 1 and 2, Technical Specifications

Table 3.3.2–2 by revising the trip

setpoints and allowable values for

secondary containment isolation

radiation monitors.

Date of issuance: December 8, 1997.

Effective date: December 8, 1997.

Amendment Nos.: 170 and 143.

Facility Operating License Nos. NPF–

14 and NPF–22: The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: December 18, 1996 (61 FR

66716).

The May 9, 1997, letter provided

clarifying information that did not

change the initial proposed no

significant hazards consideration

determination.

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated December 8,

1997.

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.

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:

April 4, 1997, as supplemented April

14, June 6, and September 2, 1997.

Brief description of amendments:

These amendments clarify the scope of

the surveillance requirements for

response time testing of instrumentation

in the reactor protection system,

isolation actuation system, and

emergency core cooling system in the

Technical Specifications for each unit

(Sections 4.3.1.3, 4.3.2.3, and 4.3.3.3).

Date of issuance: December 8, 1997.

Effective date: December 8, 1997.

Amendment Nos.: 171 and 144.

Facility Operating License Nos. NPF–

14 and NPF–22: The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: April 17, 1997 (62 FR 17885).

The April 14, June 6, and September

2, 1997, letters provided clarifying

information that did not change the

original proposed no significant hazards

consideration determination.

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated December 8,

1997.

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:

December 14, 1995, as supplemented

September 26, 1997.

Brief description of amendment: The

amendment proposes to change the

James A. FitzPatrick Technical

Specifications to incorporate the

inservice testing requirements of

Section XI of the American Society of

Mechanical Engineers Boiler and

Pressure Vessel Code.

Date of issuance: December 2, 1997.

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment No.: 241.

Facility Operating License No. DPR–

59: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: January 22, 1996 (61 FR 1635).

The September 26, 1997, letter

provided clarifying information that did

not change the initial proposed no

significant hazards consideration

determination.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated December 2,

1997.

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.

Rochester Gas and Electric Corporation

[Docket No. 50–244]

R. E. Ginna Nuclear Power Plant, Wayne

County, New York

Date of application for amendment:

September 29, 1997, as supplemented

October 8, 1997.

Brief description of amendment: The

amendment revises the Ginna Station

Technical Specifications (TS) to allow

referencing of revision of the Ginna

Station pressure and temperature limits

report for the reactor coolant system

pressure and temperature limits and low

temperature overpressure protection

limits. The amendment also corrects a

typographical error in the TSs.

Date of issuance: December 9, 1997.

Effective date: December 9, 1997.

Amendment No.: 70.

Facility Operating License No. DPR–

18: Amendment revised the Technical

Specifications.
Federal Register / Vol. 62, No. 250 / Wednesday, December 31, 1997 / Notices 68323

Date of initial notice in Federal

Register: November 5, 1997 (62 FR

59921).

The September 29 and October 8,

1997, superseded in their entirety the

applications dated December 13, 1996,

April 24, 1997, and June 3, 1997.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated December 9,

1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Rochester Public Library, 115

South Avenue, Rochester, New York

14610.

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:

December 22, 1995, as supplemented by

letter dated November 25, 1997.

Brief description of amendments:

These amendments revise License

Conditions 2.E and 2.G for the San

Onofre Nuclear Generating Station

(SONGS), Units 2 and 3. The

amendments delete the physical

protection program reporting

requirement from License Condition

2.G, and clarify in License Condition 2.E

that not all documents composing the

physical protection program plans

necessarily contain safeguards

information.

Date of issuance: December 16, 1997.

Effective date: December 16, 1997.

Amendment Nos.: Unit 2—138; Unit

3—130.

Facility Operating License Nos. NPF–

10 and NPF–15: The amendments

revised the Facility Operating Licenses.

Date of initial notice in Federal

Register: November 5, 1997 (62 FR

59921). The November 25, 1997, letter

provided additional clarifying

information and did not change the

initial no significant hazards

consideration determination. The

Commission’s related evaluation of the

amendments is contained in a Safety

Evaluation dated December 16, 1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Main Library, University of

California, P. O. Box 19557, Irvine,

California 92713.

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:

May 14, 1997, as supplemented October

15, 1997. The October 15, 1997,

submittal provided clarifying

information only, and did not change

the proposed no significant hazards

consideration determination.

Brief description of amendments: The

proposed action consists of changes to

the Technical Specifications (TS)

revising Surveillance Requirement

4.7.1.7.2.a for both units to clarify the

testing and inspection methodology of

the turbine governor control valves. The

proposed changes also provide

clarification in the TS Bases Section 3/

4 7.1.7 for the Turbine Valve Freedom

Testing of the turbine governor control

valves.

Date of issuance: December 4, 1997.

Effective date: December 4, 1997.

Amendment Nos.: 207 and 188.

Facility Operating License Nos. NPF–

4 and NPF–7: Amendments revised the

Technical Specifications.

Date of initial notice in Federal

Register: July 30, 1997 (62 FR 40860).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated December 4,

1997.

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 24th day

of December 1997.

For the Nuclear Regulatory Commission.

Elinor G. Adensam,

Acting Director, Division of Reactor Projects—

III/IV, Office of Nuclear Reactor Regulation.

[FR Doc. 97–33968 Filed 12–30–97; 8:45 am]

BILLING CODE 7590–01–P

NUCLEAR REGULATORY

COMMISSION

Membership on the Executive

Resources Board

AGENCY: Nuclear Regulatory

Commission.

ACTION: Appointment to the Executive

Resources Board for the Senior

Executive Service.

SUMMARY: The U.S. Nuclear Regulatory

Commission (NRC) has announced the

following appointments to the NRC

Executive Resources Board. The

Executive Resources Board is

responsible for providing institutional

continuity in executive personnel

management by overseeing NRC’s

Senior Executive Service (SES) and

Senior Level System (SLS) succession

planning, merit staffing, and position

management activities.

Appointees

L. Joseph Callan, Executive Director for

Operations, Chair

Karen D. Cyr, General Counsel

Anthony J. Galante, Chief Information

Officer

Jesse L. Funches, Chief Financial Officer

Hugh L. Thompson, Jr., Deputy

Executive Director for Regulatory

Programs

Ashok C. Thadani, Acting Deputy

Executive Director for Regulatory

Effectiveness

Patricia G. Norry, Deputy Executive

Director for Management Services

Samuel J. Collins, Director, Office of

Nuclear Reactor Regulation

Carl J. Paperiello, Director, Office of

Nuclear Material Safety and

Safeguards

Malcolm R. Knapp, Acting Director,

Office of Nuclear Regulatory Research

Timothy T. Martin, Director, Office for

the Analysis and Evaluation of

Operational Data

Edward L. Halman, Director, Office of

Administration

Paul E. Bird, Director, Office of Human

Resources

Irene P. Little, Director, Office of Small

Business and Civil Rights

John C. Hoyle, Secretary of the

Commission

Hubert J. Miller, Regional

Administrator, Region I

Luis A. Reyes, Regional Administrator,

Region II

A. Bill Beach, Regional Administrator,

Region III

Ellis W. Merschoff, Regional

Administrator, Region IV

EFFECTIVE DATE: December 18, 1997.

FOR FURTHER INFORMATION CONTACT:

Carolyn J. Swanson, Secretary,

Executive Resources Board, U.S.

Nuclear Regulatory Commission,

Washington, D.C. 20555 (301) 415–7530.

Dated at Rockville, Maryland, this 19th day

of December, 1997.

For the U.S. Nuclear Regulatory

Commission.

Paul E. Bird,

Director, Office of Human Resources.

[FR Doc. 97–34075 Filed 12–30–97; 8:45 am]

BILLING CODE 7590–01–P