Federal Register / Vol. 61, No. 119 / Wednesday, June 19, 1996 / Notices 31171

determined that this action is not a

major rule and has verified this

determination with the Office of

Information and Regulatory Affairs of

OMB.

Statement of Policy

The Commission’s policy on

Conversion to the Metric System

remains essentially as stated in the

Federal Register (57 FR 46202) of

October 7, 1992.

The NRC supports and encourages the

use of the metric system of

measurement by licensed nuclear

industry. In order to facilitate the use of

the metric system by licensees and

applicants, beginning January 7, 1993,

the NRC will publish the following

documents in dual units: New

regulations, major amendments to

existing regulations, regulatory guides,

NUREG-series documents, policy

statements, information notices, generic

letters, bulletins, and all written

communications directed to the public.

Documents specific to a licensee, such

as inspection reports and docketed

material dealing with a particular

licensee, will be in the system of units

employed by the licensee. This protocol

reflects a general approach that only

documents applicable to all licensees, or

to all licensees of a given type in which

a licensee may operate in the metric

system will contain dual units.

Otherwise, English or metric units alone

are permissible. In dual-unit documents,

the first unit presented will be in the

International System of Units with the

English unit shown in brackets. The

NRC will modify existing documents

and procedures as needed to facilitate

use of the metric system by licensees

and applicants. In addition, the NRC

will provide staff training as needed.

Further, through its participation in

national, international, professional,

and industry standards organizations

and committees and through its work

with other industry organizations and

groups, the NRC will encourage and

further the use of the metric system in

formulating and adopting standards and

policies for the licensed nuclear

industry.

However, if the NRC concludes that

the use of any particular system of

measurement would be detrimental to

the public health and safety, the

Commission will proscribe the use of

that system by regulation, order, or

other appropriate means. In particular,

all event reporting and emergency

response communications between

licensees, the NRC, and State and local

authorities will be in the English system

of measurement. Further, the NRC will

follow the Federal Acquisition

Regulation and the General Services

Administration metrication program in

executing procurements. Lastly, the

Commission considers this policy final

and conversion to the metric system

complete. The Commission does not

intend to revisit this policy unless it is

causing an undue burden or hardship.

Dated at Rockville, Maryland, this 12th day

of June 1996.

For the Nuclear Regulatory Commission.

John C. Hoyle,

Secretary of the Commission.

[FR Doc. 96–15397 Filed 6–17–96; 8:45 am]

BILLING CODE 7590–01–P

Biweekly Notice

Applications and Amendments to

Facility Operating Licenses Involving

No Significant Hazards Considerations

I. Background

Pursuant to Public Law 97–415, the

U.S. Nuclear Regulatory Commission

(the Commission or NRC staff) is

publishing this regular biweekly notice.

Public Law 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 May 24,

1996, through June 7, 1996. The last

biweekly notice was published on June

5, 1996 (61 FR 28604).

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 Review 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 July 19, 1996, 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
31172 Federal Register / Vol. 61, No. 119 / Wednesday, June 19, 1996 / Notices

petition for leave to intervene shall be

filed in accordance with the

Commission’s ‘‘Rules of Practice for

Domestic Licensing Proceedings’’ in 10

CFR Part 2. Interested persons should

consult a current copy of 10 CFR 2.714

which is available at the Commission’s

Public Document Room, the Gelman

Building, 2120 L Street, NW.,

Washington, DC and at the local public

document room for the particular

facility involved. If a request for a

hearing or petition for leave to intervene

is filed by the above date, the

Commission or an Atomic Safety and

Licensing Board, designated by the

Commission or by the Chairman of the

Atomic Safety and Licensing Board

Panel, will rule on the request and/or

petition; and the Secretary or the

designated Atomic Safety and Licensing

Board will issue a notice of a hearing or

an appropriate order.

As required by 10 CFR 2.714, a

petition for leave to intervene shall set

forth with particularity the interest of

the petitioner in the proceeding, and

how that interest may be affected by the

results of the proceeding. The petition

should specifically explain the reasons

why intervention should be permitted

with particular reference to the

following factors: (1) the nature of the

petitioner’s right under the Act to be

made a party to the proceeding; (2) the

nature and extent of the petitioner’s

property, financial, or other interest in

the proceeding; and (3) the possible

effect of any order which may be

entered in the proceeding on the

petitioner’s interest. The petition should

also identify the specific aspect(s) of the

subject matter of the proceeding as to

which petitioner wishes to intervene.

Any person who has filed a petition for

leave to intervene or who has been

admitted as a party may amend the

petition without requesting leave of the

Board up to 15 days prior to the first

prehearing conference scheduled in the

proceeding, but such an amended

petition must satisfy the specificity

requirements described above.

Not later than 15 days prior to the first

prehearing conference scheduled in the

proceeding, a petitioner shall file a

supplement to the petition to intervene

which must include a list of the

contentions which are sought to be

litigated in the matter. Each contention

must consist of a specific statement of

the issue of law or fact to be raised or

controverted. In addition, the petitioner

shall provide a brief explanation of the

bases of the contention and a concise

statement of the alleged facts or expert

opinion which support the contention

and on which the petitioner intends to

rely in proving the contention at the

hearing. The petitioner must also

provide references to those specific

sources and documents of which the

petitioner is aware and on which the

petitioner intends to rely to establish

those facts or expert opinion. Petitioner

must provide sufficient information to

show that a genuine dispute exists with

the applicant on a material issue of law

or fact. Contentions shall be limited to

matters within the scope of the

amendment under consideration. The

contention must be one which, if

proven, would entitle the petitioner to

relief. A petitioner who fails to file such

a supplement which satisfies these

requirements with respect to at least one

contention will not be permitted to

participate as a party.

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:

Docketing and Services Branch, or may

be delivered to the Commission’s Public

Document Room, the Gelman Building,

2120 L Street, NW., Washington DC, by

the above date. Where petitions are filed

during the last 10 days of the notice

period, it is requested that the petitioner

promptly so inform the Commission by

a toll-free telephone call to Western

Union at 1-(800) 248–5100 (in Missouri

1-(800) 342–6700). The Western Union

operator should be given Datagram

Identification Number N1023 and the

following message addressed to (Project

Director): petitioner’s name and

telephone number, date petition was

mailed, plant name, and publication

date and page number of this Federal

Register notice. A copy of the petition

should also be sent to the Office of the

General Counsel, U.S. Nuclear

Regulatory Commission, Washington,

DC 20555–0001, and to 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.

Boston Edison Company, Docket No.

50–293, Pilgrim Nuclear Power Station,

Plymouth County, Massachusetts

Date of amendment request: April 25,

1996

Description of amendment request:

The proposed amendment would

change the definition of OperableOperability, revise Technical

Specifications (TSs) and associated

Bases Section for TSs 3.5.F.1, ‘‘Core and

Containment Cooling systems,’’ TSs

3.9.B.1, 3.9.B.2, 3.9.B.3, 3.9.b.4,

‘‘Auxiliary Electrical System,’’ and TSs

3.7.B.1.a, c, and e, and 3.7.b.2.a, c, and

e, ‘‘Standby Gas Treatment System and

Control Room High Efficiency Air

Filtration System,’’ and delete TSs

4.5.F.1, ‘‘Core and Containment Cooling

Systems,’’ and 3.7.B.1.f, ‘‘Standby Gas

Treatment System and Control Room

High Efficiency Air Filtration System.’’

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed amendment does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

Operation of PNPS [Pilgrim Nuclear Power

Station] in accordance with the proposed

license amendment will not involve a

significant increase in the probability or

consequences of an accident previously

evaluated because of the following:

Definition of ‘‘Operable-Operability’’

Definitions perform a supporting function

for other sections of the TS. The definition

of ‘‘Operable-Operability’’ affects the manner
Federal Register / Vol. 61, No. 119 / Wednesday, June 19, 1996 / Notices 31173

in which the requirements for a Limiting

Condition for Operation (LCO) and its

associated remedial actions are applied when

a support system is inoperable. This

definition re-affirms the principle that a

system is operable when it is capable of

performing its specified function and when

all necessary support systems are also

capable of performing their related support

functions. The corollary is that a system is

inoperable when it is not capable of

performing its specified function or when a

necessary support system is not capable of

performing its related support function.

No changes are being made to the plant

design, system configuration, or method of

operation. The proposed change does not

affect the ability of the AC power sources to

perform their required safety functions nor

affect the ability of the features they support

to perform their respective safety functions.

Therefore, the proposed change does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

EDG [Emergency Diesel Generator]

An Individual Plant Examination (IPE) for

Internal Events was submitted to the NRC in

response to Generic Letter 88–20 in

September 1992. The IPE was used to

quantify the overall impact of the proposed

14 day allowed outage time on core damage

frequency. Part III provides the results of a

comprehensive Probabilistic Safety

Assessment (PSA) of the impact of the

proposed AOTs [allowed outage times] for

the EDGs and Startup and Shutdown

transformers. As shown in Part III, there is

not a significant increase in risk due to the

proposed change. Thus the proposed change

does not involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The existing specification 3.9.B.1 is being

separated into two segments (a and b)

because of the proposed and different AOTs

for the Startup and Shutdown transformers.

As a result of the PSA, the AOT for the

Startup transformer (a) is reduced from 7

days to 72 hours, while the AOT for the

Shutdown transformer (b) remains at 7 days.

The reduction of the AOT from 7 days to 3

days is based on the relative risk importance

of the Startup transformers support to the

balance of plant systems. Similarly, an

additional reduction from 72 hours to 48

hours is proposed in the AOT for a

simultaneous loss of both the Startup

transformer and an EDG (TS 3.9.B.4.b) based

upon the Startup transformer’s contribution

to risk in relation to the EDG 14-day AOT

risk assessment analysis and that two power

sources have been removed from the

associated bus. The AOT reductions

represent a measurable decrease in risk as

assessed in the PSA. Thus, the probability or

consequences of an accident previously

evaluated are not significantly increased.

The current technical specifications allow

one EDG to be out of service for three days

based on the availability of the SUT [startup

transformer] and SDT [shutdown

transformer] and the fact that each EDG

carries sufficient engineered safeguards

equipment to cover all design basis

accidents. With one EDG out of service and

a Loss of Offsite Power (LOOP) condition, the

capability to power vital and auxiliary

system components remains available via the

other EDG, and for one train of ESF

equipment via the SDT for all operating,

transient and accident conditions. Increasing

the EDG AOT to 14 days provides flexibility

in the maintenance and repair of the EDGs.

The EDG unavailability will be monitored

and trended in accordance with the

Maintenance Rule. The PSA analyses

supports the change to a 14 day AOT for the

EDGs based on an insignificant increase in

overall risk. Implementation of the proposed

change is expected to result in less than a one

percent increase in the baseline core damage

frequency (2.84E–05/yr), which is considered

to be insignificant relative to the underlying

uncertainties involved with probabilistic

safety assessments. Additional conditions are

added to the Standby Liquid Control,

Standby Gas Treatment, and Control Room

High Efficiency Air Filtration systems

requiring the EDG associated with these

systems to remain operable while in the 14

day EDG AOT. Thus, the 14 day EDG AOT

does not involve a significant increase in the

probability or consequences of an accident

previously evaluated.

Eliminating the 4.5.F.1 requirement for

daily testing of the operable diesel generator

when the redundant diesel generator

becomes inoperable is consistent with the

guidance provided in Generic Letter 93–05.

The change does not affect the ability of the

emergency diesel generator to perform on

demand, and by actually lowering the

number of demands to demonstrate

operability, reduces the probability of

equipment failure. The redundant EDG will

remain in service during the entire period of

inoperability of the out-of-service EDG. If a

common cause failure cannot be ruled out,

the redundant EDG will be tested to assure

operability. The proposed revisions do not

involve a significant change to the plant

design or operation, only to the manner in

which remaining equipment is confirmed to

be operable, which is consistent with NRC

guidance. Thus operation of PNPS in

accordance with the proposed license

amendment will not involve a significant

increase in the probability or consequences

of an accident previously evaluated.

The 3.9.B.1 and 2 requirements to

demonstrate both EDGs and associated

emergency buses operable are deleted. This

change is based on the NRC guidance

provided in item 10.1 of Generic Letter 93–

05, ‘‘Line-Item Technical Specification

Improvements to Reduce Surveillance

Requirements for Testing During Power

Operation.’’ Revising the methods for

verifying EDG and emergency bus operability

does not physically alter the plant or have an

affect on the probability or consequences of

an accident previously evaluated. Deleting

the testing requirements for an EDG when the

other EDG is inoperable does not increase the

probability or consequences of an accident

previously evaluated because the reliability

program and routinely performed TS

surveillances continue to provide the added

assurance sought by the testing. The

elimination of this testing will serve to

improve the overall reliability of the EDGs.

Since the proposed change does not affect the

design or negatively affect the performance of

the EDGs, the change will not result in a

significant increase in the consequences or

probability of an accident previously

analyzed.

SGT [Standby Gas Treatment] and

CRHEAF [Control Room High Efficiency Air

Filtration]

During normal plant operation, with one

SGT or CRHEAF subsystem inoperable, the

inoperable subsystem must be restored to

operable status in 7 days. In this condition,

the remaining operable SGT or CRHEAF

subsystem is adequate to perform the

required radioactivity release control

function. However, the overall system

reliability is reduced because a single failure

in the operable subsystem could result in the

radioactivity release control function not

being adequately performed. The 7 day

completion time is based on consideration of

such factors as the availability of the operable

redundant SGT subsystem and the low

probability of a DBA [design basis accident]

occurring during this period.

If the SGT or CRHEAF subsystem cannot

be restored to operable status within 7 days

when in the Run, Startup, or Hot Shutdown

MODE, the plant must be brought to a MODE

in which the LCO does not apply. To achieve

this status, the plant must be brought to at

least Hot Shutdown within 12 hours and to

Cold Shutdown within 36 hours. The

allowed completion times are reasonable,

based on operating experience, to reach the

required plant conditions from full power

conditions in an orderly manner and without

challenging plant systems.

Current TS governing refueling operations

restrict fuel movement if one train of SGTS

or one train of CRHEAF are inoperable. In

this condition the remaining operable SGT

and CRHEAF trains are adequate to perform

the required radioactivity release control

functions. However, the overall system

reliability is reduced because a single failure

in the operable train could result in the

radioactivity release control function of the

systems not being adequately performed.

New requirements are added that require if

one train of SGT or CRHEAF is inoperable,

the redundant train of SGT or CRHEAF must

be demonstrated to be operable within 2

hours. This substantiates the availability of

the operable trains. Fuel handling is limited

only to the following 7 days and if the

inoperable train is not returned to an

operable condition within that time frame,

the operable SGT train is placed in operation

or fuel handling activities are suspended. For

CRHEAF, after 7 days, the operable

subsystem is demonstrated operable in

accordance with existing surveillances on a

daily basis. The proposed changes do not

modify system design, use, or configuration

in a manner different from their original

design and therefore do not involve a

significant increase in the consequences or

probability of an accident previously

analyzed.

The revisions to make the SGT and

CRHEAF TS sections similar in wording are

made to enhance usability and alleviate

possible confusion. These changes are strictly

editorial, have no impact, and do not alter
31174 Federal Register / Vol. 61, No. 119 / Wednesday, June 19, 1996 / Notices

technical content or meaning of the

specifications. These editorial changes do not

involve a significant increase in the

probability or consequences of an accident

previously analyzed.

2. The proposed amendment does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

The operation of PNPS in accordance with

the proposed license amendment will not

create the possibility of a new or different

kind of accident from any accident

previously evaluated because of the

following:

Definition of ‘‘Operable-Operability’’

The revised definition redefines the AC

power needs to allow either onsite or offsite

power available for systems/subsystems to be

considered operable. This does not

compromise the level of safety already

afforded to such systems/subsystems because

the functional operability requirements

continue to be assured through the technical

specifications applicable to such systems/

subsystems. AC power availability continues

to be assured through existing and proposed

surveillances and action statements

applicable to AC power systems. Reducing

the need for both onsite and offsite power

sources in order to consider operable, the

systems/subsystems powered by these AC

power sources, provides additional

operational flexibility by allowing redundant

systems/subsystems to still be considered

‘‘operable’’ within the requirements of their

functional operability requirements. No new

change or modes of plant operation are

involved. Therefore, operation in accordance

with the revised definition does not

introduce any new or different kind of

accident from any accident previously

evaluated.

EDG

The proposed amendment will extend the

action completion/allowed outage time for an

inoperable emergency diesel generator from

72 hours to 14 days. The EDGs are designed

as backup AC power sources for essential

safety systems in the event of loss of offsite

power. The proposed AOT does not change

the conditions, operating configurations or

minimum amount of operating equipment

assumed in the safety analysis for accident

mitigation. The EDGs and AC equipment are

not accident initiators. No change is being

made in the manner in which the EDG’s

provide plant protection. No new modes of

plant operation are involved. An extended

AOT for one EDG does not increase the

probability of occurrence of a new or

different kind of accident previously

evaluated. The PSA results concluded that

the risk contribution of the EDG AOT

extension is insignificant.

The current Pilgrim Technical

Specifications requiring immediate and daily

testing of the redundant operable EDG is

based on the assumption that the increased

testing provides additional assurance that the

equipment is available should it be needed.

Industry experience indicates that repetitive

testing can place demands and wear on the

EDG without necessarily providing

additional confidence of availability. Also,

the new surveillance requires verification

that offsite power is available and that a

common cause failure is not present. These

actions provide assurance that the required

emergency buses can be energized with no

loss of functions to mitigate accident or

transient conditions. In addition, Pilgrim has

implemented an EDG reliability program to

maintain reliability of EDGs. The proposed

change does not introduce any new mode of

plant operation or new accident precursors,

involve any physical alterations to plant

configurations, or make changes to system set

points that could initiate a new or different

kind of accident. Therefore, operation in

accordance with the proposed change does

not create the possibility of a new or different

kind of accident from any previously

evaluated.

The AOT for an inoperable Startup

Transformer is reduced from 7 days to 72

hours based upon the PSA that was

performed to quantitatively assess the risk

impact of the proposed amendment. The

proposed reduction in AOT improves overall

AC power source availability because the

SUT will potentially be inoperable for shorter

time periods. Therefore, reducing the AOT

does not create the possibility of a new or

different kind of accident from any accident

previously evaluated.

SGT and CRHEAF

The SGT system is designed to filter

radioactive materials from the secondary

containment following a postulated DBA or

fuel handling accident prior to release to the

environment to ensure compliance with 10

CFR 100 limits.

The CRHEAF is designed to filter intake air

for the control room atmosphere during

conditions when normal intake air may be

contaminated.

The proposed revisions do not affect the

ability of the SGTS or CRHEAF to perform

their intended function, do not create the

possibility of a new or different kind of

accident from the loss of coolant or fuel

handling accidents previously analyzed, and

do not modify system configuration, use, or

design. Therefore, operating Pilgrim in

accordance with this change will not create

the possibility of a new or different kind of

accident from any accident previously

analyzed.

The revisions to make the SGT and

CRHEAF TS sections similar in wording are

made to enhance usability and alleviate

possible confusion. These changes are strictly

editorial, have no impact, and do not alter

technical content or meaning of the

specifications. These editorial changes do not

create the possibility of a new or different

kind of accident from any previously

analyzed.

3. The proposed amendment does not

involve a significant reduction in a margin of

safety.

The operation of PNPS in accordance with

the proposed license amendment will not

involve a significant reduction in a margin of

safety because of the following:

Definition of ‘‘Operable-Operability’’

The implementation of the ‘‘Operability’’

definition clarifies the relationship between

AC power supplies and the operability status

of the equipment requiring AC power. No

change is being made in which the plant

systems relied upon in the safety analyses

provide plant protection. Plant safety

margins are maintained through the

limitations established in the TS LCOs. Since

there will be no significant reduction to the

physical design or operation of the plant

there will be no significant reduction to any

of these margins.

EDG

Operation of PNPS in accordance with the

proposed license amendment will not

involve a significant reduction in a margin of

safety. As shown in Part III [of the

application dated April 25, 1996],

incorporation of the proposed change

involves an insignificant reduction in the

margin of safety.

The proposed changes do not significantly

reduce the basis for any technical

specification related to the establishment of,

or the maintenance of, a safety margin nor do

they require physical modifications to the

plant. Additional conditions are added to the

Standby Liquid Control, Standby Gas

Treatment, and Control Room High

Efficiency Air Filtration systems requiring

the diesel generator associated with the

redundant operable trains of these systems to

remain operable while in the 14 day EDG

AOT. Moreover, the PSA results showed that

the risk contribution of extending the AOT

for an inoperable EDG is insignificant. The

reduction in the AOT for the SUT could

improve availability, therefore, reducing

overall risk. Likewise the proposed changes

in the deletion of testing have no impact on

the safety margin.

As previously stated, implementation of

the proposed changes is expected to result in

an insignificant increase in: (1) power

unavailability to the emergency buses (given

that a loss of offsite power has occurred), and

(2) core damage frequency. Implementation

of the proposed changes does not increase

the consequences of a previously analyzed

accident nor significantly reduce a margin of

safety. Functioning of the EDGs and the

manner in which limiting conditions of

operation are established are unaffected.

SGT and CRHEAF

SGT and CRHEAF contribute to the margin

of safety by supporting the secondary

containment system during fuel handling by

mitigating the consequences of a fuel

handling event. Allowing fuel movement to

continue as established in the LCOs does not

involve a significant reduction in the margin

of safety because the first line of defense, the

other SGT and CRHEAF trains will be

operable. The proposed change will allow

placing the Operable SGT subsystem in

operation, or in the case of CRHEAF,

conducting daily testing, as an alternative to

suspending movement of irradiated fuel. This

alternative is less restrictive than the existing

requirement, however, the proposed

requirements ensure that the remaining

subsystem is operable, that no failures that

could prevent actuation have occurred, and

that any failure would be readily detected.

The proposed change does not result in a

significant reduction in a margin of safety

because it allows operations which have the

potential for releasing radioactive material to

the secondary containment to continue only

if the system designed to mitigate the
Federal Register / Vol. 61, No. 119 / Wednesday, June 19, 1996 / Notices 31175

consequences of this release is functioning.

Proper operation of only one SGT or one

CRHEAF subsystem is sufficient to mitigate

the consequences of any analyzed accident.

Therefore, this change does not change any

of the assumptions in the accident analysis

and does not involve a significant reduction

in a margin of safety.

The revisions to make the SGT and

CRHEAF TS sections similar in wording are

made to enhance usability and alleviate

possible confusion. These changes are strictly

editorial, have no impact, and do not alter

technical content or meaning of the

specifications. These editorial changes 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 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: Plymouth Public Library, 11

North Street, Plymouth, Massachusetts

02360.

Attorney for licensee: W. S. Stowe,

Esquire, Boston Edison Company, 800

Boylston Street, 36th Floor, Boston,

Massachusetts 02199.

NRC Project Director: Jocelyn A.

Mitchell, Acting

Connecticut Yankee Atomic Power

Company, Docket No. 50–213, Haddam

Neck Plant, Middlesex County,

Connecticut

Date of amendment request: April 22,

1996

Description of amendment request:

The licensee is proposing to change the

technical specifications to reflect a

revision to the overload cutoff limit on

the manipulator crane inside the

containment at the Haddam Neck Plant.

Due to a change in fuel design and

supplier, the heaviest fuel assembly

design starting in Cycle 20 will be the

Westinghouse-supplied LOPAR design.

Therefore, the heaviest combination

beginning in Cycle 20 will be the

Westinghouse LOPAR fuel assembly

with a full-length rod cluster control

assembly (RCCA) inserted. It will now

be used as the standard for the overload

cutoff limit on the manipulator crane.

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 will revise the

method of determining the overload cutoff

limit for the manipulator crane. The actual

absolute value of the cutoff limit will not be

increased and will not affect the [probability]

of any plant accidents.

Since there is no actual increase in the

absolute overload cutoff limit, there will be

no adverse effects to the crane, cables, or

associated hardware. Therefore, there is no

impact on the crane’s ability to perform its

intended function. Even though the net

lifting forces on an individual assembly have

increased 25 pounds, the limit is within the

recommended Westinghouse guidelines with

respect to fuel handling and will not result

in potential damage to assembly grids during

fuel handling activities.

As such, CYAPCO [Connecticut Yankee

Atomic Power Company] has concluded that

these changes do not involve an 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 conservatively revise the

method of determining the overload cutoff

limit for the manipulator crane. There is no

impact on the basic functioning of plant

systems or equipment. Therefore, the change

does not create a malfunction that is different

from those previously evaluated.

As such, the proposed changes described

above do not create the possibility of a new

or different kind of accident from any

previously evaluated.

3. [The proposed change does not] involve

a significant reduction in a margin of safety.

The proposed revisions in the methodology

for determining the overload cutoff limit for

the manipulator crane is conservative and in

accordance with vendor standards. The

changes do not adversely affect any

equipment credited in the safety analysis.

Also, the changes do not adversely affect the

probability or consequences of any plant

accident, including the fuel handling

accident or offsite doses associated with

those accidents.

As such, the proposed changes have no

significant impact on 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: Russell Library, 123 Broad

Street, Middletown, CT 06457

Attorney for licensee: Lillian M.

Cuoco, Esq., Senior Nuclear Counsel,

Northeast Utilities Service Company,

P.O. Box 270, Hartford, CT 06141–0270

NRC Project Director: Phillip F.

McKee

Duke Power Company, Docket Nos. 50–

413 and 50–414, Catawba Nuclear

Station, Units 1 and 2, York County,

South Carolina

Date of amendment request:

December 14, 1995, as supplemented by

letter dated May 16, 1996

Description of amendment request:

The proposed amendments would

change the Technical Specifications

(TS) to improve the TS Action

Statements and Surveillance

Requirements for diesel generators in

accordance with the recommendations

and guidance in Generic Letter 93–05,

Generic Letter 94–01, NUREG–1366,

and NUREG–1431. The proposed

amendments would also incorporate

technical and administrative changes.

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

Operation of the facilities in accordance

with the requested amendments will not

involve a significant increase in the

probability or consequences of an accident

previously evaluated. Improvements to the

LCOs [limiting condition for operation] and

surveillance requirements for the emergency

diesel generators do not affect their capability

to provide emergency power to plant vital

instruments and safety related equipment. In

fact, these improvements make the diesel

generators more reliable since they

significantly reduce the amount of wear and

stress due to excessive and unnecessary

testing. The proposed monthly testing of the

diesel generator continues to ensure that the

system is ready for service when needed. The

fast starts and fast loadings continue to

ensure that the timing and loading

requirements for engineered safety features

actuation are met. The proposed changes do

not affect any of the design basis accident

analyses previously evaluated. Therefore,

these proposed changes do not involve any

increase in the probability or consequences

of any accident previously evaluated. The

proposed changes are fully consistent with

the recommendations and guidance

contained in GL [Generic Letter] 93–05, GL

94–01, NUREG–1366, NUREG–1431, and are

compatible with plant operating experience.

Criterion 2

Operation of the facilities in accordance

with the requested amendments will not

create the possibility of a new or different

kind of accident from any accident

previously evaluated. The proposed changes

in fact improve the reliability of the diesel

generators by eliminating unnecessary wear

and stress. Improved reliability decreases the

failure probability which also decreases the

probability of an accident not previously

evaluated. None of the requested

amendments increase the common mode

failure probability thus would not increase

the chance of both EDG’s [emergency diesel
31176 Federal Register / Vol. 61, No. 119 / Wednesday, June 19, 1996 / Notices

generators] for a particular nuclear unit being

out of service simultaneously. The proposed

changes are fully consistent with the

recommendations and guidance contained in

GL 93–05, GL 94–01, NUREG–1366, NUREG–

1431, and are compatible with plant

operating experience.

Criterion 3

Operation of the facilities in accordance

with the requested amendments will not

involve a significant reduction in a margin of

safety. The proposed monthly testing of the

diesel generators continues to ensure that the

system is ready for service when needed. The

fast starts and fast loadings continue to

ensure that the timing and loading

requirements for engineered safety features

actuation are met. The proposed changes

improve the reliability of the diesel

generators. Implementation of the

Maintenance Rule also ensures continued

reliability of the diesel generators. No margin

of safety is decreased as a result of these TS

changes.

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: York County Library, 138 East

Black Street, Rock Hill, South Carolina

29730

Attorney for licensee: Mr. Albert Carr,

Duke Power Company, 422 South

Church Street, Charlotte, North Carolina

28242

NRC Project Director: Herbert N.

Berkow

Entergy Operations, Inc., et al., Docket

No. 50–416, Grand Gulf Nuclear

Station, Unit 1, Claiborne County,

Mississippi and Docket No. 40–458,

River Bend Station, Unit 1, West

Feliciana Parish, Louisiana

Date of amendment request: April 18,

1996, as supplemented by letter dated

June 4, 1996

Description of amendment request:

The licensee has proposed to (1) amend

Limiting Condition for Operation (LCO)

3.10.6 and Surveillance Requirement

3.10.6.3, and (2) add a Surveillance

Requirement 3.10.6.4 of the Technical

Specifications (TSs) for the Grand Gulf

Nuclear Station, Unit 1, and the River

Bend Station, Unit 1, to allow another

method of fuel movement and loading

in the core when control rods are

removed or withdrawn from defueled

core cells. Currently, LCO 3.10.6 allows

only fuel loading as part of the approved

spiral reloading sequence to prevent

fuel loading into core cells in which the

control rod has been removed or

withdrawn. This amendment request

does not withdraw this approved

method, revise the frequency of

performing the surveillance during fuel

loading, or alter the method of verifying

the fuel is being loaded in compliance

with the approved method. Grand Gulf

Unit 1 and River Bend Unit 1 are both

General Electric (GE) Boiling Water

Reactor (BWR)-6 plants, the latest

version of the GE design series.

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:

Entergy Operations, Inc. [(EOI)] propose[d]

to change the current Grand Gulf Nuclear

Station (GGNS) and River Bend Station (RBS)

Technical Specifications [(TSs)]. The specific

proposed change is to add an additional

method of performing fuel loading into LCO

3.10.6, ‘‘Multiple Control Rod Withdrawal -

Refueling’’. The proposed change would

allow fuel loading [in the core] if a positive

means of assuring fuel assemblies cannot be

loaded into a core cell with a withdrawn or

removed control rod is in effect. [Currently,

the TSs for both plants allow fuel assembles

to be loaded in compliance with an approved

spiral reload sequence which is used to

ensure the reactivity additions are

minimized. Spiral loadings encompass

reloading a core cell on the edge of a

continuous fueled region.]

The Commission has provided standards

for determining whether a no significant

hazards consideration exists as stated in 10

CFR 50.92(c). A proposed amendment to an

operating license involves no significant

hazards consideration if operation of the

facility in accordance with the proposed

amendment would not: (1) involve a

significant increase in the probability or

consequences of an accident previously

evaluated; or (2) create the possibility of a

new or different kind of accident from any

accident previously evaluated; or (3) involve

a significant reduction in a margin of safety.

Entergy Operations, Inc. [EOI] has

evaluated the no significant hazards

consideration in its request for this license

amendment and determined that no

significant hazards consideration results from

this change. In accordance with 10 CFR

50.91(a), Entergy Operations, Inc. [EOI] is

providing the analysis of the proposed

amendment against the three standards in 10

CFR 50.92(c). A description of the no

significant hazards consideration

determination follows:

I. The proposed change does not

significantly increase the probability or

consequences of an accident previously

evaluated.

The refueling interlocks (i.e., the refueling

equipment and one-rod-out interlocks)

allowed to be bypassed by Technical

Specification [TS] LCO 3.10.6 are explicitly

assumed in the analysis of the control rod

removal error or fuel loading error during

refueling. This analysis evaluates the

consequences of control rod withdrawal

during refueling. Criticality and, therefore,

subsequent prompt reactivity excursions are

prevented during the insertion of fuel,

provided all control rods are fully inserted

during the fuel insertion. The refueling

interlocks accomplish this by preventing

loading fuel into the core with any control

rod withdrawn, or by preventing withdrawal

of a rod from the core during fuel loading.

LCO 3.10.6 allows multiple control rod

withdrawals, control rod removals,

associated control rod drive (CRD) removal,

or any combination of these, and the ‘‘full in’’

position indication input to the refueling

interlocks is allowed to be bypassed for each

withdrawn control rod if all fuel has been

removed from the cell. This supports the

GGNS Updated Final Safety Analyses Report

(UFSAR) and RBS Updated Safety Analyses

Report (USAR) analyses since, with no fuel

assemblies in the core cell, the associated

control rod has no reactivity control function

and does not need to remain inserted. Prior

to reloading fuel into the cell, however, the

associated control rod must be inserted to

ensure that an inadvertent criticality does not

occur, as evaluated in the analysis.

The Technical Specification [TS]

requirements prohibiting fuel loading was

placed in the Technical Specifications [TSs]

for GGNS and RBS as part of the originally

enforced Technical Specification [TS]

requirements to resolve NRC concerns

identified in IE Information Notice No. 83–

35, ‘‘Fuel Movement with Control Rods

Withdrawn at BWRs,’’ (IEN 83–35). IEN 83–

35 details instances where fuel assemblies

were loaded into core cells while the control

rod was withdrawn and discusses that the

General Electric Company (GE) had issued

Service Information Letter (SIL) No. 372.

SIL No. 372 discusses a potential event

where 8 fuel assemblies are loaded into 2

[two] adjacent core cells where the control

rods are withdrawn and no action is taken to

recover from the errors. In this SIL GE

identified that the probability of such an

event occurring was extremely low but

potentially slightly higher than 10–6

probability of the event even further to where

it need not be considered credible (i.e., below

10–6 per reactor year), GE recommended that

the additional administrative control of

prohibiting loading fuel with withdrawn rods

be enforced.

The proposed change will only provide an

additional way to meet the intent of the

original GE recommendation. [The currently

approved method is listed in LCO 3.10.6 and

Surveillance Requirement 3.10.6.3.]. The

proposed change will provide the additional

allowance to perform fuel loading only if an

additional positive means of assuring fuel

assemblies cannot be loaded into a core cell

with a withdrawn or removed control rod is

in effect. The positive means will entail a

physical barrier such that, even if refueling

procedures were violated and an attempt was

made to load a fuel assembly into a core cell

with a withdrawn or removed control rod,

the action would be prevented. This

requirement provides sufficient additional

restrictions to meet the intent of the GE

recommendation to add additional

administrative controls to prevent the

postulated event from occurring.

The probability of an inadvertent criticality

occurring will continue to be precluded by
Federal Register / Vol. 61, No. 119 / Wednesday, June 19, 1996 / Notices 31177

the same number of layers of administrative

controls [as the currently approved method];

therefore, the proposed change does not

significantly increase the probability or

consequences of an accident previously

evaluated.

II. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The administrative changes in the

Technical Specification [TS] requirements do

not involve a change in the design of the

plant. The proposed requirements will

continue to ensure that fuel is not loaded into

a core cell that is associated with a removed

or withdrawn control rod.

Therefore, the proposed changes do not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

III. The proposed change does not involve

a significant reduction in a margin of safety.

The margin of safety associated with

criticality events during fuel handling is

provided by the event being a non credible

event. The proposed change will only

provide an additional means to meet the

same intent of ensuring that the event is of

such low probability as to be considered non

credible. The proposed change will provide

the additional allowance to perform fuel

loading only if an additional positive means

of assuring fuel assemblies cannot be loaded

into a core cell with a withdrawn or removed

control rod is in effect. The positive means

will entail a physical barrier such that even

if refueling procedures were violated and an

attempt was made to load a fuel assembly

into a core cell with a withdrawn or removed

control rod the action would be prevented.

This requirement provides sufficient

additional restrictions to ensure that the

event is of such low probability as to be

considered non credible.

The probability of an inadvertent criticality

occurring will continue to be precluded by

the same number of layers of administrative

controls [as the currently approved method];

therefore, this change does not reduce the

level of safety imposed by the current

Technical Specification [TS] requirements.

Therefore, the proposed changes do not

cause 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: (1) Judge George W. Armstrong

Library, 220 S. Commerce Street,

Natchez, MS 39120, for Grand Gulf

Nuclear Station and (2) Government

Documents Department, Louisiana State

University, Baton Rouge, LA 70803, for

River Bend Station.

Attorney for licensee: (1) Nicholas S.

Reynolds, Esquire, Winston and Strawn,

1400 L Street, N.W., 12th Floor,

Washington, DC 20005–3502, for Grand

Gulf Nuclear Station and (2) Mark

Wetterhahn, Esq., Winston & Strawn,

1400 L Street, N.W., Washington, DC

20005, for River Bend Station.

NRC Project Director: William D.

Beckner

Entergy Operations, Inc., et al., Docket

No. 50–416, Grand Gulf Nuclear

Station, Unit 1, Claiborne County,

Mississippi

Date of amendment request: May 9,

1996

Description of amendment request:

The amendment request would allow

allow the licensee to perform the

surveillance of the relief mode of

operation of each of the 20 safety/relief

valves (S/RVs) on the 4 main steam

lines without physically lifting the disk

off the seat at power. The proposed

changes are to Surveillance

Requirements (SRs) 3.4.4.3, Safety/

Relief Valves, 3.5.1.7, Automatic

Depressurization System Valves, and

3.6.1.6.1, Low-Low Set Valves, of the

Technical Specifications, and the

changes would state that the required

operation of the valve to verify is that

the relief-mode actuator strokes when

the valve is manually actuated. Each S/

RV is a Dikkers, 8 X 10, direct-acting,

spring loaded, safety valve with

attached pneumatic actuator for reliefmode operation. Eight of the S/RVs use

the relief mode to perform the

Automatic Depressurization System

(ADS) function. Also, six S/RVs, two of

which are also ADS S/RVs, use the relief

mode to perform the Low-Low Set valve

function. The licensee also proposed

changes to the Bases of the Technical

Specifications that are associated with

the above proposed changes.

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 Dikkers S/RV provides

pressure relief based on the principle of

vertically moving the stem that attaches

directly to the valve disk. The force that

provides the stem movement is

provided by one of two sources; the

vessel pressure directly against the force

of the stem spring (safety mode), or the

pneumatic actuator arm against the

force of the stem spring (relief mode).

ASME Boiler and Pressure Vessel Code

requires testing the safety mode of

operation once every five year operating

cycle. Once a safety valve is installed,

the safety mode is never tested while

the S/RV is installed in the plant. The

testing of the relief mode of operation

for a direct-acting S/RV provides

verification that the control functions of

electrical and pneumatic connections

have been properly reconnected, and

that the actuator arm will provide the

necessary force to operate the S/RV.

This proposed change provides verification

of proper control connections by requiring

the pneumatic and electrical controls to cycle

the actuator arm on each S/RV after

installation in the drywell. The test

population of S/RVs removed each outage for

safety setpoint testing will be tested in the

relief mode. This testing will demonstrate

that the installed S/RVs will function

properly in the relief mode. The remaining

installed S/RVs will continue to be tested for

proper system function. As presently

required by GGNS Technical Specifications

and administrative procedures, proper

operation of the solenoid control block will

be demonstrated by providing an open signal

to each S/RV, with a check to verify that each

solenoid valve repositions. Verification of

proper solenoid valve operation, in addition

to the proper relief-mode operation of the test

population, provides assurance that the S/RV

will perform as expected when control air

pressure is applied to the solenoid valve

control block.

Entergy Operations, Inc. is proposing that

the Grand Gulf Nuclear Station Operating

License be amended to perform the

surveillance of each safety relief valve (S/RV)

relief mode of operation without physically

lifting the disk off the seat at power.

During the refueling outage, a sample

population of the S/RVs will be removed for

safety-mode setpoint testing in accordance

with the GGNS IST program, using ASME

Boiler and Pressure Vessel Code, Section XI.

Each of these removed S/RVs will be tested

in the relief mode to verify that the

pneumatic actuator functions correctly, and

this test sample will be used to provide

assurance that the installed S/RV pneumatic

actuators will function properly. After the

test sample of S/RVs has been replaced with

recertified spares, and S/RV controls have

been connected, the upper stem nut that

couples the valve stem to each newlyinstalled S/RV’s pneumatic actuator will be

moved up the stem to allow an uncoupled

actuation of the relief-mode actuator. Control

air pressure to each actuator will be reduced

from normal system pressure to prevent

damaging the pneumatic relief-mode

actuator. The actuator will be remotely

operated from the control room, as required

by current test methods, and visual

verification will be performed for proper

actuator response and range of motion. After

proper actuator operation has been verified,

the upper stem nut will be returned to its

operating stem location. Verification of

proper system logic controls and function for

every installed S/RV will continue to be

performed, as required by Technical

Specifications.

The commission has provided standards

for determining whether a no significant

hazards consideration exists as stated in 10

CFR 50.92(c). A proposed amendment to an

operating license involves no significant

hazards if the operation of the facility in

accordance with the proposed amendment

would not: (1) involve a significant increase
31178 Federal Register / Vol. 61, No. 119 / Wednesday, June 19, 1996 / Notices

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.

Entergy Operations has evaluated the no

significant hazards considerations in its

request for a license amendment. In

accordance with 10 CFR 50.91(a), Entergy

Operations, Inc. is providing the following

analysis of the proposed amendment against

the three standards in 10 CFR 50.92:

a. No significant increase in the probability

or consequences of an accident previously

evaluated results from this change.

Each refueling outage, a test sample of the

population of S/RVs is removed from the

plant to perform testing as required by ASME

Boiler and Pressure Vessel Code, Section XI.

These S/RVs will be stroked in the relief

mode during as-found testing, and are

therefore verified to operate properly when

each S/RV stem is raised by the relief-mode

pneumatic actuator. This proposed

surveillance verifies proper S/RV relief-mode

operation of all installed S/RVs based upon

this test sample. This testing, in conjunction

with replacement of each S/RV prior to the

end of its expected service life, provides

reasonable assurance that the installed S/RVs

will perform as well as the test population of

S/RVs.

After the S/RVs have been replaced in the

plant, and after all controls are reconnected,

the relief-mode actuator on each newlyinstalled S/RV will be uncoupled from the S/

RV stem, and stroked. This actuator stroke

will verify that no damage has occurred to

the relief-mode actuator during S/RV

transportation from its storage location to its

operating location. The direct coupling of the

valve stem to disk provides assurance that

proper relief actuation will occur when the

actuator is operated. The safety-mode

components are completely encased within

the valve body and bonnet, which provides

a rugged structure to prevent damage to these

components. The remaining installed S/RVs

will continue to be tested for proper control

system function as previously required by

Technical Specifications. The direct coupling

of the S/RV stem to disk provides assurance

that proper relief-mode actuation will occur

when the actuator is operated. The safety

mode of the GGNS S/RVs is not affected by

a malfunction of the relief-mode components.

Blockage of each S/RV discharge line will

be prevented by the same Foreign Material

Exclusion (FME) controls that exist for other

reactor vessel and support systems. These

FME controls, combined with the horizontal

orientation of the S/RV discharge piping

mating surfaces, provide reasonable

assurance that discharge line blockage will

not occur.

Therefore, no significant increase in the

probability or consequences of an accident

previously evaluated results from this

proposed change.

b. This change would not create the

possibility of a new or different kind of

accident from any previously analyzed.

The proposed change demonstrates that

each S/RV will perform its intended reliefmode function, which is the intent of the

present surveillance. The relief mode of S/RV

operation is demonstrated to be operable

based upon successful performance of a test

population, S/RV component service life, and

existing Technical Specification

surveillances. No new failure mechanisms to

the relief- mode of operation are introduced,

as the proposed surveillance verifies relief

actuator operability. Plant FME controls,

combined with the horizontal orientation of

the S/RV discharge piping mating flange,

provides reasonable assurance that discharge

line blockage will not occur. This proposed

change does not add any new systems,

structures, or components, nor does it

introduce new S/RV operating modes.

Therefore, this change would not create the

possibility of a new or different kind of

accident from any previously analyzed.

c. This change would not involve a

significant reduction in the margin of safety.

This proposed change will verify that the

relief mode of all installed S/RVs will operate

properly based upon demonstrated relief

mode performance of a sample of S/RVs. The

failure mode of the S/RV relief function

would require a failure of either the

pneumatic actuator, lifting linkage, or

solenoid block. Each of these items has been

verified to have a service life exceeding the

replacement cycle of each S/RV. Therefore,

proper operation of a sample population of

S/RVs provides reasonable assurance that the

remaining S/RVs would perform identically,

within the original margin of expected S/RV

operability. In addition, each S/RVFEs

solenoid block and control functions will

continue to be tested and cycled each

refueling outage. The removal of the valve

stroke surveillance for all S/RVs does not

increase the possibility of valve malfunction,

since valve stroke is verified during the asfound testing of the sample population of S/

RVs. This proposed surveillance test reduces

the number of S/RV actuations, and

therefore, reduces challenges to the system

both mechanically and thermally. Also, the

proposed alternative method of testing

reduces the possibility of a stuck-open S/RV,

since this proposed method will not stroke

the S/RVs with the reactor pressurized

during reactor power operations.

Therefore, this change would not involve

a significant reduction in the margin of

safety.

Based on the above evaluation, Entergy

Operations, Inc. has concluded that operation

in accordance with the proposed amendment

involves no significant hazards

considerations.

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: Judge George W. Armstrong

Library, 220 S. Commerce Street,

Natchez, MS 39120

Attorney for licensee: Nicholas S.

Reynolds, Esquire, Winston and Strawn,

1400 L Street, N.W., 12th Floor,

Washington, DC 20005–3502

NRC Project Director: William D.

Beckner

Entergy Operations, Inc., et al., Docket

No. 50–416, Grand Gulf Nuclear

Station, Unit 1, Claiborne County,

Mississippi

Date of amendment request: May 31,

1996

Description of amendment request:

The amendment would provide an

alternative method to compensate for

inoperable refueling equipment

interlocks. The alternative method

would be to insert a control rod

withdrawal block and verify that all

control rods are fully inserted; however,

the control rods required to be inserted

would not apply to those control rods

withdrawn in accordance with LCO

3.10.6, ‘‘Multiple Control Rod

Withdrawal -Refueling.’’ The

amendment would add an additional

Required Action for Limiting Condition

for Operation (LCO) 3.9.1, ‘‘Refueling

Equipment Interlocks,’’ of the Technical

Specifications (TSs) for Grand Gulf

Nuclear Station, Unit 1 (GGNS). The

alternative method then could be used

to respond to inoperable interlocks

instead of only the current method of

halting in-vessel fuel movement with

equipment associated with the

inoperable interlock.

The proposed change does not remove

the current Required Action method for

LCO 3.9.1 and does not change the

surveillance requirements on the

refueling equipment. The licensee has

also provided changes to the Bases of

the TSs for the proposed amendment.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration. The licensee has

proposed the amendment for the TSs for

both GGNS and River Bend Station

(RBS). References made to the RBS TSs

and to RBS in the licensee’s analysis of

no significant hazards consideration

have been removed and replaced by [...].

The licensee’s analysis is presented

below:

Entergy Operations, Inc. proposes to

change the current Grand Gulf Nuclear

Station (GGNS) [...] Technical Specifications.

The specific proposed change adds

additional acceptable Required Actions to the

Actions of LCO 3.9.1, ‘‘Refueling Equipment

Interlocks,’’ [for inoperable interlocks]. The

additional Required Actions will add an

alternative [method] to [the current method

of] suspending fuel movement in the reactor

vessel when the refueling interlocks are

inoperable. The requested alternative is to

insert a control rod withdrawal block
Federal Register / Vol. 61, No. 119 / Wednesday, June 19, 1996 / Notices 31179

immediately and verify all control rods

required to be inserted are fully inserted.

[The control rods required to be inserted

would not apply to control rods withdrawn

in accordance with LCO 3.10.6, ‘‘Multiple

Control Rod Withdrawal—Refueling.’’]

The Commission has provided standards

for determining whether a no significant

hazards consideration exists as stated in 10

CFR 50.92(c). A proposed amendment to an

operating license involves no significant

hazards consideration if operation of the

facility in accordance with the proposed

amendment would not: (1) involve a

significant increase in the probability or

consequences of an accident previously

evaluated; or (2) create the possibility of a

new or different kind of accident from any

accident previously evaluated; or (3) involve

a significant reduction in a margin of safety.

Entergy Operations, Inc. has evaluated the

[criteria for] no significant hazards

consideration in its request for this license

amendment and determined that no

significant hazards consideration results from

this change. In accordance with 10 CFR

50.91(a), Entergy Operations, Inc. is

providing the analysis of the proposed

amendment against the three standards in 10

CFR 50.92(c). A description of the no

significant hazards consideration

determination follows:

I. The proposed change does not

significantly increase the probability or

consequences of an accident previously

evaluated.

The refueling interlocks are explicitly

assumed in the GGNS Updated Final Safety

Analyses Report (UFSAR) [...] analysis of the

control rod removal error or fuel loading

error during refueling. This analysis

evaluates the probability and consequences

of control rod withdrawal during refueling.

Criticality and, therefore, subsequent prompt

reactivity excursions are prevented during

the insertion of fuel, provided all control

rods are fully inserted during the fuel

insertion. The refueling interlocks

accomplish this by preventing loading fuel

into the core with any control rod

withdrawn, or by preventing withdrawal of a

rod from the core during fuel loading.

When the refueling interlocks are

inoperable the current method of preventing

the insertion of fuel when a control rod is

withdrawn is to prevent fuel movement. This

method is currently required by the

Technical Specifications. An alternate

method to ensure that fuel is not loaded into

a cell with the control rod withdrawn is to

prevent control rods from being withdrawn

and verify that all control rods required to be

inserted are fully inserted. The proposed

actions will require that a control rod block

be placed in effect thereby ensuring that

control rods are not subsequently

inappropriately withdrawn. Additionally,

following placing the control rod withdrawal

block in effect, the proposed actions will

require that all required control rods be

verified to be fully inserted. This verification

is in addition to the requirements to

periodically verify control rod position by

other Technical Specification requirements.

These proposed actions will ensure that

control rods are not withdrawn and cannot

be inappropriately withdrawn because an

electrical or hydraulic block to control rod

withdrawal is in place. Like the current

requirements the proposed actions will

ensure that unacceptable operations are

blocked (e.g., loading fuel into a cell with a

control rod withdrawn [would be blocked]).

The proposed additional acceptable

Required Actions provide the same level of

assurance that fuel will not be loaded into a

core cell with a control rod withdrawn as the

current Required Action or the Technical

Specification Surveillance Requirement.

Therefore, the proposed change does not

significantly increase the probability or

consequences of an accident previously

evaluated.

II. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The change in the Technical Specification

requirements does not involve a change in

plant design. The proposed requirements will

continue to ensure that fuel is not loaded into

the core when a control rod is withdrawn

except following the requirements of LCO

3.10.6, ‘‘Multiple Control Rod Removal—

Refueling,’’ which is unaffected by this

change.

Therefore, the proposed changes do not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

III. The proposed change does not involve

a significant reduction in a margin of safety.

As discussed in the Bases for the affected

Technical Specification requirements,

inadvertent criticality is prevented during the

insertion of fuel provided all control rods are

fully inserted during the fuel insertion. The

refueling interlocks function to support the

refueling procedures by preventing control

rod withdrawal during fuel movement and

the inadvertent loading of fuel when a

control rod is withdrawn.

The proposed change will allow the

refueling interlocks to be inoperable and fuel

movement to continue only if a control rod

withdrawal block is in effect and all required

control rods are verified to be fully inserted.

These proposed Required Actions provide

the same level of protection as the refueling

interlocks by preventing a configuration

which could lead to an inadvertent criticality

event. The refueling procedures will

continue to be supported by the proposed

required actions because control rods cannot

be withdrawn and as a result fuel cannot be

inadvertently loaded when a control rod is

withdrawn.

Therefore, the proposed changes do not

cause 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: Judge George W. Armstrong

Library, 220 S. Commerce Street,

Natchez, MS 39120

Attorney for licensee: Nicholas S.

Reynolds, Esquire, Winston and Strawn,

1400 L Street, N.W., 12th Floor,

Washington, DC 20005–3502

NRC Project Director: William D.

Beckner

Entergy Operations, Inc., et al., Docket

No. 50–416, Grand Gulf Nuclear

Station, Unit 1, Claiborne County,

Mississippi

Date of amendment request: May 31,

1996, as supplemented by letter dated

May 2, 1996.

Description of amendment request:

The amendment request would revise

the current reactor vessel material

surveillance program schedule for

GGNS. This is the schedule for

withdrawing surveillance capsules from

the reactor vessel for testing to measure

the impact of neutron irradiation of the

vessel material and is required by

Section III.B.3 of Appendix H, ‘‘Reactor

Vessel Material Surveillance Program

Requirements,’’ of 10 CFR Part 50. The

schedule must be approved by the

Nuclear Regulatory Commission (NRC)

before implementation.

For GGNS, there are three

surveillance capsules inside the reactor

vessel, each of which contains

specimens of the reactor vessel material.

The first capsule was removed from the

reactor vessel on May 7, 1995, during

the 7th refueling outage. Because no

useful data is expected from testing the

material specimens in the first capsule,

the request would allow the first

capsule to be placed back into the

vessel.

As part of revising the schedule, the

licensee is also renumbering the three

surveillance capsules so that the capsule

removed at the 7th refueling outage

becomes the third capsule when it is

placed back in the vessel. The proposed

change would, however, not extend the

time that the next capsule (the

renumbered first capsule) would be

withdrawn from the GGNS reactor

vessel.

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:

Entergy Operations, Inc., proposes to

change the withdrawal schedule for the

reactor vessel material surveillance capsules

[and renumber the capsules]. The revised

schedule for withdrawal of the surveillance

capsules is withdrawal of the first capsule at

24 Effective Full Power Years. The

withdrawal schedule for the second capsule

is to be determined at a later date. The third

capsule which was withdrawn on May 7,

1995 is to be returned to reactor vessel during
31180 Federal Register / Vol. 61, No. 119 / Wednesday, June 19, 1996 / Notices

the Fall, 1996 outage and retained as a

standby. [The current schedule for

withdrawal of the three capsules is 8 and 24

Effective Full Power Years for the first two

capsules, and the third capsule is a spare

with no specific schedule for withdrawal.]

The Commission has provided standards

for determining whether a no significant

hazards consideration exists as stated in 10

CFR 50.92(c). A proposed amendment to an

operating license involves no significant

hazards consideration if operation of the

facility in accordance with the proposed

amendment would not: (1) involve a

significant increase in the probability or

consequences of an accident previously

evaluated; or (2) create the possibility of a

new or different kind of accident from any

accident previously evaluated; or (3) involve

a significant reduction in a margin of safety.

In consideration of the October 4, 1995,

decision of the Atomic Safety and Licensing

Board concerning an amendment request

from Perry Nuclear Power Plant, Entergy

Operations, Inc. has evaluated the no

significant hazards consideration in its

request for a change to the withdrawal

schedule required by 10 CFR 50, Appendix

H, and determined that no significant

hazards consideration results from this

change. In accordance with 10 CFR 50.91(a),

Entergy Operations, Inc. is providing the

analysis of the proposed amendment against

the three standards in 10 CFR 50.92(c):

I. The proposed change does not

significantly increase the probability or

consequences of an accident previously

evaluated.

The change revises the withdrawal

schedule for the reactor vessel material

surveillance capsules and returns a

withdrawn capsule to the reactor vessel. The

capsules [only contain specimens of the

reactor vessel material and] are not an

initiator of any previously analyzed accident.

The withdrawal or return of the surveillance

capsule does not effect the probability or

consequences of any previously analyzed

accident. Extending the time for withdrawal

of the first capsule and returning the

withdrawn capsule to the vessel do not

adversely affect the pressure temperature

limit curves for the reactor vessel. Regulatory

Guide 1.99 [, ‘‘Effects of Residual Elements

on Predicted Radiation Damage to Reactor

Vessel Materials,’’] is currently used to

prepare the pressure temperature limit curves

and is inherently conservative for boiling

water reactors (BWRs)[, as GGNS]. The

current pressure temperature limit curves

will continue to be adhered to. Additionally,

[GGNS] participates in the supplemental test

program designed to significantly increase

the amount of BWR surveillance data. [This

program has supplemental capsules which

were installed in the Cooper and Oyster

Creek Nuclear Power Plants, which contain

the limiting GGNS weld and plate vessel

material, and which will be withdrawn in

1996, 2000, and 2002.] This program will be

used to complement the GGNS surveillance

program such that postponement of the

capsule withdrawals will have minimal

impact on the understanding of the

irradiation effects on the GGNS vessel.

[The licensee stated in its May 2, 1996,

letter that testing of the specimens in the

removed capsule may not provide useful

indicators of the damage to the vessel

material because the low neutron fluence on

the vessel and the good material chemistry

will result in a minimal null-ductility

temperature shift. Testing the material

specimens will destroy them; however,

placing the capsule back in the vessel will

allow the specimens to have more irradiation

until useful data could be obtained from

testing the specimens.]

Therefore, the proposed change does not

significantly increase the probability or

consequences of an accident previously

evaluated.

II. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

Returning the withdrawn capsule to the

vessel and postponing the withdrawal of the

first capsule do not contribute to the

possibility of a new or different kind of

accident or [plant] malfunction from those

previously analyzed [in the Updated Final

Safety Analysis Report for GGNS]. Failure of

the reactor vessel is not a credible accident

since the vessel itself is a highly reliable

component. This change does not affect that

determination. The potential for reactor

vessel cracking will be adequately assessed

by the proposed withdrawal schedule.

[The licensee stated in its May 2, 1996,

letter that testing of the specimens in the

removed capsule may not be useful

indicators of the damage to the vessel

material because the low neutron fluence on

the vessel and good material chemistry will

result in a minimal shift.]

In addition, the results from the

supplemental test program will provide

indication of the condition of the vessel until

the data from the first GGNS capsule[,

withdrawn and tested,] are available. The

proposed change provides the same level of

confidence in the integrity of the vessel. The

pressure temperature curves are currently

controlled by the Technical Specifications

and are determined using the conservative

methodology in Regulatory Guide 1.99.

Therefore, the possibility of failure of the

reactor vessel is not increased. The proposed

change does not involve a change in the

design of the plant. The current pressure

temperature limit curves are inherently

conservative and will continue to be adhered

to.

Therefore, the proposed changes do not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

III. The proposed change does not involve

a significant reduction in a margin of safety.

The current pressure temperature limit

curves [for the reactor vessel] are inherently

conservative and provide sufficient margin to

ensure the integrity of the reactor vessel. The

[proposed] changes do not adversely affect

these curves. The supplemental test program

will be used to complement the GGNS

surveillance program such that

postponement of the capsule withdrawal

[and testing] will have minimal impact on

the understanding of irradiation effects on

the GGNS vessel. The capsules removed in

1996 as part of the supplemental program

will have a [neutron] fluence higher than the

25% of the design life fluence used in

establishing the original GGNS [reactor vessel

material surveillance program] schedule;

therefore, the use of the supplemental test

program results will meet the intent of the

original test schedule.

Therefore, the proposed changes do not

result in a significant reduction in the margin

of safety.

Based on the above evaluation, Entergy

Operations, Inc. has concluded that operation

in accordance with the proposed change

involves no significant hazards

consideration.

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: Judge George W. Armstrong

Library, 220 S. Commerce Street,

Natchez, MS 39120

Attorney for licensee: Nicholas S.

Reynolds, Esquire, Winston and Strawn,

1400 L Street, N.W., 12th Floor,

Washington, DC 20005–3502

NRC Project Director: William D.

Beckner

Florida Power and Light Company,

Docket Nos. 50–250 and 50–251, Turkey

Point Plant Units 3 and 4, Dade County,

Florida

Dates of amendment request: March

21, 1996, and May 13, 1996

Description of amendment request:

The licensee proposed to change the

Turkey Point Units 3 and 4 Technical

Specifications (TS) to relocate the

requirements of the Radiological

Effluent Technical Specifications

(RETS) to other documents.

The proposed amendments would

relocate the LIMITING CONDITIONS

FOR OPERATION (LCO) and

SURVEILLANCE REQUIREMENTS

associated with the RETS in accordance

with GL 89–01, NUREG–1301, and

NUREG–1431, Rev. 1. The definition in

TS 1.15, ‘‘Members of the Public,’’

would be deleted since it is already

located in 10 CFR Part 20 and has been

inserted into the Offsite Dose

Calculation Manual (ODCM). The

definitions for the ODCM and Process

Control Program (PCP) would be

relocated to the Administrative Controls

section of the TS. TS 3/4.3.3.5 and the

radioactive gaseous effluent portion of

TS 3/4.3.3.6 and associated tables,

instrumentation operational conditions,

remedial actions and surveillance

requirements would be controlled

through the ODCM or PCP and

associated procedures. Technical
Federal Register / Vol. 61, No. 119 / Wednesday, June 19, 1996 / Notices 31181

Specification Administrative Control

sections would contain the

programmatic controls for the ODCM

and PCP. The remaining portion of TS

3.3.3.6 would retain the operational

conditions, remedial actions, and

surveillance requirements for the

explosive gas monitor instrumentation.

The procedural details of the current

TS on radioactive effluents and

radiological environmental monitoring

would be deleted. Associated

operational conditions, remedial actions

and surveillance requirements presently

in the Technical Specifications would

be controlled through the ODCM or

PCP.

Administrative changes to the TS

were also proposed due to paragraph

and section numbering changes and

relocations associated with the

proposed technical changes.

New sections TS 6.8.4f and 6.8.4g

were proposed to provide programmatic

controls for the Radiological Effluents

Controls Program and the Radiological

Environmental Monitoring Program.

TS 6.9.1.3 and TS 6.9.1.4 would be

simplified and the reporting details now

contained in these specifications would

be relocated to the ODCM or PCP with

the exception of the requirement to

report licensee-initiated changes to the

PCP in the Annual Radioactive Effluent

Release Report.

New record retention requirements

changes for the ODCM and PCP would

be added to TS 6.10.3q.

In summary, as provided in the

guidance, the current technical content

of the specifications which would be

transferred to the ODCM or the PCP.

New programmatic controls for

radioactive effluents and radioactive

effluent monitoring would be added to

the TS, as well as further clarification to

the definitions of the ODCM and PCP.

The Technical Specification

requirements for Gas Decay Tanks and

Explosive Gas Mixture would be

relocated to the Plant Systems section of

the TS.

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

amendments would not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

The changes being proposed are

administrative in nature in that they relocate

Technical Specification requirements

associated with RETS from the Technical

Specifications to the ODCM or PCP. These

changes are in accordance with the

recommendations contained in GL 89–01,

NUREG 1301, and NUREG 1431 Rev. 1. The

only change being made to existing

requirements or commitments are

administrative in nature. The proposed

changes do not involve any change to the

configuration or method of operation of any

plant equipment that is used to mitigate the

consequences of an accident, nor do they

affect any assumptions or conditions in any

of the accident analyses. Since the accident

analyses remain bounding, their probability

or consequences are not adversely affected.

Therefore, the probability or consequences of

an accident previously evaluated are not

affected.

(2) Operation of the facility in accordance

with the proposed amendments would not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

The changes being proposed are

administrative in nature in that they relocate

Technical Specification requirements

associated with RETS from the Technical

Specifications to the ODCM or PCP. These

changes are in accordance with the

recommendations contained in GL 89–01,

NUREG 1301, and NUREG 1431, Rev. 1. The

only change being made to existing

requirements or commitments are

administrative in nature. The proposed

changes do not involve any change to the

configuration or method of operation of any

plant equipment used to mitigate the

consequences of an accident.

Therefore, the possibility of a new or

different kind of accident from any accident

previously evaluated would not be created.

(3) Operation of the facility in accordance

with the proposed amendments would not

involve a significant reduction in a margin of

safety.

The changes being proposed are

administrative in nature in that they relocate

Technical Specification requirements

associated with RETS from the Technical

Specifications to the ODCM or PCP. These

changes are in accordance with the

recommendations contained in GL 89–01,

NUREG 1301, and NUREG 1431, Rev. 1. The

only change being made to existing

requirements or commitments are

administrative in nature. All technical

content is preserved. The operating limits

and functional capabilities of the affected

systems, structures, and components are

unchanged by the proposed amendments.

Therefore, a significant reduction in a

margin of safety would not be involved.

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: Florida International

University, University Park, Miami,

Florida 33199

Attorney for licensee: J. R. Newman,

Esquire, Morgan, Lewis & Bockius, 1800

M Street, NW., Washington, DC 20036

NRC Project Director: Frederick J.

Hebdon

Florida Power and Light Company,

Docket Nos. 50–250 and 50–251, Turkey

Point Plant Units 3 and 4, Dade County,

Florida

Dates of amendment request: May 28,

1996

Description of amendment request:

The licensee proposed to change the

Turkey Point Units 3 and 4 Technical

Specifications (TS) to change the

licensed qualifications of the Operations

Manager. The proposed change would

delete the qualification option that the

Operations Manger could have held a

Senior Reactor Operator License on a

boiling water reactor and replace it with

an option that this individual could

have completed the Turkey Point

Nuclear Plant Senior Management

Operation Training Course (i.e., certified

at an appropriate simulator for

equivalent senior operator knowledge

level).

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.

The change being proposed is

administrative in nature, addresses

organizational personnel qualification issues,

and does not affect assumptions contained in

plant safety analyses, the physical design

and/or operation of the plant, or Technical

Specifications that preserve safety analysis

assumptions.

The individual Florida Power & Light

Company (FPL) chooses to fill the position of

Operations Manager will have extensive

educational and management- level nuclear

power experience meeting the criteria of

ANSI N18.1–1971. The Operations

Supervisor and Nuclear Plant Supervisors

maintain SRO licenses on Turkey Point. The

current Technical Specifications do not

require the Operations Manager to hold an

SRO License at Turkey Point. The current

Technical Specifications permit the

Operations Manager to have held an SRO

License on another plant. The proposed

change will continue to require that the

Operations Manager has completed the

Turkey Point Nuclear Plant Senior

Management Operations Training Course if

the incumbent did not previously hold an

SRO license. The Turkey Point Nuclear Plant

Senior Management Operations Training

Course ensures that the Operations Manager

has the training on plant-specific systems
31182 Federal Register / Vol. 61, No. 119 / Wednesday, June 19, 1996 / Notices

and procedures at Turkey Point and a

knowledge level equivalent to the license

requirements for operations management.

The on-shift Operations’ organization is,

and will continue to be, supervised and

directed by the Operations Supervisor, who

is currently required by Technical

Specification 6.2.2.h. to hold an SRO

License.

Additionally, the proposed changes do not

impact or change, in any way, the minimum

on-shift manning or qualifications for those

individuals responsible for the actual

licensed operation of the facility as required

by 10 CFR 50.54(l).

Based on the above, the proposed changes

do not affect the probability or consequences

of accidents previously analyzed.

(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 change being proposed is

administrative in nature, addresses personnel

qualification issues, does not affect

assumptions contained in plant safety

analyses, the physical design and/or

operation of the plant, or Technical

Specifications that preserve safety analysis

assumptions.

The proposed changes address

organizational and qualifications issues

related to the criteria used for assignment of

individuals to the Operations organization

off-shift management chain of command.

Since the proposed change does not impact

or change, in any way, the minimum on-shift

manning or qualifications for those

individuals responsible for the actual

licensed operation of the facility, 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.

(3) Operation of the facility in accordance

with the proposed amendment would not

involve a significant reduction in a margin of

safety.

The proposed change addresses

organizational and qualification issues

related to the criteria used for assignment of

individuals to the Operations organization

off-shift management chain of command. The

proposed change does not impact or change,

in any way, the minimum on-shift manning

or qualifications for those individuals

responsible for the actual licensed operation

of the facility.

FPL’s operating organization at Turkey

Point Plant is shown on Figure 1–2,

Appendix A of the NRC-approved FPL

Topical Quality Assurance Report (TQAR).

Since changes to the TQAR are governed by

10 CFR § 50.54(a)(3), any changes to the

TQAR that reduce commitments previously

accepted by the NRC require approval by the

NRC prior to implementation.

While the Operations Manager is

responsible for the plant’s operating

organization, his responsibilities also include

management of the plant’s Health Physics

and Chemistry departments. The Operations

organization is supervised and directed by

the Operations Supervisor, who is required

by Technical Specification 6.2.2.h. to hold a

Senior Reactor Operator License. The Turkey

Point Units 3 and 4 Technical Specifications

do not require that the Operations Manager

maintain an SRO License (nor even that the

incumbent has ever held a Senior Reactor

Operator License at Turkey Point). The

Turkey Point Technical Specification 6.3.1,

FACILITY STAFF QUALIFICATIONS, will

ensure that, other than license certification,

the individual filling the Operations Manager

position has the requisite education, training,

and experience for the management position.

As a result, operation of the facility in

accordance with the proposed amendment

would 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 request

involves no significant hazards

consideration.

Local Public Document Room

location: Florida International

University, University Park, Miami,

Florida 33199

Attorney for licensee: J. R. Newman,

Esquire, Morgan, Lewis & Bockius, 1800

M Street, NW., Washington, DC 20036

NRC Project Director: Frederick J.

Hebdon

GPU Nuclear Corporation and Saxton

Nuclear Experimental Corporation,

Docket No. 50–146, Saxton Nuclear

Experimental Facility (SNEF), Bedford

County, Pennsylvania

Date of amendment request: February

2, 1996, as supplemented on February

28, April 24 and May 24, 1996.

Description of amendment request:

The proposed amendment would (1)

increase the scope of work permitted

within the exclusion area at the SNEF

to include action preparatory to major

component and facility

decommissioning limited to asbestos

removal, removal of defunct plant

electrical services, and installation of

decommissioning support facilities and

systems such as heating, ventilation,

and air conditioning,

(2) eliminate administrative access controls

requiring that the grating covering the

auxiliary compartment stairwell and rod

room door remain locked except for

authorized entry, and (3) revise the facility

layout diagram to allow the exclusion area to

consist of, at a minimum, the containment

vessel, and at a maximum, extend to the

SNEF outer security fence, and to include on

the diagram the footprint of the proposed

decommissioning support facilities.

Basis for proposed no significant

hazards Consideration Determination:

As required by 10 CFR 50.91(a), the

licensees have provided their analysis of

the issue of no significant hazards

consideration, which is presented

below:

The proposed changes do not involve

a significant hazards considerations

because the changes would not:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The SNEF ended power operation in May

1972, and the reactor core has been removed.

In its present condition, the only accidents

applicable to the site are fire, flooding, and

radiological hazard. The additional activities

associated with the expansion of the

permissible work scope will not involve a

significant increase in the probability or

consequences of a fire. There is no effect on

the probability or consequences of flooding

nor would there be a significant increase in

the probability or consequences of an offsite

radiological hazard. The relocation of

administratively controlled accesses in

accordance with the revised wording and the

proposed clarification of the facility layout

diagram would have no affect on analyzed

accidents. Activities associated with the

construction of the decommissioning support

facilities and the existence of the completed

buildings depicted on the revised figure will

not involve a significant increase in the

probability or consequences of a fire, flood,

or radiological hazard. The proposed changes

identified by this technical specification

change request do not involve a significant

increase in the probability or consequences

of an accident previously evaluated.

2. Create the possibility of a new or

different kind of accident from any

previously analyzed.

For the reasons discussed in 1 above, the

possibility of a new or different kind of

accident from any accident previously

evaluated will not be created by the

performance of the activities delineated in

the proposed revised technical specifications.

There is similarly no possibility of a new or

different kind of accident from any accident

previously evaluated that would result from

relocation of administratively controlled

accesses within the containment vessel; from

the flexibility to relocate/modify the

exclusion area fence or from the

identification of the footprint, construction

and existence of the completed

decommissioning support facilities.

3. Involve a significant reduction in a

margin of safety.

For the reasons discussed in 1 above, none

of the proposed changes involve a significant

reduction in a margin of safety.

The NRC staff has reviewed the

analysis of the licensees 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: Saxton Community Library,

911 Church Street, Saxton,

Pennsylvania 16678 Attorney for the

Licensee: Ernest L. Blake, Jr., Esquire,

Shaw, Pittman, Potts, and Trowbridge,

2300 N Street, NW, Washington, DC

20037
Federal Register / Vol. 61, No. 119 / Wednesday, June 19, 1996 / Notices 31183

NRC Project Director: Seymour H.

Weiss

Gulf States Utilities Company, Cajun

Electric Power Cooperative, and

Entergy Operations, Inc., Docket No.

50–458, River Bend Station, Unit 1,

West Feliciana Parish, Louisiana

Date of amendment request: May 20,

1996

Description of amendment request:

The proposed amendment would revise

the Facility Operating License No. NPF–

47 and Appendix C to the license to

reflect the name change from Gulf States

Utilities Company to Entergy Gulf

States, Inc.

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:

I. The proposed change does not

significantly increase the probability or

consequences of an accident previously

evaluated.

The proposed change documents changing

the legal name of the company. The proposed

change will not affect any other obligations.

The company will still own all of the same

assets, serve the same customers, and all

existing obligations and commitments will

continue to be honored.

Therefore, the proposed change does no

significantly increase the probability or

consequences of an accident previously

evaluated.

II. The proposed change does not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The administrative changes in the

Operating License requirements do not

involve any change in the design of the plant.

Therefore, the proposed changes do not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

III. The proposed change does not involve

a significant reduction in a margin of safety.

The proposed change is administrative in

nature, as described above, therefore, this

change does not reduce the level of safety

imposed by any current requirements.

Therefore, the proposed changes do not

cause 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: Government Documents

Department, Louisiana State University,

Baton Rouge, LA 70803

Attorney for licensee: Mark

Wetterhahn, Esq., Winston & Strawn,

1400 L Street, N.W., Washington, D.C.

20005

NRC Project Director: William D.

Beckner

Northeast Nuclear Energy Company

(NNECO), Docket No. 50–245, Millstone

Nuclear Power Station, Unit 1, New

London County, Connecticut

Date of amendment request: April 25,

1996

Description of amendment request:

The change modifies the calibration

requirement for the source range

monitors and intermediate range

monitors by noting that the sensors are

excluded.

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:

Pursuant to 10 CFR 50.92, NNECO has

reviewed the proposed change and concludes

that the change does not involve a significant

hazards consideration (SHC) since the

proposed change satisfies the criteria in 10

CFR 50.92(c). That is, the proposed change

does not:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

By removing the requirement for sensor

calibration the function and safety

performance of these systems will not be

affected. Existing surveillances, operator

verification of overlap and system interlocks

ensure correct system performance without

sensor calibration.

Therefore, based on the above, the

proposed change to the Technical

Specifications does not involve a significant

increase in the probability or consequences

of any previously analyzed accident.

2. Create the possibility of a new or

different kind of accident from any accident

previously evaluated.

This change does not cause the source

range monitors (SRM) or the intermediate

range monitors (IRM) to function any

differently than intended by design and,

therefore, does not create the possibility of a

new or different kind of accident. The

Technical Specification change deletes a

Technical Specification requirement which

could not literally be complied with for one

component and that has no effect on the

functional performance of the SRMs or IRMs.

Therefore, this change will not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

3. Involve a significant reduction in a

margin of safety.

This change corrects a Technical

Specification requirement which could not

literally be complied with for one component

and that has no effect on the functional

performance of the SRMs or IRMs.

Instrument calibrations and functional

checks are still performed during each

refueling outage to assure adequate system

performance.

Therefore, this change has no impact on

the margin to 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: Learning Resources Center,

Three Rivers Community-Technical

College, 574 New London Turnpike,

Norwich, CT 06360, and Waterford

Library, ATTN: Vince Juliano, 49 Rope

Ferry Road, Waterford, CT 06385.

Attorney for licensee: Lillian M.

Cuoco, Esq., Senior Nuclear Counsel,

Northeast Utilities Service Company,

P.O. Box 270, Hartford, CT 06141–0270.

NRC Project Director: Phillip F.

McKee

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:

February 14, 1996

Description of amendment requests:

The proposed amendments would

revise the combined Technical

Specifications (TS) for the Diablo

Canyon Power Plant (DCPP), Unit Nos.

1 and 2, to revise 30 TS and add two

new TS surveillance requirements to

support implementation of extended

fuel cycles at DCPP, Unit Nos. 1 and 2.

The specific TS changes proposed

include those for 9 trip actuating device

tests, 12 fluid system actuation tests,

and 11 miscellaneous tests. Two of the

fluid system actuation tests are

proposed new TS surveillance

requirements. The TS changes also

include the addition of a new frequency

notation, ‘‘R24, REFUELING

INTERVAL,’’ to Table 1.1 of the TS.

Also, a revision that applies to all

subsequent TS changes involves

revising the Bases section of TS 4.0.2 to

change the surveillance frequency from

an 18-month surveillance interval to at

least once each refueling interval.

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
31184 Federal Register / Vol. 61, No. 119 / Wednesday, June 19, 1996 / Notices

consequences of an accident previously

evaluated.

The surveillance interval notation addition

in TS Table 1.1 and the updated TS 4.0.2

Bases section are administrative changes that

do not affect the probability or consequences

of accidents.

The 30 proposed TS surveillance interval

increases from 18 to 24 months do not alter

the intent or method by which the

inspections, tests, or verifications are

conducted, do not alter the way any

structure, system, or component functions,

and do not change the manner in which the

plant is operated. The surveillance,

maintenance, and operating histories indicate

that the equipment will continue to perform

satisfactorily with longer surveillance

intervals. Few surveillance and maintenance

problems were identified. No problems

recurred, with the exception of those

associated with the pressurizer heater

emergency breakers, which will continue to

be surveilled on a quarterly frequency until

they are replaced.

There are no known mechanisms that

would significantly degrade the performance

of the evaluated equipment during normal

plant operation. All potential time-related

degradation mechanisms have insignificant

effects in the timeframe of interest (24

months +25 percent, or 30 months). Based on

the past performance of the equipment, the

probability or consequences of accidents

would not be significantly affected by the

proposed surveillance interval increases.

The 24-month surveillance intervals for the

two new TS proposed to verify that the CCW

[component cooling water] and ASW

[auxiliary saltwater] pumps will start

automatically are based on an evaluation of

historical operation, maintenance, and

surveillance data for the pumps. These

historical data are available because the

pumps have been operated, maintained, and

tested on 18- month intervals in accordance

with procedures since initial plant startup.

These new surveillances represent additional

TS requirements to ensure the CCW and

ASW pumps start when required. No known

degradation mechanisms would significantly

affect the ability of the pumps to start over

the timeframe of interest (30 months

maximum). Based on the past performance of

the equipment, these proposed new TS

would not affect the probability or

consequences of accidents.

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 surveillance interval notation addition

in TS Table 1.1 and the updated TS 4.0.2

Bases section are administrative changes that

do not affect the type of accidents possible.

For the 30 proposed TS changes involving

surveillance interval increases from 18 to 24

months, the surveillance and maintenance

histories indicate that the equipment will

continue to effectively perform its design

function over the longer operating cycles.

Additionally, the increased surveillance

intervals do not result in any physical

modifications, affect safety function

performance or the manner in which the

plant is operated, or alter the intent or

method by which surveillance tests are

performed. Only a few problems have been

identified and generally have not recurred.

All potential time-related degradations have

insignificant effects in the timeframe of

interest. The proposed surveillance interval

increases would not affect the type of

accidents possible.

The 24-month surveillance intervals for the

two new TS proposed to verify starting of the

CCW and ASW pumps are based on an

evaluation of historical operation,

maintenance, and surveillance data. These

new TS represent additional requirements to

ensure the CCW and ASW pumps start when

required. No known degradation mechanisms

would significantly affect the ability of the

pumps to start over the timeframe of interest.

These proposed new TS would not affect the

type of accidents possible.

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 surveillance interval notation addition

in TS Table 1.1 and the updated TS 4.0.2

Bases section are administrative changes that

do not affect the margin of safety.

For the 30 proposed TS changes involving

18- to 24-month surveillance interval

increases, evaluation of historical

surveillance and maintenance data indicates

there have been only a few problems

experienced with the evaluated equipment.

There are no indications that potential

problems would be cycle-length dependent

or that potential degradation would be

significant for the timeframe of interest and,

therefore, increasing the surveillance interval

will have little, if any, impact on safety.

There is no safety analysis impact since these

changes will have no effect on any safety

limit, protection system setpoint, or limiting

condition for operation, and there are no

hardware changes that would impact existing

safety analysis acceptance criteria. Safety

margins would not be significantly affected

by the proposed surveillance interval

increases.

As previously noted, the 24-month

surveillance intervals for the two new TS are

based on an evaluation of historical data,

represent additional requirements, and are

not believed to be significantly affected by

potential time-dependent degradation. As

such, these proposed new TS would not

affect any margin of safety.

Therefore, the proposed changes do 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: Christopher J.

Warner, Esq., Pacific Gas and Electric

Company, P.O. Box 7442, San

Francisco, California 94120

NRC Project Director: William H.

Bateman

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: May 9,

1996

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

by revising Technical Specifications

(TS) 3/4.3.2, ‘‘Engineered Safety

Features Actuation System

Instrumentation,’’ and 3/4.6.2,

‘‘Containment Spray System.’’ The

changes would clarify the description of

the initiation signal required for

operation of the containment spray

system at Diablo Canyon Power Plant

(DCPP) and correctly incorporate

changes made in previous license

amendments. All of the changes are

administrative in nature.

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.

Revising the description of the

containment spray (CS) initiating signal

clarifies the design of the plant and provides

uniformity across the Technical

Specifications (TS) associated with the CS

initiation function. The enhanced description

does not affect system operation or

performance, nor the probability of any event

initiators. The changes do not affect any

engineered safety feature actuation setpoints

or accident mitigation capabilities.

The administrative changes to TS 3/4.3.2,

Table 4.3–2, correct the column headings and

restore test frequency notation. The changes

only revise the TS to correspond with

previously issued license amendments (LAs).

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.
Federal Register / Vol. 61, No. 119 / Wednesday, June 19, 1996 / Notices 31185

The administrative changes in the

description of the CS initiating signal provide

uniformity across the TS associated with the

spray system. There are no design, operation,

maintenance, or testing changes associated

with the administrative changes.

The administrative changes to TS 3/4.3.2,

Table 4.3–2, correct the column headings and

restore test frequency notation. The changes

only revise the TS to correspond with

previously issued LAs.

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 administrative changes in CS signal

description are not associated with any

design, operation, maintenance, or testing

revisions.

The administrative changes to TS 3/4.3.2,

Table 4.3–2, correct the column headings and

restore test frequency notation. The changes

only revise the TS to correspond with

previously issued LAs.

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: Christopher J.

Warner, Esq., Pacific Gas and Electric

Company, P.O. Box 7442, San

Francisco, California 94120

NRC Project Director: William H.

Bateman

Tennessee Valley Authority, Docket

Nos. 50–259, 50–260, and 50–296,

Browns Ferry Nuclear Plant, Units 1, 2,

and 3, Limestone County, Alabama

Date of amendment request: May 20,

1996 (TS 373)

Description of amendment request:

The proposed amendment revises the

technical specifications to incorporate a

24-hour delay in implementing the

action requirements due to a missed

surveillance requirement when the

action requirements provide a

restoration time that is less than 24

hours. This change also clarifies that the

time limit of the action requirements

applies from the point in time it is

identified a surveillance has not been

performed and not at the time that the

allowed surveillance interval was

exceeded. The licensee claims this

amendment is consistent with generic

guidance.

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:

A. The proposed amendment does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed amendment to TS definition

1.0.LL is in accordance with the guidance of

GL 87–09 and NUREG 1433, Revision 1. The

proposed change will allow BFN to continue

operation for an additional 24 hours after

discovery of a missed surveillance. The

change being proposed does not affect the

precursor for any accident or transient

analyzed in Chapter 14 of the BFN Updated

Final Safety Analysis Report. The proposed

change does not reflect a revision to the

physical design and/or operation of the plant.

Therefore, operation of the facility in

accordance with the proposed change does

not affect the probability or consequences of

an accident previously evaluated.

B. The proposed amendment does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

The proposed amendment to TS definition

1.0.LL is in accordance with the guidance of

GL 87–09 and NUREG 1433, Revision 1. The

proposed change will allow the plant to

continue operation for an additional 24 hours

after discovery of a missed surveillance. The

change being proposed will not change the

physical plant or the modes of operation

defined in the facility license. The change

does not involve the addition or modification

of equipment, nor do they alter the design or

operation of plant systems. Therefore,

operation of the facility in accordance with

the proposed change does not create the

possibility of a new or different kind of

accident from any previously evaluated.

C. The proposed amendment does not

involve a significant reduction in a margin of

safety.

The proposed amendment to TS definition

1.0.LL is in accordance with the guidance of

GL 87–09 and NUREG 1433, Revision 1. The

proposed change does not affect plant safety

analysis or change the physical design or

operation of the plant. The proposed change

will allow the plant up to 24 hours to

perform a missed surveillance. The overall

effect is a net gain in plant safety by avoiding

unnecessary shutdowns and the associate

system transients due to missed surveillance.

Therefore, operation of the facility in

accordance with 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: Athens Public Library, South

Street, Athens, Alabama 35611

Attorney for licensee: General

Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET llH,

Knoxville, Tennessee 37902

NRC Project Director: Frederick J.

Hebdon

Wisconsin Public Service Corporation,

Docket No. 50–305, Kewaunee Nuclear

Power Plant, Kewaunee County,

Wisconsin

Date of amendment request: May 8,

1996

Description of amendment request:

The proposed amendment would revise

Kewaunee Nuclear Power Plant (KNPP)

Technical Specification (TS) 5.3,

‘‘Reactor,’’ and TS 5.4, ‘‘Fuel Storage,’’

by removing the enrichment limit for

reload fuel and imposing fuel storage

restrictions on the spent fuel storage

racks and the new fuel storage racks.

The revised TS are structured consistent

with the Westinghouse Standard

Technical Specifications and the fuel

storage restrictions are based on the

criticality analyses used to support TS

Amendment 92 dated March 7, 1991.

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 proposed changes were reviewed in

accordance with the provisions of 10 CFR

50.92 to determine that no significant

hazards exist. The proposed changes will not:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The criticality analysis which was

performed in support of Technical

Specification Amendment 92, dated March 7,

1991, demonstrated that adequate margins to

criticality can be maintained with fuel

enrichments up to 49.2 grams of U235 per

axial centimeter stored in the New Fuel

Storage Racks and enrichments up to 52.3

grams of U235 per axial centimeter stored in

the Spent Fuel Storage Racks.

The bounding cases of the analysis

demonstrated that keff remains less than 0.95

in the Spent Fuel Storage Racks and the New

Fuel Storage Racks if flooded with unborated

water. The bounding cases of the analysis

also demonstrated that keff remains less than

0.98 in the New Fuel Storage Racks if

moderated by optimally misted moderator.

Therefore, the 49.2 grams of U235 per axial

centimeter enrichment is acceptable for

storage in the New Fuel Storage Racks and

52.3 grams of U235 per axial centimeter for

storage in the Spent Fuel Storage Racks.

The only other accident that needs to be

considered is a fuel handling accident. Since

the mass of the fuel assembly would not be

appreciably altered by the increased fuel
31186 Federal Register / Vol. 61, No. 119 / Wednesday, June 19, 1996 / Notices

enrichment, the probability of this accident

occurring is not changed. The consequences

of a fuel handling accident also would not be

affected by the use of higher fuel enrichment

since the fission product inventories in a fuel

assembly are not a significant function of

initial fuel enrichment. This accident was

analyzed in the criticality analysis which was

performed in support of Technical

Specification Amendment 92, dated March 7,

1991.

It should be noted that any changes in the

nuclear properties of the reactor core that

may result from higher fuel enrichments

would be analyzed in the appropriate reload

analysis.

The administrative relocation of

information to licensee controlled documents

(i.e., USAR) conforms to NRC policy for the

content of technical specifications and does

not increase the probability or consequences

of an accident.

2. Create the possibility of a new or

different kind of accident from any accident

previously evaluated.

As discussed above, the only safety issue

significantly affected by the proposed change

is the criticality analysis of the Spent Fuel

Storage Racks and the New Fuel Storage

Racks. Since it has been demonstrated that

kG≅2eff remains below 0.95 and 0.98,

respectively, in those areas, no new or

different accident would be created through

the use of fuel enrichments up to 52.3 grams

of U235 per axial centimeter at the Kewaunee

Nuclear Power Plant. Administrative controls

will ensure that only fuel enriched to 49.2

grams of U235 per axial centimeter or less will

be placed into the New Fuel Storage Racks.

The relocation of information to licensee

controlled documents does not create the

possibility of a new or different kind of

accident.

3. Involve a significant reduction in the

margin of safety.

Since the criticality analyses have shown

that increasing the allowable weight percent

enrichment to 52.3 grams of U235 per axial

centimeter would not increase keff above 0.95

in the Spent Fuel Storage Racks and

increasing the allowable weight percent

enrichment to 49.2 grams of U235 per axial

centimeter would not increase keff above 0.98

in the New Fuel Storage Racks, it is

concluded that this proposed change would

not reduce the margin of safety. Any changes

in the nuclear properties of the reactor core

that may result from higher fuel enrichments

would be analyzed in the appropriate reload

analysis to ensure compliance with

applicable reload considerations and

requirements.

Relocation of information to licensee

controlled documents is an administrative

action and therefore does not reduce 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 Wisconsin,

Cofrin Library, 2420 Nicolet Drive,

Green Bay, Wisconsin 54311–7001

Attorney for licensee: Bradley D.

Jackson, Esq., Foley and Lardner, P. O.

Box 1497, Madison, Wisconsin 53701–

1497

NRC Project Director: Gail H. Marcus

Previously Published Notices Of

Consideration Of Issuance Of

Amendments To Facility Operating

Licenses, Proposed No Significant

Hazards Consideration Determination,

And Opportunity For A Hearing

The following notices were previously

published as separate individual

notices. The notice content was the

same as above. They were published as

individual notices either because time

did not allow the Commission to wait

for this biweekly notice or because the

action involved exigent circumstances.

They are repeated here because the

biweekly notice lists all amendments

issued or proposed to be issued

involving no significant hazards

consideration.

For details, see the individual notice

in the Federal Register on the day and

page cited. This notice does not extend

the notice period of the original notice.

Houston Lighting & Power Company,

City Public Service Board of San

Antonio, Central Power and Light

Company, City of Austin, Texas, Docket

Nos. 50–498 and 50–499, South Texas

Project, Units 1 and 2, Matagorda

County, Texas

Date of amendment request: May 17,

1996

Description of amendment request:

The proposed amendments would

modify Technical Specification Section

3/4.4.5, Steam Generators, 3/4.4.6,

Reactor Coolant System Leakage, and

associate Bases to allow the installation

of tube sleeves as an alternative to

plugging to repair defective steam

generator tubes.

Date of individual notice in the

Federal Register: May 29, 1996 (61 FR

26936)

Expiration date of individual notice:

June 28, 1996

Local Public Document Room

location: Wharton County Junior

College, J. M. Hodges Learning Center,

911 Boling Highway, Wharton, TX

77488 Washington Public Power Supply

System, Docket No. 50–397, Nuclear

Project No. 2, Benton County,

Washington

Date of application for amendment:

April 24, 1996

Brief description of amendment

request: The proposed amendment

would modify Technical Specifications

(TSs) 5.3.1 and 6.9.3.2 to reflect use of

new fuel obtained from ABB/

Combustion Engineering, and to

incorporate staff-approved core reload

analysis computer programs (codes).

Date of individual notice in Federal

Register: May 1, 1996 (61 FR 19326)

Expiration date of individual notice:

May 31, 1996

Local Public Document Room

location: Richland Public Library, 955

Northgate Street, Richland, Washington

99352

Notice Of Issuance Of Amendments To

Facility Operating Licenses

During the period since publication of

the last biweekly notice, the

Commission has issued the following

amendments. The Commission has

determined for each of these

amendments that the application

complies with the standards and

requirements of the Atomic Energy Act

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

Commission’s rules and regulations.

The Commission has made appropriate

findings as required by the Act and the

Commission’s rules and regulations in

10 CFR Chapter I, which are set forth in

the license amendment.

Notice of Consideration of Issuance of

Amendment to Facility Operating

License, Proposed No Significant

Hazards Consideration Determination,

and Opportunity for A Hearing in

connection with these actions was

published in the Federal Register as

indicated.

Unless otherwise indicated, the

Commission has determined that these

amendments satisfy the criteria for

categorical exclusion in accordance

with 10 CFR 51.22. Therefore, pursuant

to 10 CFR 51.22(b), no environmental

impact statement or environmental

assessment need be prepared for these

amendments. If the Commission has

prepared an environmental assessment

under the special circumstances

provision in 10 CFR 51.12(b) and has

made a determination based on that

assessment, it is so indicated.

For further details with respect to the

action see (1) the applications for

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

the Commission’s related letter, Safety

Evaluation and/or Environmental

Assessment as indicated. All of these

items are available for public inspection

at the Commission’s Public Document

Room, the Gelman Building, 2120 L

Street, NW., Washington, DC, and at the

local public document rooms for the

particular facilities involved.
Federal Register / Vol. 61, No. 119 / Wednesday, June 19, 1996 / Notices 31187

Arizona Public Service Company, et al.,

Docket Nos. STN 50–528, STN 50–529,

and STN 50–530, Palo Verde Nuclear

Generating Station, Units 1, 2, and 3,

Maricopa County, Arizona

Date of application for amendments:

January 5, 1996, as supplemented by

letters dated April 19, May 1, and May

10, 1996.

Brief description of amendments: The

amendments revise the operating

licenses and Technical Specification

(TS) Section 1.26 to increase the

authorized rated thermal power. The

amendments also revise TS 4.1.1.4,

3.1.3.4, and 3.2.6 (Figure 3.2–1) to lower

the allowable reactor coolant system

cold leg temperature limits for each of

the three Palo Verde Nuclear Generating

Station units, and TS 3.4.2.1 and 3.4.2.2

to lower the pressurizer safety valve

setpoints for Units 1 and 3 to support

the increased power operation. The Unit

2 pressurizer safety valve setpoints in

TS 3.4.2.1 and 3.4.2.2 were revised in

Amendment 78, approved March 28,

1995, to the same values being

requested for Units 1 and 3 in this

submittal.

Date of issuance: May 23, 1996

Effective date: May 23, 1996, to be

implemented for Unit 1 within 30 days

of issuance; to be implemented for Unit

2 within 30 days of issuance; to be

implemented for Unit 3 within 45 days

as of the date of issuance, except for the

pressurizer safety valve setpoints

change which are effective prior to

startup from Unit 3’s sixth refueling

outage.

Amendment Nos.: Unit 1 - 108; Unit

2 - 100; Unit 3 - 80

Facility Operating License Nos. NPF–

41, NPF–51, and NPF–74: The

amendments revised the Operating

Licenses and Technical Specifications.

Date of initial notice in Federal

Register: February 28, 1996 (61 FR

7544) The April 19, May 1, and May 10,

1996, supplemental letters 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 May 23, 1996. No

significant hazards consideration

comments received: No.

Local Public Document Room

location: Phoenix Public Library, 1221

N. Central Avenue, Phoenix, Arizona

85004

Carolina Power & Light Company,

Docket No. 50–261, H. B. Robinson

Steam Electric Plant, Unit No. 2,

Darlington County, South Carolina

Date of application for amendment:

January 31, 1996.

Brief description of amendment: This

amendment revises the Technical

Specifications Section 4.4 to allow the

use of 10 CFR Part 50, Appendix J,

Option B, Performance-Based

Containment Leakage Rate Testing.

Date of issuance: May 28, 1996

Effective date: May 28, 1996

Amendment No. 169

Facility Operating License No. DPR–

23. Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: February 28, 1996 (61 FR

7545) The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

May 28, 1996. No significant hazards

consideration comments received: No

Local Public Document Room

location: Hartsville Memorial Library,

147 West College Avenue, Hartsville,

South Carolina 29550

Duke Power Company, et al., Docket

Nos. 50–413 and 50–414, Catawba

Nuclear Station, Units 1 and 2, York

County, South Carolina

Date of application for amendments:

November 15, 1995, as supplemented by

letters dated March 15, and April 10,

1996

Brief description of amendments: The

amendments revise the Technical

Specifications and the associated Bases

to increase the setpoint tolerance of the

main steam safety valves (MSSVs) from

plus or minus 1% to plus or minus 3%,

to incorporate a requirement to reset the

as-left MSSV lift settings to within plus

or minus 1% following surveillance

testing, and to delete two obsolete

footnotes.

Date of issuance: May 31, 1996

Effective date: As of the date of

issuance to be implemented within 30

days

Amendment Nos.: 146 and 140

Facility Operating License Nos. NPF–

35 and NPF–52: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: December 20, 1995 (60 FR

65676). The March 15 and April 10,

1996 letters provided clarifying

information that did not change the

scope of the November 15, 1995

application and the initial proposed no

significant hazards consideration

determination. The Commission’s

related evaluation of the amendments is

contained in a Safety Evaluation dated

May 31, 1996. No significant hazards

consideration comments received: No

Local Public Document Room

location: York County Library, 138 East

Black Street, Rock Hill, South Carolina

29730

Duke Power Company, et al., Docket

Nos. 50–413 and 50–414, Catawba

Nuclear Station, Units 1 and 2, York

County, South Carolina

Date of application for amendments:

January 12, 1995, as supplemented by

letter dated June 29, 1995

Brief description of amendments: The

amendments revise and clarify portions

of Technical Specification Section 6.0,

‘‘Administrative Controls.’’

Date of issuance: May 30, 1996

Effective date: As of the date of

issuance to be implemented within 30

days

Amendment Nos.: 145 and 139

Facility Operating License Nos. NPF–

35 and NPF–52: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: November 24, 1995 (60 FR

58109) The Commission’s related

evaluation of the amendments is

contained in a Safety Evaluation dated

May 30, 1996. No significant hazards

consideration comments received: No

Local Public Document Room

location: York County Library, 138 East

Black Street, Rock Hill, South Carolina

29730

Duke Power Company, et al., Docket

Nos. 50–413 and 50–414, Catawba

Nuclear Station, Units 1 and 2, York

County, South Carolina

Date of application for amendments:

April 3, 1996

Brief description of amendments: The

amendments revise the Technical

Specifications and the associated Bases

to provide that if neither Train A or

Train B of the hydrogen igniter is

operable in any one containment region,

there is an allowance of 7 days to restore

one hydrogen igniter to operable status,

or be in hot shutdown within the next

6 hours.

Date of issuance: June 3, 1996

Effective date: As of the date of

issuance to be implemented within 30

days

Amendment Nos.: 147 and 141

Facility Operating License Nos. NPF–

35 and NPF–52: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: April 16, 1996 (61 FR 16649)

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated June 3, 1996 No

significant hazards consideration

comments received: No

Local Public Document Room

location: York County Library, 138 East

Black Street, Rock Hill, South Carolina

29730
31188 Federal Register / Vol. 61, No. 119 / Wednesday, June 19, 1996 / Notices

Entergy Operations, Inc., Docket No.

50–382, Waterford Steam Electric

Station, Unit 3, St. Charles Parish,

Louisiana

Date of amendment request: May 19,

1995, as supplemented by letter dated

December 7, 1995

Brief description of amendment: The

amendment revised the recombiner

surveillance requirements to conform

with the staff guidance provided in

NUREG–1432, ‘‘Standard Technical

Specifications Combustion Engineering

Plants.’’

Date of issuance: June 5, 1996

Effective date: June 5, 1996

Amendment No.: 119

Facility Operating License No. NPF–

38. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: January 3, 1996 (61 FR 180)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated June 5, 1996. 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 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 application for amendments:

January 4, 1996

Brief description of amendments:

These amendments rectify a

discrepancy in Technical Specification

3.5.3, and provide assurance that

administrative controls for High

Pressure Safety Injection pumps remain

effective in the lower operational

modes.

Date of Issuance: May 30, 1996

Effective Date: May 30, 1996

Amendment Nos.: 143 and 183

Facility Operating License Nos. DPR–

67 and NPF–16: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: February 14, 1996 (61 FR

5813) The Commission’s related

evaluation of the amendments is

contained in a Safety Evaluation dated

May 30, 1996. No significant hazards

consideration comments received: No.

Local Public Document Room

location: Indian River Junior College

Library, 3209 Virginia Avenue, Fort

Pierce, Florida 34954–9003

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 application for amendments:

November 22, 1995

Brief description of amendments:

These amendments upgrade existing TS

3/4.4.6.1 for the Reactor Coolant System

Leakage Detection Systems by adopting

the Standard Technical Specifications

for Combustion Engineering Plants to

both St. Lucie Units.

Date of Issuance: May 30, 1996

Effective Date: May 30, 1996

Amendment Nos.: 144 and 84

Facility Operating License Nos. DPR–

67 and NPF–16: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: January 22, 1996 (61 FR 1629)

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated May 30, 1996. No

significant hazards consideration

comments received: No.

Local Public Document Room

location: Indian River Junior College

Library, 3209 Virginia Avenue, Fort

Pierce, Florida 34954–9003

GPU Nuclear Corporation, et al.,

Docket No. 50–219, Oyster Creek

Nuclear Generating Station, Ocean

County, New Jersey

Date of application for amendment:

March 28, 1996 (TSCR 234)

Brief description of amendment: The

amendment modifies Technical

Specification pages 3.1–5 and 3.1–16 to

indicate 40 percent of the rated reactor

thermal power as the anticipatory

reactor scram bypass setpoint on turbine

trip or generator load rejection.

Date of Issuance: June 4, 1996

Effective date: As of the date of

issuance, to be implemented within 30

days.

Amendment No.: 184

Facility Operating License No. DPR–

16. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: April 24, 1996 (61 FR 18167)

The Commission’s related evaluation of

this amendment is contained in a Safety

Evaluation dated June 4, 1996 No

significant hazards consideration

comments received: No.

Local Public Document Room

location: Ocean County Library,

Reference Department, 101 Washington

Street, Toms River, NJ 08753

Houston Lighting & Power Company,

City Public Service Board of San

Antonio Central Power and Light

Company, City of Austin, Texas, Docket

No. 50–498, South Texas Project, Unit 1,

Matagorda County, Texas

Date of amendment request: January

22, 1996, as supplemented April 4 and

May 2, 1996

Brief description of amendment: The

amendment modified the steam

generator tube plugging criteria in TS 3/

4.4.5, Steam Generators, the allowable

primary-to-secondary leakage in TS 3/

4.4.6.2, Operational Leakage, and the

associated Bases. These changes

allowed the implementation of alternate

steam generator tube plugging criteria

for the tube support plate/tube

intersections for Unit 1.

Date of issuance: May 22, 1996

Effective date: May 22, 1996

Amendment No.: 83

Facility Operating License No. NPF–

76. The amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: April 16, 1996 (61 FR 16651)

as corrected April 22, 1996 (61 FR

17735). The additional information

contained in the supplemental letter

dated May 2, 1996, was clarifying in

nature and thus, within the scope of the

initial notice and did not affect the

staff’s proposed no significant hazards

consideration determination. The

Commission’s related evaluation of the

amendment is contained in a Safety

Evaluation dated May 22, 1996. No

significant hazards consideration

comments received: No

Local Public Document Room

location: Wharton County Junior

College, J. M. Hodges Learning Center,

911 Boling Highway, Wharton, TX

77488

IES Utilities Inc., Docket No. 50–331,

Duane Arnold Energy, Center, Linn

County, Iowa

Date of application for amendment:

July 21, 1995, as supplemented August

8, 1995 and December 15, 1995

Brief description of amendment: The

amendment made administrative

changes to various sections of the DAEC

Technical Specifications (TS). The

amendment replaced the surveillance

condition when an Emergency Service

Water pump or loop is inoperable with

an OPERABILITY verification of the

opposite train’s Emergency Diesel

Generator (EDG). The amendment

modified the TS to allow credit for

demonstration of EDG OPERABILITY

that occurred within the previous 24

hours. The amendment revised the

format and language of TS Section 5.5
Federal Register / Vol. 61, No. 119 / Wednesday, June 19, 1996 / Notices 31189

to clarify the requirements and state the

capacity of the spent fuel pool and vault

storage in order to remove ambiguities

in the wording and to be more

consistent with the Improved Standard

TS guidance. The amendment revised

the list of Operations Committee

responsibilities (Section 6.5.1.6) to

eliminate Committee review of

procedures implementing Security and

Emergency Plans.

Date of issuance: June 5, 1996

Effective date: June 5, 1996

Amendment No.: 214

Facility Operating License No. DPR–

49. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: September 27, 1995 (60 FR

49938) and February 2, 1996 (61 FR

3953) The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

June 5, 1996. No significant hazards

consideration comments received: No.

Local Public Document Room

location: Cedar Rapids Public Library,

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

Iowa 52401

Northern States Power Company,

Docket Nos. 50–282 and 50–306, Prairie

Island Nuclear Generating Plant, Unit

Nos. 1 and 2, Goodhue County,

Minnesota

Date of application for amendments:

May 4, 1995, as supplemented

November 27, 1995, and March 1, 1996

Brief description of amendments: The

amendments revise the pressurizer and

main steam safety valve lift setting

tolerance from plus or minus 1 percent

to plus or minus 3 percent (as-found

setpoint only), revise the safety limit

curves, reformat Section 2, and correct

typographical errors.

Date of issuance: May 21, 1996

Effective date: May 21, 1996, with full

implementation within 30 days

Amendment Nos.: Unit 1 - 123, Unit

2 - 116

Facility Operating License Nos. DPR–

42 and DPR–60. Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: September 13, 1995 (60 FR

47621) The November 27, 1995, and

March 1, 1996, letters provided

clarifying information in response to

NRC staff questions. This information

was within the scope of the original

application and did not change the

staff’s initial proposed no significant

hazards consideration determination.

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated May 21, 1996. No

significant hazards consideration

comments received: No.

Local Public Document Room

location: Minneapolis Public Library,

Technology and Science Department,

300 Nicollet Mall, Minneapolis,

Minnesota 55401

Pacific Gas and Electric Company,

Docket Nos. 50–275 and 50–323, Diablo

Canyon Nuclear Power Plant, Unit Nos.

1 and 2, San Luis Obispo County,

California

Date of application for amendments:

March 13, 1996

Brief description of amendments:

These amendments delete the

requirement in Technical Specifications

(TS) 4.0.5a for NRC written approval

prior to implementation of relief from

ASME Code requirements by deleting

‘‘...(g),.except where specific written

relief has been granted by the

Commission pursuant to 10 CFR

50.55a(g)(6)(i).’’ Also, the amendments

add the ASME Section XI definition of

‘‘Biennially or every 2 years - At least

once per 731 days,’’ in TS 4.0.5b.

Date of issuance: May 28, 1996

Effective date: May 28, 1996

Amendment Nos.: Unit 1 - 112; Unit

2 - 110

Facility Operating License Nos. DPR–

80 and DPR–82: The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: April 24, 1996 (61 FR 18173)

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated May 28, 1996. No

significant hazards consideration

comments received: No.

Local Public Document Room

location: California Polytechnic State

University, Robert E. Kennedy Library,

Government Documents and Maps

Department, San Luis Obispo, California

93407

Pacific Gas and Electric Company,

Docket Nos. 50–275 and 50–323, Diablo

Canyon Nuclear Power Plant, Unit Nos.

1 and 2, San Luis Obispo County,

California

Date of application for amendments:

April 3, 1996

Brief description of amendments:

These amendments revise the combined

Technical Specifications (TS) for the

Diablo Canyon Nuclear Power Plant,

Unit Nos. 1 and 2 to revise Technical

Specifications 3/4.7.5, ‘‘Control Room

Ventilation System;’’ 3/4.7.6, ‘‘Auxiliary

Building Safeguards Air Filtration

System;’’ and 3/4.9.12, ‘‘Fuel Handling

Building Ventilation System’’ to clarify

the testing methodology utilized by

PG&E to determine the operability of the

charcoal and high efficiency particulate

air (HEPA) filters in the engineering

safeguards features (ESF) air handling

units at the Diablo Canyon Power Plant

(DCPP).

Date of issuance: May 28, 1996

Effective date: May 28, 1996

Amendment Nos.: Unit 1 - 113; Unit

2 - 111

Facility Operating License Nos. DPR–

80 and DPR–82: The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: April 24, 1996 (61 FR 18173)

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated May 28, 1996. No

significant hazards consideration

comments received: No.

Local Public Document Room

location: California Polytechnic State

University, Robert E. Kennedy Library,

Government Documents and Maps

Department, San Luis Obispo, California

93407

Rochester Gas and Electric

Corporation, Docket No. 50–244, R. E.

Ginna Nuclear Power Plant, Wayne

County, New York

Date of application for amendment:

May 8, 1996, as supplemented May 10,

1996, and May 29, 1996, and June 3,

1996.

Brief description of amendment: This

amendment modifies the Technical

Specifications to correct several

typographical errors that were

implemented in the Improved Technical

Specifications at Ginna Station per

Amendment No. 61.

Date of issuance: June 3, 1996

Effective date: As of date of issuance.

Amendment No.: 65

Facility Operating License No. DPR–

18: Amendment revised the Technical

Specifications. Public comments

requested as to proposed no significant

hazards consideration: Yes (61 FR

24965, dated May 17, 1996). That notice

provided an opportunity to submit

comments on the Commission’s

proposed no significant hazards

consideration determination. No

comments have been received. The

notice published May 17, 1996, also

provided for a hearing by June 17, 1996,

but indicated that if a Commission

makes a final no significant hazards

consideration determination, any such

hearing would take place after issuance

of the amendment. The Commission’s

related evaluation of the amendment is

contained in a Safety Evaluation dated

June 3, 1996.

Local Public Document Room

location: Rochester Public Library, 115

South Avenue, Rochester, New York

14610.
31190 Federal Register / Vol. 61, No. 119 / Wednesday, June 19, 1996 / Notices

Union Electric Company, Docket No.

50–483, Callaway Plant, Unit 1,

Callaway County, Missouri

Date of application for amendment:

February 9, 1996 as superseded by letter

dated March 22, 1996.

Brief description of amendment: The

amendment revises Technical

Specification (TS) 1.7, 4.6.1.1, 3.6.1.3,

4.6.1.3, 6.8.4 and the associated Bases

section to directly reference Regulatory

Guide 1.163, ‘‘Performance-Based

Containment Leak Test Program,’’ as

required by 10 CFR 50, Appendix J,

Option B for the Type A containment

integrated leak rate tests and the Type

B and C local leak tests.

Date of issuance: May 28, 1996

Effective date: May 28, 1996, to be

implemented within 30 days from the

date of issuance.

Amendment No.: 111

Facility Operating License No. NPF–

30: The amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: April 24, 1996 (61 FR 18174)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated May 28, 1996. No

significant hazards consideration

comments received: No.

Local Public Document Room

location: Callaway County Public

Library, 710 Court Street, Fulton,

Missouri 65251.

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:

January 30, 1996

Brief description of amendments: The

amendments modify the Technical

Specifications to increase the minimal

allowable reactor coolant system total

flow rate.

Date of issuance: June 5, 1996

Effective date: June 5, 1996

Amendment Nos.: 201 and 182

Facility Operating License Nos. NPF–

4 and NPF–7. Amendments revised the

Technical Specifications.

Date of initial notice in Federal

Register: February 28, 1996 (61 FR

7559) The Commission’s related

evaluation of the amendments is

contained in a Safety Evaluation dated

June 5, 1996. 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.

Washington Public Power Supply

System, Docket No. 50–397, Nuclear

Project No. 2, Benton County,

Washington

Date of application for amendment:

April 24, as supplemented by letter

dated May 29, 1996.

Brief description of amendment: The

amendment would modify the WNP–2

technical specifications to support Cycle

12 operation, reflect use of new fuel

obtained from ABB/Combustion

Engineering, and incorporate staffapproved core reload analysis computer

programs (codes). Date of issuance: June

4, 1996 Effective date: June 4, 1996, to

be implemented within 30 days of

issuance.

Amendment No.: 146

Facility Operating License No. NPF–

21: The amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: May 1, 1996 (61 FR 19326).

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated June 4, 1996. No

significant hazards consideration

comments received: No.

Local Public Document Room

location: Richland Public Library, 955

Northgate Street, Richland, Washington

99352

Notice Of Issuance Of Amendments To

Facility Operating Licenses And Final

Determination Of No Significant

Hazards Consideration And

Opportunity For A Hearing (Exigent

Public Announcement Or Emergency

Circumstances)

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 for the

amendment 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.

Because of exigent or emergency

circumstances associated with the date

the amendment was needed, there was

not time for the Commission to publish,

for public comment before issuance, its

usual 30-day Notice of Consideration of

Issuance of Amendment, Proposed No

Significant Hazards Consideration

Determination, and Opportunity for a

Hearing.

For exigent circumstances, the

Commission has either issued a Federal

Register notice providing opportunity

for public comment or has used local

media to provide notice to the public in

the area surrounding a licensee’s facility

of the licensee’s application and of the

Commission’s proposed determination

of no significant hazards consideration.

The Commission has provided a

reasonable opportunity for the public to

comment, using its best efforts to make

available to the public means of

communication for the public to

respond quickly, and in the case of

telephone comments, the comments

have been recorded or transcribed as

appropriate and the licensee has been

informed of the public comments.

In circumstances where failure to act

in a timely way would have resulted, for

example, in derating or shutdown of a

nuclear power plant or in prevention of

either resumption of operation or of

increase in power output up to the

plant’s licensed power level, the

Commission may not have had an

opportunity to provide for public

comment on its no significant hazards

consideration determination. In such

case, the license amendment has been

issued without opportunity for

comment. If there has been some time

for public comment but less than 30

days, the Commission may provide an

opportunity for public comment. If

comments have been requested, it is so

stated. In either event, the State has

been consulted by telephone whenever

possible.

Under its regulations, the Commission

may issue and make an amendment

immediately effective, notwithstanding

the pendency before it of a request for

a hearing from any person, in advance

of the holding and completion of any

required hearing, where it has

determined that no significant hazards

consideration is involved.

The Commission has applied the

standards of 10 CFR 50.92 and has made

a final determination that the

amendment involves no significant

hazards consideration. The basis for this

determination is contained in the

documents related to this action.

Accordingly, the amendments have

been issued and made effective as

indicated.

Unless otherwise indicated, the

Commission has determined that these

amendments satisfy the criteria for

categorical exclusion in accordance

with 10 CFR 51.22. Therefore, pursuant

to 10 CFR 51.22(b), no environmental

impact statement or environmental

assessment need be prepared for these

amendments. If the Commission has

prepared an environmental assessment

under the special circumstances

provision in 10 CFR 51.12(b) and has
Federal Register / Vol. 61, No. 119 / Wednesday, June 19, 1996 / Notices 31191

made a determination based on that

assessment, it is so indicated.

For further details with respect to the

action see (1) the application for

amendment, (2) the amendment to

Facility Operating License, 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 room for the

particular facility involved.

The Commission is also offering an

opportunity for a hearing with respect to

the issuance of the amendment. By July

19, 1996, the licensee may file a request

for a hearing with respect to issuance of

the amendment to the subject facility

operating license and any person whose

interest may be affected by this

proceeding and who wishes to

participate as a party in the proceeding

must file a written request for a hearing

and a petition for leave to intervene.

Requests for a hearing and a petition for

leave to intervene shall be filed in

accordance with the Commission’s

‘‘Rules of Practice for Domestic

Licensing Proceedings’’ in 10 CFR Part

2. Interested persons should consult a

current copy of 10 CFR 2.714 which is

available at the Commission’s Public

Document Room, the Gelman Building,

2120 L Street, NW., Washington, DC and

at the local public document room for

the particular facility involved. If a

request for a hearing or petition for

leave to intervene is filed by the above

date, the Commission or an Atomic

Safety and Licensing Board, designated

by the Commission or by the Chairman

of the Atomic Safety and Licensing

Board Panel, will rule on the request

and/or petition; and the Secretary or the

designated Atomic Safety and Licensing

Board will issue a notice of a hearing or

an appropriate order.

As required by 10 CFR 2.714, a

petition for leave to intervene shall set

forth with particularity the interest of

the petitioner in the proceeding, and

how that interest may be affected by the

results of the proceeding. The petition

should specifically explain the reasons

why intervention should be permitted

with particular reference to the

following factors: (1) the nature of the

petitioner’s right under the Act to be

made a party to the proceeding; (2) the

nature and extent of the petitioner’s

property, financial, or other interest in

the proceeding; and (3) the possible

effect of any order which may be

entered in the proceeding on the

petitioner’s interest. The petition should

also identify the specific aspect(s) of the

subject matter of the proceeding as to

which petitioner wishes to intervene.

Any person who has filed a petition for

leave to intervene or who has been

admitted as a party may amend the

petition without requesting leave of the

Board up to 15 days prior to the first

prehearing conference scheduled in the

proceeding, but such an amended

petition must satisfy the specificity

requirements described above.

Not later than 15 days prior to the first

prehearing conference scheduled in the

proceeding, a petitioner shall file a

supplement to the petition to intervene

which must include a list of the

contentions which are sought to be

litigated in the matter. Each contention

must consist of a specific statement of

the issue of law or fact to be raised or

controverted. In addition, the petitioner

shall provide a brief explanation of the

bases of the contention and a concise

statement of the alleged facts or expert

opinion which support the contention

and on which the petitioner intends to

rely in proving the contention at the

hearing. The petitioner must also

provide references to those specific

sources and documents of which the

petitioner is aware and on which the

petitioner intends to rely to establish

those facts or expert opinion. Petitioner

must provide sufficient information to

show that a genuine dispute exists with

the applicant on a material issue of law

or fact. Contentions shall be limited to

matters within the scope of the

amendment under consideration. The

contention must be one which, if

proven, would entitle the petitioner to

relief. A petitioner who fails to file such

a supplement which satisfies these

requirements with respect to at least one

contention will not be permitted to

participate as a party.

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. Since the Commission has

made a final determination that the

amendment involves no significant

hazards consideration, if a hearing is

requested, it will not stay the

effectiveness of the amendment. Any

hearing held would take place while the

amendment is in effect.

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 20555–001, Attention:

Docketing and Services Branch, or may

be delivered to the Commission’s Public

Document Room, the Gelman Building,

2120 L Street, NW., Washington, DC, by

the above date. Where petitions are filed

during the last 10 days of the notice

period, it is requested that the petitioner

promptly so inform the Commission by

a toll-free telephone call to Western

Union at 1-(800) 248–5100 (in Missouri

1-(800) 342–6700). The Western Union

operator should be given Datagram

Identification Number N1023 and the

following message addressed to (Project

Director): petitioner’s name and

telephone number, date petition was

mailed, plant name, and publication

date and page number of this Federal

Register notice. A copy of the petition

should also be sent to the Office of the

General Counsel, U.S. Nuclear

Regulatory Commission, Washington,

DC 20555–001, 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 the

factors specified in 10 CFR

2.714(a)(1)(i)-(v) and 2.714(d).

Commonwealth Edison Company,

Docket No. 50–249, Dresden Nuclear

Power Station, Unit No. 3

Date of application for amendment:

May 22, 1996

Brief description of amendment: The

amendment authorizes, on a one- time

temporary basis, operation of Dresden,

Unit 3, with the structural steel

members in the Low Pressure Coolant

Injection (LPCI) corner rooms outside

the Updated Final Safety Analysis

Report (UFSAR) design parameters, but

capable of performing their intended

safety function. Following a reactor

scram on May 15, 1996, Commonwealth

Edison Company (ComEd) performed a

Safety Evaluation (SE) in accordance

with the requirements of 10 CFR 50.59

to determine if the current configuration

of the corner room structural steel

members had reduced the margin of

safety as described in the UFSAR. The

SE determined that the configuration

does not reduce the margin of safety

with respect to the stress allowables for

the structural steel if subjected to a Safe

Shutdown Earthquake (SSE). An

unreviewed safety question was

determined to exist because stress

allowables for the structural steel

subjected to an Operating Basis

Earthquake (OBE) were found outside

the UFSAR requirements; however, the

current configuration of the corner room

structural steel members has not
31192 Federal Register / Vol. 61, No. 119 / Wednesday, June 19, 1996 / Notices

significantly reduced the margin of

safety as described in the UFSAR.

Date of Issuance: May 31, 1996

Effective date: May 31, 1996

Amendment No.: 144

Facility Operating License No. DPR–

25. The amendment revised the license.

Press release issued requesting

comments as to proposed no significant

hazards consideration: Yes. Joliet

Herald News on May 25, 1996, and the

Morris Daily Herald on May 29, 1996.

Comments received: No comments were

received on the proposed no significant

hazards consideration determination;

however, comments were received

concerning the licensee’s timeliness and

decision-making in restoring the UFSAR

design margin to the structural steel

members installed the LPCI corner

rooms at Dresden Unit 3.

The Commission’s related evaluation

of the amendment, finding of exigent

circumstances, consultation with the

State of Illinois and final determination

of no significant hazards consideration

are contained in a Safety Evaluation

dated May 31, 1996.

Attorney for licensee: Michael I.

Miller, Esquire; Sidley and Austin, One

First National Plaza, Chicago, Illinois

60690

Local Public Document Room

location: Morris Area Public Library

District, 604 Liberty Street, Morris,

Illinois 60450.

NRC Project Director: Robert A. Capra

Dated at Rockville, Maryland, this 12th day

of June 1996.

For the Nuclear Regulatory Commission

John A. Zwolinski,

Deputy Director, Division of Reactor Projects

- I/II, Office of Nuclear Reactor Regulation

[Doc. 96–15398 Filed 6–18–96; 8:45 am]

BILLING CODE 7590–01–F

OCCUPATIONAL SAFETY AND

HEALTH REVIEW COMMISSION

Sunshine Act Meeting

TIME AND DATE: 10:00 a.m., on June 25,

1996.

PLACE: The Commission’s National

Office at One Lafayette Centre, 1120

20th St., N.W., 9th Floor, Washington,

DC 20036–3419.

STATUS: Under 29 C.F.R. § 2203.4(d) this

meeting is subject to being closed by a

vote of the Commissioners taken at the

beginning of the meeting. Since the only

matters to be discussed at this meeting

will be specific cases in the

Commission’s adjudicative process, it is

likely that, pursuant to 29 C.F.R.

§ 2203.3(b)(10), the meeting will be

closed upon a proper vote taken.

MATTERS TO BE CONSIDERED: Cases in the

Commission’s adjudicative process.

CONTACT PERSON FOR MORE INFORMATION:

Earl R. Ohman, Jr., General Counsel,

(202) 606–5410.

Earl R. Ohman, Jr.,

General Counsel.

[FR Doc. 96–15749 Filed 6–17–96; 8:45 am]

BILLING CODE 7600–01–M

RAILROAD RETIREMENT BOARD

Sunshine Act Meeting

Notice is hereby given that the

Railroad Retirement Board will hold a

meeting on June 26, 1996, 9:00 a.m., at

the Board’s meeting room on the 8th

floor of its headquarters building, 844

North Rush Street, Chicago, Illinois,

60611.

The agenda for this meeting follows:

Portion Open to the Public

(1) Annual Actuarial Report (Sec. 22 of the

Railroad Retirement Act of 1974 and Sec.

502 of the Railroad Retirement Solvency

Act of 1983)

(2) Fiscal Year 1996 Budget Allocations

(3) Proposed Reorganization—Bureau of

Information Systems

(4) Letters to Congress on H.R. 2942 and S.

1552

(5) Draft Legislation Proposed on April 4,

1996—Draft Legislation to Enhance Debt

Collection Efforts

(6) Medicare Part B Services (Contract No.

92RRB006)

(7) Regulations, Claims Manuals, Rulings,

and Procedures

(8) Status of Intermodal Services Under the

Railroad Retirement and Railroad

Unemployment Insurance Acts

(9) Regulations—Part 230 (Reduction and

Non-Payment of Annuities by Reason of

Work)

(10) Employee Service Determinations:

A. Maryland Midland Railway, Inc.—James

W. Schaeffer, Jr.

B. Joyce Goss

(11) Labor Member Truth in Budgeting Status

Report

Portion Closed to the Public

(A) Request for Change in Position Index

(Bureau of Hearings and Appeals)

(B) Pending Board Appeals:

(1) Anderson, Raymond

(2) Garcia, Fedelina

(3) Herbert, Harold

(4) Howard, Alvira M.

(5) McLeod, Jasper N.

(6) Trybala, Therese A.

The person to contact for more

information is Beatrice Ezerski,

Secretary to the Board, Phone No. 312–

751–4920.

Dated: June 14, 1996.

Beatrice Ezerski,

Secretary to the Board.

[FR Doc. 96–15720 Filed 6–17–96; 11:09 am]

BILLING CODE 7905–01–M

SECURITIES AND EXCHANGE

COMMISSION

Proposed Collection; Comment

Request

Upon Written Request, Copies Available

From: Securities and Exchange

Commission, Office of Filings and

Information Services, Washington, DC

20549

Extension:

Rule 15c2–5

SEC File No. 270–195

OMB Control No. 3235–0198

Notice is hereby given that pursuant

to the Paperwork Reduction Act of 1995

(44 U.S.C. 3501 et seq.), the Securities

and Exchange Commission

(‘‘Commission’’) is publishing the

following summary of collection for

public comment.

Rule 15c2–5 prohibits a broker-dealer

from arranging a loan for a customer to

whom a security is sold unless, before

the transaction is entered into, the

broker-dealer first: (1) delivers to the

customer a written statement setting

forth certain information about the

specific arrangement being offered to

him; (2) obtains from the customer

sufficient information concerning his or

her financial situation and needs so as

to determine that the entire transaction

is suitable for the customer; and (3)

retains in his or her files a written

statement setting forth the basis upon

which the broker-dealer made such

determination. The information

required by the rule is necessary for the

execution of the Commission’s mandate

under the Securities Exchange Act of

1934 (‘‘Exchange Act’’) to prevent

fraudulent, manipulative, and deceptive

acts and practices by broker-dealers.

There are approximately 50 respondents

that require an aggregate total of 600

hours to comply with the rule. Each of

these approximately 50 registered

broker-dealers makes an estimated 6

annual responses, for an aggregate total

of 300 responses per year. Each

response takes approximately 2 hours to

complete. Thus, the total compliance

burden per year is 600 burden hours.

The approximate cost per hour is $20,

resulting in a total cost of compliance

for the respondents of $12,000 (600

hours @ $20).

Written comments are invited on: (a)

whether the proposed collection of