8790 Federal Register / Vol. 62, No. 38 / Wednesday, February 26, 1997 / Notices

Proposed Language: Safety-Conscious

Work Environment

(a) Licensees shall establish and

maintain a safety-conscious work

environment in which employees are

encouraged to raise safety and

regulatory concerns, and where such

concerns are promptly reviewed, given

priority based on their potential safety

significance, and appropriately resolved

with timely feedback to the originator of

the concern. Attributes of a safetyconscious work environment include:

(1) A management attitude that

promotes employee involvement and

confidence in raising and resolving

concerns;

(2) A clearly communicated

management policy that safety has the

utmost priority, overriding, if necessary,

the demands of production and project

schedules;

(3) A strong, independent quality

assurance organization and program;

(4) A training program that

encourages a positive attitude toward

safety;

(5) A safety ethic at all levels that is

characterized by an inherently

questioning attitude, attention to detail,

prevention of complacency, a

commitment to excellence, and personal

accountability in safety matters.

(b) When circumstances occur that

could adversely impact the safetyconscious environment, or when

conditions arise that indicate the

potential emergence of an adverse trend

in the safety-conscious work

environment, the licensee shall take

action as required to ensure that the

safety-conscious environment is

preserved. Indicators that may be

considered as possible evidence of an

emerging adverse trend include, but are

not limited to:

(1) Adverse findings by the

Department of Labor or the NRC Office

of Investigation (OI) concluding that

discrimination has occurred against

employees for engaging in protected

activity, including a finding of the

existence of a hostile work environment;

(2) A significant increase in the rate

(or a sustained high number) of

allegations made to the NRC that

licensee employees are being subjected

to harassment and intimidation for

engaging in protected activity;

(3) A significant increase in the rate

(or a sustained high number) of

allegations made to the NRC concerning

matters of safety or regulatory concern,

particularly if accompanied by low

usage or a decrease in use of the

licensee’s employee concern program

(ECP) or other licensee channels for

reporting safety and regulatory

concerns;

(4) Other indications that the

licensee’s ECP or other programs for

identifying and resolving safety and

regulatory concerns are ineffective.

Such indications might include: delays

in or absence of feedback for concerns

raised to the ECP; breaches of

confidentiality for concerns raised to the

ECP; the lack of effective evaluation,

follow-up, or corrective action for

concerns raised to the ECP or findings

made by the licensee’s QA organization;

overall licensee ineffectiveness in

identifying safety issues; the occurrence

of repetitive or willful violations; a

licensee emphasis on cost-cutting

measures at the expense of safety

considerations; and/or poor

communication mechanisms within or

among licensee groups.

(c) The presence of one or more of the

indicators discussed in paragraph (b) of

this section may or may not, in

isolation, be considered evidence of

deterioration in the licensee’s safetyconscious work environment.

Evaluation of the licensee’s safetyconscious work environment should

consider these indicators in the context

of the overall work environment,

including the presence or absence of

other indicators, and the presence or

absence of related licensee safety and

performance issues.

(d) If, based on a review of indicators

as discussed in paragraphs (b) and (c) of

this section, the Executive Director for

Operations determines that the licensee

has failed to establish and maintain a

safety-conscious work environment as

discussed in paragraph (a) of this

section, the NRC at its discretion may

require the licensee to take action. This

action may include (but is not limited

to) ordering one or more of the

following:

(1) Establishment of a formal

employee concerns program (if one does

not already exist);

(2) Performance of an independent

survey of the licensee’s environment for

raising safety and regulatory concerns,

with periodic follow-up surveys to

monitor change;

(3) Establishment of an independent

group for oversight of licensee

performance in establishing and

maintaining a safety-conscious work

environment;

(4) Establishment of a ‘‘holding

period’’ policy, to be applied in cases

where an employee of the licensee or its

contractor registers a complaint of

having been discriminated against for

engaging in protected activity. The

holding period policy requires that,

when such an employee submits to the

licensee a complaint that he or she has

been discriminated against for engaging

in protected activity, the licensee will

maintain that employee’s pay and

benefits until the licensee has

investigated the complaint,

reconsidered the facts, negotiated with

the employee, and informed the

employee of a final decision on the

matter. After the licensee has informed

the employee of its final decision, the

holding period of continued pay and

benefits will continue for an additional

2 weeks to allow a reasonable time for

the employee to file a complaint of

discrimination with the DOL. If, by the

end of that 2-week period, the employee

has filed with the DOL a complaint of

discrimination for engaging in protected

activity, the licensee will maintain the

holding period of continued pay and

benefits until the DOL has made a

finding based on its initial investigation

of the employee’s complaint.

(5) Additional enforcement action

pursuant to Subpart B of Part 2,

including civil penalties.

Dated at Rockville, Maryland, this 19th day

of February, 1996.

For the Nuclear Regulatory Commission.

James Lieberman,

Director, Office of Enforcement.

[FR Doc. 97–4702 Filed 2–25–97; 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 February 1,

1997, through February 13, 1997. The

last biweekly notice was published on

February 12, 1997 (62 FR 6567).
Federal Register / Vol. 62, No. 38 / Wednesday, February 26, 1997 / Notices 8791

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 March 28, 1997, 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.

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
8792 Federal Register / Vol. 62, No. 38 / Wednesday, February 26, 1997 / Notices

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: January

30, 1997.

Description of amendment request:

The proposed amendment would

change the Updated Final Safety

Analysis Report (FSAR) to include the

credit for containment overpressure in

the Pilgrim Nuclear Power Station net

positive suction head (NPSH) analysis

for the emergency core cooling pumps.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

(1) Will crediting post-LOCA [loss-ofcoolant accident] wetwell airspace pressure

in ECCS [emergency core cooling system]

analyses involve a significant increase in the

probability or consequences of an accident

previously evaluated?

Chapter 14 of the FSAR contains

evaluations of the design basis accidents,

which include the refueling accident, the

main steam line break outside primary

containment, the recirculation line break

inside primary containment, and the control

rod drop accident. No increase in the

probability of the evaluated accidents will

result from crediting the post-LOCA wetwell

airspace pressure because post-LOCA

wetwell airspace pressure does not represent

an accident initiator but is rather a byproduct

of the conditions which will exist in the

containment after the pipe break inside

containment.

The worst radiological consequences for

the Pilgrim plant are associated with the

design basis LOCA which is the double

guillotine failure of the recirculation system

piping. The radiological analysis of this

event, contained in FSAR Chapter 14, uses a

TID–14844 source term and assumes a 1.5%

per day leakage from the containment, which

is greater than the maximum leakage allowed

by the Technical Specifications. The results

of this analysis are presented in Table 14.5–

2 of the FSAR and indicate substantial

margin when compared to 10 CFR Part 100

limits.

The radiological consequences of the

design basis accident are not increased by

taking credit for the post-LOCA wetwell

airspace pressure. Assuming containment

integrity exists, the mechanism for increasing

the consequences of the accident would be

an increased leakage rate caused by an

increase of the average differential pressure

between primary and secondary containment

during the accident response. However, the

NPSH analyses performed for Pilgrim, which

credits the post-LOCA wetwell airspace, does

not require that the differential pressure

between primary and secondary containment

be maintained above the minimum that exists

due to the equilibrium conditions based on

the suppression pool temperature.

Specifically, the wetwell airspace pressure

credited in the ECCS pump NPSH analyses

is provided by an increase in wetwell vapor

pressure and air/nitrogen partial pressure in

equilibrium with increasing pool temperature

with an accounting for containment initial

conditions and leakage.

By crediting the post-LOCA wetwell

airspace pressure in the calculation of NPSH,

no requirement is created to purposely

maintain a higher containment pressure than

would otherwise occur; no requirement is

incurred to delay operating containment heat

removal equipment at the highest rate

possible; no requirement is incurred to

deliberately continue any condition of high

containment pressure to maintain adequate

NPSH; and no requirement is incurred for the

purposeful addition of air/nitrogen into the

containment to increase the available

pressure.

Based on these reasons, the probability of

accidents previously evaluated is not

increased, and the consequences of the

design basis accident are not increased.

(2) Will crediting post-LOCA wetwell

airspace pressure create the possibility for

new or different kinds of accidents?

As stated above, Chapter 14 of the Pilgrim

FSAR contains evaluations of design basis

accidents that include the refueling accident,

the main steam line break outside primary

containment, the recirculation line break

inside primary containment, and the control

rod drop accident. New or different types of

accidents are not created by crediting the

post-LOCA wetwell airspace pressure

because post-LOCA wetwell airspace

pressure does not represent an accident

initiator but is rather a byproduct of the

conditions which will exist in the

containment after the pipe break inside

containment.

Therefore, crediting post-LOCA wetwell

airspace pressure does not create the

possibility for new or different kinds of

accidents from those previously analyzed.

(3) Will crediting post-LOCA wetwell

airspace pressure in ECCS NPSH analyses

involve a significant reduction in a margin of

safety?

The integrity of the primary containment

and the operation of the ECCS systems in

combination limit the off-site doses to values

less than those suggested in 10 CFR 100 in

the event of a break in the primary system

piping. In order for the ECCS pumps to meet

their performance requirements, the NPSH

available to the pumps throughout the

accident response must meet their specific

NPSH requirements. Excess NPSH margin

will not improve the performance of the

ECCS pumps because NPSH available must

only meet NPSH requirements for the pump

to operate on its pump curve and meet design

expectations.

Crediting post-LOCA wetwell airspace

pressure in ECCS NPSH analyses increases

the NPSH available to the pumps connected

to the suppression pool but limits the

increase in NPSH available consistent with

the bounding leakage assumptions for the

containment system. The amount of postaccident pressure that is utilized in ECCS

NPSH analyses is calculated in a manner

such that the pressure credited represents a

conservative lower bound of the pressure

available. Therefore, it is expected that the

NPSH margin will exceed that credited in the

NPSH analyses.

Credit for wetwell airspace pressure in

NPSH analyses is not required under all

circumstances. If the suction strainers for the

ECCS pumps remain relatively free of postLOCA debris, adequate NPSH will be

available without credit for the wetwell

airspace pressure provided by the post-LOCA

heatup of the air/nitrogen gas in the

containment. If debris accumulates on the

pump suction strainers, the NPSH available

to the ECCS pumps will be decreased due to

the head loss caused by the debris. Credit for

the post-LOCA wetwell airspace pressure in

the analyses indicates that there is adequate

NPSH margin such that NPSH available will

remain above NPSH required, and ECCS

pump performance will meet applicable

requirements. Based on the above discussion,

credit for wetwell airspace pressure in ECCS

NPSH analyses does not involve a significant

reduction in the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied.
Federal Register / Vol. 62, No. 38 / Wednesday, February 26, 1997 / Notices 8793

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: Patrick D.

Milano, Acting.

Carolina Power & Light Company, et

al., Docket Nos. 50–325 and 50–324,

Brunswick Steam Electric Plant, Units 1

and 2, Brunswick County, North

Carolina.

Date of amendments request:

November 1, 1996.

Description of amendments request:

The amendments would revise the

Technical Specifications (TS) to allow

full implementation of the Boiling

Water Reactor Owners Group (BWROG)

Enhanced Option 1–A Reactor Stability

Long Term Solution. In Safety

Evaluation Reports (SERs) transmitted

to Kevin P. Donovan, Chairman,

BWROG, by letters from Robert C. Jones,

Office of Nuclear Reactor Regulation,

NRC, dated June 21, 1996, and

September 20, 1996, the NRC staff

concluded that Enhanced Option 1–A

generic technical specifications

described in Topical Report NEDO–

32339, Supplement 4, were acceptable

for referencing in license applications.

The characteristics of a reactor system

most important in determining stability

performance are power, core flow and

power distribution. The proposed

changes would delete the current limits

on power and flow conditions in the

technical specifications associated with

the implementation of the guidance in

General Electric Service Information

Letter (SIL) #380, Revision 1 and the

power/flow figure (Figure 3.4.1.1–1),

add two new specifications on the

fraction of core boiling boundary (FCBB)

and the Period Based Detection System

(PBDS) and relocate certain

requirements pertaining to the Average

Power Range Monitors (APRM) to the

Core Operating Limits Report (COLR).

The current Technical Specifications

for Units 1 and 2 permit single loop

operation (SLO) only for a 12-hour

period and there are no provisions for

potential alterations of safety limits or

operating limits because of SLO

conditions. Approval of the amendment

applications discussed above would

permit SLO operation subject to the

compensatory actions and requirements

that address this mode of operation in

the revised Technical Specifications.

However, Brunswick Unit 2’s License

currently has a condition, 2.C.(5) that

states that the reactor shall not be made

critical unless both recirculation loops

are in service. This License Condition

also requires the plant to be placed in

the hot shutdown condition within 24

hours if one recirculation loop becomes

out-of-service. The License Condition

also allows one or both recirculation

loops to be out-of-service for the

purposes of testing (not to exceed 24

hours). Whereas the License Condition

would permit SLO for up to 24 hours,

the current TS limit SLO to 12 hours.

The License Condition was added to

permit natural circulation testing as

required by the startup test program but

to preclude long-term SLO or operation

in the natural circulation mode. The

startup test program was completed

many years ago for Brunswick Unit 2

and natural circulation operation is no

longer allowed. The License Condition

is no longer relevant and if not deleted

would negate the objectives of the

proposed license amendments

discussed above. The licensee has

submitted proposed license

amendments on the same date of the

subject application (i.e., November 1,

1996) to convert the Brunswick Units 1

and 2 Technical Specifications to the

Improved Standard Technical

Specifications (ISTS) consistent with

NUREG–1433, Revision 1, ‘‘Standard

Technical Specifications for General

Electric Plants, BWR 4.’’ Attachment 6

of the later application was a proposed

revision of the Brunswick Unit 2

License to delete License Condition

2.C.(5). While the Notice of

Consideration of Issuance of the ISTS

amendments (62 FR 3719) discussed

deletion of License Condition 2.C.(5),

the deletion is discussed in this Notice

as well, since if the subject amendment

applications are approved, the License

Condition would thwart the

considerable effort represented by the

subject amendments to finally resolve

the thermal-hydraulic stability issues for

Brunswick Units 1 and 2.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed amendments do not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed amendments allow the

implementation of the Enhanced Option 1–

A (E1A) long term solution to the neutronic/

thermal hydraulic instability issue. Current

Technical Specification restrictions on power

and flow conditions, number of operating

recirculation loops and operator actions

implemented to reduce the probability of

neutronic/thermal hydraulic instability are

eliminated and new stability control

requirements consistent with NEDO–32339,

Supplement 4, are imposed. These

requirements include restrictions on power

and flow conditions and actions associated

with the modified APRM flow biased scram

and control rod block functions. These

actions include adherence to the boiling

boundary limit stability control prior to entry

and during operation in the region of the

power and flow operating domain which is

potentially susceptible to neutronic/thermal

hydraulic instability in the absence of the

stability control. In addition, the proposed

amendments require operator actions based

upon a new Period Based Detection System

(PBDS). The PBDS is designed to provide

alarm indication that conditions consistent

with a significant degradation in the stability

performance of the reactor has occurred and

the potential for imminent onset of

neutronic/thermal hydraulic instability may

exist.

The proposed amendments will permit

operation in regions of the power and flow

operating domain postulated to be

susceptible to neutron/thermal hydraulic

instability (i.e., Restricted and Monitored

Regions). Operation in these regions does not

increase the probability of occurrence of

initiators and precursors of previously

analyzed accidents when neutronic/thermal

hydraulic instability is not possible. The

proposed amendments also permit the

implementation of the features of the E1A

solution which prevent neutronic/thermal

hydraulic instability including pre-emptive

reactor scram upon entry into the region of

the power and flow operating domain most

susceptible to neutronic/thermal hydraulic

instability (i.e., Exclusion Region).

Furthermore, the E1A solution requires

implementation of stability control prior to

entry into a region of the power and flow

operating domain which is potentially

susceptible, in the absence of stability

control, to neutronic/thermal hydraulic

instability (i.e., Restricted Region). The E1A

solution prevents neutronic/thermal

hydraulic instability during operation in

regions of the power and flow operating

domain previously excluded from operation

and therefore does not significantly increase

the probability of a previously analyzed

accident.

Operation in the regions of the power and

flow operating domain excluded by current

Technical Specification 3/4.4.1.1 and Figure

3.4.1.1–1 can occur as a result of anticipated

operational occurrences. The severity of

these transients may increase in the absence

of operator actions due to the potential

occurrence of neutronic/thermal hydraulic

instability as a result of operation in these

regions. The proposed amendments will

permit the implementation of the E1A long

term solution to the stability issue. Required

features of the E1A solution include

adherence to a boiling boundary limit

stability control prior to selection by the

operator of APRM flow biased scram and

control rod block function setpoints which
8794 Federal Register / Vol. 62, No. 38 / Wednesday, February 26, 1997 / Notices

allow operation in a region of the power and

flow operating domain potentially

susceptible, in the absence of the stability

control, to neutronic/thermal hydraulic

instability. Upon entry, as a result of an

anticipated operational occurrence, into the

region most susceptible to neutronic/thermal

hydraulic instability during operation with

the boiling boundary limit stability control

met, the pre-emptive reactor scram prevents

neutronic/thermal hydraulic instability.

Therefore, the consequences of an accident

do not significantly increase while operating

with the stability control met. After exiting

the region requiring the stability control to be

met, the setpoints are automatically returned

to the values applicable when anticipated

operational occurrences can be initiated from

conditions with the stability control not met.

This automatic actuation of the more

conservative setpoints ensures that the preemptive reactor scram will prevent operation

as a result of an anticipated operational

occurrence in the region most susceptible to

neutronic/thermal hydraulic instability

should the operator not select the more

conservative setpoints appropriate for

operation following exit from the region

requiring stability control. These required

features of the E1A solution prevent

operation in the region of the power and flow

operating domain most susceptible to

postulated neutronic/thermal hydraulic

instability by pre-emptive reactor scram

regardless of how the region was entered.

Therefore, the proposed amendments prevent

the occurrence of neutronic/thermal

hydraulic instability as a consequence of an

anticipated operational occurrence and do

not significantly increase the consequences

of any previously analyzed accident.

2. The proposed amendments do not create

the possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed amendments eliminate

restrictions on power and flow conditions

and impose alternative restrictions which

permit the implementation of the E1A long

term stability solution. The current

restrictions on the power and flow conditions

do not prevent the entry into regions of the

power and flow operating domain most

susceptible to neutronic/thermal hydraulic

instability and therefore the possibility of

neutronic/thermal hydraulic instability exists

in the absence of operator action. The

required features of the E1A solution

implement a pre-emptive scram upon entry

into the region most susceptible, without

operator action, to neutronic/thermal

hydraulic instability. The accessible

operating domain allowed by the proposed

amendments is a subset of the power and

flow operating domain currently allowed.

Current initiators and precursors of accidents

and anticipated operational occurrences can

not occur with new or different initial

conditions. Therefore, the proposed

amendments do not create the possibility of

a new or different kind of accident from that

previously evaluated.

Concurrent with the implementation of the

proposed amendments, a modified Flow

Control Trip Reference (FCTR) card and a

new Period Based Detection System (PBDS)

will be installed as required by the E1A

solution. The function of the FCTR card is to

aid the operator in the identification of entry

into regions of the power and flow operating

domain potentially susceptible to neutronic/

thermal hydraulic instability and to initiate

a pre-emptive scram upon entry into the

regions most susceptible to neutronic/

thermal hydraulic instability. This is

accomplished by altering the values of

setpoints of the APRM flow biased scram and

the control rod block functions generated by

the modified FCTR card, which are existing

functions of the current FCTR card. The

modified FCTR card design includes

components which may be susceptible to

electromagnetic interference or other

environmental effects. The plant specific

environmental conditions (temperature,

humidity, pressure, seismic, and

electromagnetic compatibility) have been

confirmed to be enveloped by the PBDS

environmental qualification values and will

be confirmed to be enveloped by the E1A

FCTR card environmental qualification

values prior to installation. Therefore, the

potential for spurious scrams or common

mode failures induced by environmental

effects (e.g., electromagnetic interference) is

considered negligible. The installation of the

modified FCTR card will therefore not create

the possibility of a new or different kind of

accident from any accident previously

evaluated. The function of the PBDS is to

provide the operator with an indication that

conditions consistent with a significant

degradation in the stability performance of

the reactor has occurred and the potential for

imminent onset of neutronic/thermal

hydraulic instability may exist. This is

accomplished by the installation of a new

PBDS card in the Neutron Monitoring

System. The PBDS card takes inputs from

individual local power range monitors and

provides displays indicating alarm and status

conditions to the operator in the control

room. These displays can not create the

possibility of a new or different kind of

accident from any accident previously

evaluated. The PBDS card design includes

components which may be susceptible to

electromagnetic interference or other

environmental effects. The plant specific

environmental conditions (temperature,

humidity, pressure, seismic, and

electromagnetic compatibility) have been

confirmed to be enveloped by the PBDS

environmental qualification values and will

be confirmed to be enveloped by the E1A

FCTR card environmental qualification

values prior to installation. Therefore, the

installation of the PBDS card will not create

the possibility of a new or different kind of

accident from any accident previously

evaluated.

3. The proposed amendments do not

involve a significant reduction in a margin of

safety. The proposed amendments permit the

implementation of the E1A long term

solution to the stability issue. Under certain

conditions, existing BWR designs are

susceptible to neutronic/thermal hydraulic

instability. General Design Criterion (GDC) 12

OF 10 CFR 50, Appendix A, requires thermal

hydraulic instability to be prevented by

design or be readily and reliably detected and

suppressed. When the design of the reactor

system does not prevent the occurrence of

neutronic/thermal hydraulic instability,

instability is an anticipated operational

occurrence. GDC 10 of 10 CFR 50, Appendix

A, requires that specified acceptable fuel

design limits not be exceeded during

anticipated operational occurrences.

Analyses performed by the BWROG

indicate that neutronic/thermal hydraulic

instability induced power oscillations could

result in conditions exceeding the Minimum

Critical Power Ratio (MCPR) Safety Limit

(SL) prior to detection and suppression by

the current design of the Neutron Monitoring

System and Reactor Protection System. To

ensure compliance with GDC 12, the BWROG

developed Interim Corrective Actions (ICAs)

to enhance the capability of the operator to

readily and reliably detect and suppress

neutronic/thermal hydraulic instability. The

BWROG ICAs also provided additional

guidance for monitoring local power range

monitors beyond the requirements of current

Technical Specification 3/4.4.1.1 to ensure

adequate margin to the onset of neutronic/

thermal hydraulic instability. Reliance on

operator actions to comply with GDC 12 was

accepted on an interim basis by the NRC

pending final implementation of a long term

solution to the stability issue.

The modified design of the Reactor

Protection System (APRM flow biased scram)

implemented with the E1A solution prevents

neutron/thermal hydraulic instability. The

E1A solution also requires implementation of

the stability control prior to entry into a

region of the power and flow operating

domain which is potentially susceptible, in

the absence of the stability control, to

neutronic/thermal hydraulic instability. As a

result, the margin to the onset of neutronic/

thermal hydraulic instability provided by the

existing Technical Specification

requirements and BWROG ICAs

recommendations is not significantly

reduced by the implementation of the E1A

solution. The E1A solution assures

compliance with GDC 12 by the prevention

of neutronic/thermal hydraulic instability

and therefore precludes neutronic/thermal

hydraulic instability from becoming a

credible consequence of an anticipated

operational occurrence. The consequences of

anticipated operational occurrences and the

margin to the MCPR SL will not change upon

the implementation of the E1A solution.

Therefore, the proposed amendments 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 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: University of North Carolina at

Wilmington, William Madison Randall

Library, 601 S. College Road,

Wilmington, North Carolina 28403–

3297.
Federal Register / Vol. 62, No. 38 / Wednesday, February 26, 1997 / Notices 8795

Attorney for licensee: William D.

Johnson, Vice President and Senior

Counsel, Carolina Power & Light

Company, Post Office Box 1551,

Raleigh, North Carolina 27602.

NRC Project Director: Mark Reinhart

(Acting).

Commonwealth Edison Company,

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

County Station, Units 1 and 2, LaSalle

County, Illinois.

Date of amendment request: January

20, 1997.

Description of amendment request:

The proposed amendments would

relocate the surveillance requirements

for selected instrumentation from the

Technical Specifications to licensee

controlled documents because the

instrumentation provides indication or

an alarm only. The affected surveillance

requirements are: 4.1.3.5.b, ‘‘Control

Rod Scram Accumulators’’; 4.5.1.d.2.c, ‘‘

Emergency Core Cooling Systems—

Operating’’; 4.5.3.1.b, ‘‘ECCS—

Suppression Chamber’’; and 4.6.2.1.c,

‘‘Containment Systems—Suppression

Chamber’’. In addition, the proposed

amendments would replace TS SR

4.4.3.2.1, ‘‘Reactor Coolant System

Leakage’’ and SR 4.5.1.d.1, ‘‘ECCS—

Operating’’ with surveillances more

appropriate to the associated LCOs and

action statements. Also, the proposed

amendments add an action statement to

TS 3.5.1, ‘‘ECCS—Operating’’ regarding

pressure of the ADS accumulator

backup compressed gas system bottle,

and delete action statements 3.5.3.c,

3.5.3.d, 3.6.2.1.c and 3.6.2.1.d regarding

suppression chamber water level

instrumentation.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

(1) Involve a significant increase in the

probability or consequences of an accident

previously evaluated because:

The proposed change relocates

instrumentation requirements, which provide

no post-accident function from the Technical

Specifications to the Bases, UFSAR,

procedures, or other plant controlled

documents. These requirements are part of

routine operational monitoring and are not

considered in the safety analysis. The Bases,

UFSAR, procedures, and other plant

controlled documents containing the

relocated information will be maintained in

accordance with 10 CFR 50.59. In addition to

10 CFR 50.59 provisions, the Technical

Specification Bases are subject to the change

control provisions in the Administrative

Controls Chapter of the Technical

Specifications. The UFSAR is subject to the

change control provisions of 10 CFR 50.71(e),

and plant procedures and other plant

controlled documents are subject to controls

imposed by plant administrative procedures,

which endorse applicable regulations and

standards. Since any changes to the Bases,

UFSAR, procedures, or other plant controlled

documents will be evaluated per the

requirements of 10 CFR 50.59, no significant

increase in the probability or consequences

of an accident previously evaluated will be

allowed. Therefore, this change does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The Reactor Coolant Operational Leakage

limits monitoring surveillance 4.4.3.2.1 has

been modified to eliminate procedural details

of what instrumentation/leakage detection

systems to use in verifying limits. The

proposed surveillance requires verification

that the reactor coolant system leakage is

within limits at the same frequency as the

current surveillance requirement. The reactor

coolant leakage detection systems operability

requirements are controlled by Technical

Specification 3/4.4.3.1. Since any changes to

procedures describing the method of

monitoring leakage will be evaluated per the

requirements of 10 CFR 50.59, no significant

increase in the probability or consequences

of an accident previously evaluated will be

allowed. Therefore, this change does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The monitoring action and the surveillance

requirements added for the Automatic

Depressurization System (ADS) pneumatic

supply help assure the continued operability

of ADS for the mitigation of accidents

involving high reactor vessel pressure and

the loss of the high pressure core spray

system. The surveillance frequency is

reasonable for the ADS supply header

pressure due to the redundancy of the

instrument nitrogen system, [and] several

alarms [that warn] of system trouble. The

ADS accumulator backup compressed gas

system bottle pressure monitoring

surveillance frequency and the proposed

action on low bottle pressure is reasonable

due to the [presence of the] ADS accumulator

check valves and the [availability of the]

normal ADS supply header. Therefore, this

change does 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 accident

previously evaluated because:

The proposed change does not involve a

physical alteration of the plant (no new or

different type of equipment will be installed)

or a change in the methods governing normal

plant operation. The proposed change will

not impose or eliminate any requirements,

and adequate control of the requirements will

be maintained. Thus, these changes do not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

(3) Involve a significant reduction in the

margin of safety because:

The proposed change will not reduce a

margin of safety because it has no impact on

any safety analysis assumption. In addition,

the requirements to be transposed from the

Technical Specifications to procedures, or

other plant controlled documents are the

same as the existing Technical

Specifications. Since any future changes to

these requirements in the Bases, UFSAR,

procedures, or other plant controlled

documents will be evaluated per the

requirements of 10 CFR 50.59, no significant

reduction in a margin of safety will be

allowed.

Based on 10 CFR 50.92, the existing

requirement for NRC review and approval of

revisions to these requirements proposed for

relocation does not have a specific margin of

safety upon which to evaluate. However,

since the proposed change is consistent with

the BWR Standard Technical Specifications,

NUREG–1434, approved by the NRC Staff,

revising the Technical Specifications to

reflect the approved level of instrumentation

requirements ensures no significant

reduction in the margin of safety.

The Reactor Coolant Operational Leakage

limits monitoring surveillance 4.4.3.2.1 has

been modified to eliminate procedural details

of what instrumentation/leakage detection

systems to use in verifying limits. The

proposed surveillance requires verification

that the reactor coolant system leakage is

within limits at the same frequency as the

current surveillance requirement. The reactor

coolant leakage detection systems operability

requirements are controlled by Technical

Specification 3/4.4.3.1. Because there are no

changes to either the reactor coolant leakage

detection systems and the reactor coolant

leakage continues to be maintained within

the specified limits, at the required

frequency, there is no reduction in the

margin of safety.

The monitoring action and the surveillance

requirements added for the Automatic

Depressurization System (ADS) pneumatic

supply help assure the continued operability

of ADS for the mitigation of accidents

involving high reactor vessel pressure and

the loss of the high pressure core spray

system. This helps assure ADS is maintained

in a ready status. The previous TS SRs only

tested the instrumentation, and did not verify

the parameter remained within limits.

Therefore, the margin of safety is not

reduced.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

requested amendments involve no

significant hazards consideration.

Local Public Document Room

location: Jacobs Memorial Library,

Illinois Valley Community College,

Oglesby, Illinois 61348.

Attorney for licensee: Michael I.

Miller, Esquire; Sidley and Austin, One

First National Plaza, Chicago, Illinois

60603.

NRC Project Director: Robert A. Capra.

Consumers Power Company, Docket

No. 50–255, Palisades Plant, Van Buren

County, Michigan.
8796 Federal Register / Vol. 62, No. 38 / Wednesday, February 26, 1997 / Notices

Date of amendment request: January

10, 1996.

Description of amendment request:

The proposed amendment would revise

test requirements for the containment

emergency escape airlock.

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 following evaluation supports the

finding that operation of the facility in

accordance with the proposed change to the

Technical Specifications would not:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed change does not alter any

plant operating conditions, operating

practices, equipment design, equipment

settings, or equipment capabilities.

Therefore, operation of the facility in

accordance with the proposed change will

not involve an increase in the probability of

an accident. This determination is made

because the full pressure test and the seal

contact check provides reasonable assurance

that the Emergency Escape Airlock doors will

act as designed to maintain containment

integrity. Procedures are established to test

seal integrity with full pressure airlock test

and to verify the seal contact following the

test. Acceptance criteria are established for

each evolution. Failure to meet the

acceptance criteria would result in corrective

action to restore the Emergency Escape

Airlock to the intended condition.

The proposed change defines the pressure

tests required for the Emergency Escape

Airlock and specifies the method used to

restore the airlock door seals after full

pressure testing. Due to the design of the

airlock, the doors must be opened after

testing. This change recognizes the practice

of verifying the final integrity of the airlock

by verifying door seal contact. Since the

pressure test does not load the door seals in

the same direction as a design basis accident,

this seal contact check provides better

assurance that the door is sealed than

alternative pressure tests. The Emergency

Escape Airlock continues to be capable of

performing its design function and the

consequences of those accidents previously

evaluated will not increase.

2. Create the possibility of a new or

different kind of accident from any

previously evaluated.

The proposed change does not alter any

plant operating conditions, operating

practices, equipment design, equipment

settings, or equipment capabilities.

Therefore, operation of the facility in

accordance with the proposed change will

not create the possibility of a new or different

kind of accident from any previously

evaluated.

3. Involve a significant reduction in a

margin of safety.

The proposed change requires testing of

the Emergency Escape Airlock at full

pressure (greater than or equal to Pa) rather

than a reduced pressure between-the-seals

test. This reduced pressure test is allowed by

the existing Technical Specifications when

the door is opened during periods when

containment integrity is required. The door

seal contact check and restoration will

provide assurance that the Emergency Escape

Airlock is capable of performing its design

function after the doors are opened during

recovery from full pressure testing.

Implementation of these test requirements

and meeting the acceptance criteria will

ensure that containment integrity with

respect to the Emergency Escape Airlock will

be maintained. Therefore, there will be no

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: Van Wylen Library, Hope

College, Holland, Michigan 49423.

Attorney for licensee: Judd L. Bacon,

Esquire, Consumers Power Company,

212 West Michigan Avenue, Jackson,

Michigan 49201.

NRC Project Director: John N.

Hannon.

Duke Power Company, Docket Nos.

50–269, 270 and 50–287, Oconee

Nuclear Station, Units 1, 2 and 3,

Oconee County, South Carolina.

Date of amendment request: February

5, 1997 (TSC 96–11)

Description of amendment request:

The proposed changes would reflect

replacement of the existing nuclear

instrumentation with an enhanced wide

range nuclear instrumentation system

that provides more channels and

continuous coverage from the source to

above the power range. As a result: (1)

The various references to Intermediate

Range of nuclear instrumentation would

be eliminated and replaced with

reference to Wide Range

instrumentation; (2) the minimum

number of operable Source and Wide

Range Nuclear Instrumentation

channels that are available and that are

required to be operable in Table 3.5.1–

1 would be increased; (3) the minimum

power level specified in Note (c) of

Table 3.5.1–1 would be changed from

10¥10 amps on the intermediate range

instrument channels to 4 x 10¥4% rated

power on the wide range instrument

channels; and (4) entries that specify the

Wide Range Nuclear Instrumentation,

the number of Required Operable

Channels, reference to a new Action

Statement, and Applicability would be

added to Table 3.5.6–1.

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) Will the change involve a significant

increase in the probability or consequences

of an accident previously evaluated?

No. The proposed amendment to the

Oconee Technical Specifications is

associated with the implementation of an

enhanced nuclear instrumentation system.

The new Gamma Metrics system provides

twice the number of channels of neutron

detectors for use during both normal plant

operations and post-accident monitoring. The

proposed change will make Oconee’s

Technical Specifications consistent with a

nuclear instrumentation system that meets

the reliability and redundancy requirements

of Regulatory Guide 1.97. Additionally, the

new Technical Specifications will be more

conservative in terms of stating the minimum

number of operable channels required, since

there are now a greater number of redundant

channels available. Assuring that the nuclear

instrumentation at Oconee is more reliable

and more redundant, does not affect the

probability of an occurrence of an accident,

since this system is a monitoring system and

not an accident initiator. However, these

characteristics (increased reliability and

redundancy) could provide additional

capability to deal with the consequences of

post-accident situations.

(2) Will the change create the possibility of

a new or different kind of accident from any

[kind of accident] previously evaluated?

No. The proposed amendment to Oconee

Technical Specifications involves the

implementation of an enhanced nuclear

instrumentation system. By implementing a

nuclear instrumentation system that meets

the provisions of Regulatory Guide 1.97,

Oconee’s ability for neutron monitoring is

enhanced during normal operations and postaccident recovery. The Source Range nuclear

instrumentation system is utilized for

monitoring purposes only, while the Wide

Range provides a control rod withdrawal

interlock based on high startup rate. The new

Gamma Metrics detectors have been shown

to be more reliable, accurate, and redundant

than Oconee’s original detectors. Therefore,

changing the Oconee Technical

Specifications to be consistent with the

current nuclear instrumentation arrangement,

as proposed in this amendment request, has

no effect on the possibility of any type of

accident: new, different, or previously

evaluated.

(3) Will the change involve a significant

reduction in a margin of safety?

No. Margin of safety is associated with

confidence in the ability to maintain the

fission product barriers (i.e., fuel and fuel

cladding, Reactor Coolant System pressure

boundary, and containment structure) to

limit the level of radiation dose to the public.

The proposed Technical Specifications

amendment will establish operability

requirements for an enhanced nuclear

instrumentation system at Oconee. By
Federal Register / Vol. 62, No. 38 / Wednesday, February 26, 1997 / Notices 8797

implementing a more reliable and redundant

nuclear instrumentation system, Oconee’s

post-accident monitoring capability is

enhanced. Therefore, the ability to protect

the public from radiation dose is further

assured, and no reduction in any existing

margin of safety will occur.

The NRC has reviewed the licensee’s

analysis and, based on this review, it

appears that the three standards of 10

CFR 50.92(c) are satisfied. Therefore, the

NRC staff proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Oconee County Library, 501

West South Broad Street, Walhalla,

South Carolina 29691.

Attorney for licensee: J. Michael

McGarry, III, Winston and Strawn, 1200

17th Street, NW., Washington, DC

20036.

NRC Project Director: Herbert N.

Berkow.

Entergy Operations, Inc., et al., Docket

No. 50–416, Grand Gulf Nuclear Station,

Unit 1, Claiborne County, Mississippi.

Date of amendment request: October

22, 1996.

Description of amendment request:

The proposed amendment would revise

Figure 3.4.11–1, ‘‘Minimum Reactor

Vessel Metal Temperature vs. Reactor

Vessel Pressure,’’ in Limiting Condition

for Operation 3.4.11, ‘‘RCS [Reactor

Coolant System] Pressure and

Temperature (P/T) Limits,’’ of the

Technical Specifications (TSs). The

existing curve is valid only up to 10

Effective Full Power Years (EFPYs) and

would be revised to be valid up to 32

EFPYs.

The proposed curves, pages 1 through

5 of Figure 3.4.11–1, have been drawn

for five different EFPY periods: 16, 20,

24, 28 and 32. There are two sets of

curves attached to the licensee’s

application. The first set of curves

(Attachment 3) would replace the

existing curve in TS Figure 3.4.11–1.

The second set of curves (Attachment 4)

are duplicates of the Attachment 3

curves except that these curves also

contain detailed information used in

development of the curves and would

be included in the next update of the

Updated Final Safety Analysis Report

(UFSAR) for information.

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, in its application for the

proposed amendment, which is

presented below:

(A) The proposed change does not

significantly increase the probability or

consequences of an accident previously

evaluated.

Regulatory Guide 1.99, Revision 2 is

currently used to prepare the pressuretemperature limit curves and is inherently

conservative for Boiling Water Reactors

(BWRs). [Grand Gulf Unit 1 is a BWR.] The

proposed Technical Specification Figure

3.4.11–1 was prepared in accordance with

the requirements of 10CFR50 [10 CFR Part

50], Appendix G [(Fracture Toughness

Requirements)], and using NRC approved

methodology outlined in NRC Regulatory

Guide 1.99, Revision 2, ‘‘Radiation

Embrittlement of Reactor Vessel Materials.’’

Operation of the plant within the limitations

of the proposed figure will ensure that the

Requirements of 10CFR50 [10 CFR Part 50],

Appendix G are met up to and including 32

Effective Full Power Years (EFPY) of

operation. The proposed changes assure that

the existing safety limits are not exceeded

due to changing Reactor Vessel conditions by

continued incorporation of the effect of

neutron radiation embrittlement of vessel

materials into the proposed curves.

The curves have also been editorially

enhanced by removal of phrases used for

validation of the curves. Having the phrases

on the TS (Technical Specification) curves

distracts from the intended purpose which is

to maintain operation of the reactor to the

right of the curves. Operators, in performance

of their job function, do not need this

information to comply with TS Limiting

Condition for Operation (LCO) 3.4.11. This

change also revises the curve labeling

consistent with the terminology used in

Table 1 of 10CFR50 [10 CFR Part 50],

Appendix G. These enhancements and

revisions have no impact on the operation of

the plant since they are editorial in nature

and do not change the technical content of

the curves.

Therefore, the proposed change does not

significantly increase the probability or

consequences of an accident previously

evaluated.

(B) The proposed change does not create

the possibility of a new or different kind of

accident from any accident previously

evaluated.

The pressure-temperature curves are

controlled by the Technical Specifications

and are determined using the conservative

methodology in NRC Regulatory Guide 1.99,

Revision 2, ‘‘Radiation Embrittlement of

Reactor Vessel Materials.’’ The proposed

pressure-temperature limit curves are

inherently conservative, therefore, the

possibility of failure of the reactor vessel is

not increased. The proposed curves establish

new periods of applicability (16, 20, 24, 28,

and 32 EFPY) for the current pressuretemperature limitations based on NRC

methodology in Regulatory Guide 1.99 and

actual fluence measurements. These

limitations are appropriate up to and

including 32 EFPY exposure and operation of

the plant within the figure’s limitations will

ensure that the requirements of 10CFR50 [10

CFR Part 50], Appendix G are met for that

time frame. No physical plant modifications

or new operating configurations result from

these changes. These changes do not

adversely affect the design or operation of

any system or component important to safety,

rather they establish limits to assure that

operations remain within acceptable safety

boundaries.

The curves have also been editorially

enhanced by removal of phrases used for

validation of the curves. Having the phrases

on the TS curves distracts from the intended

purpose which is to maintain operation of

the reactor to the right of the curves.

Operators, in performance of their job

function, do not need this information to

comply with TS Limiting Condition for

Operation (LCO) 3.4.11. This change also

revises the curve labeling consistent with the

terminology used in Table 1 of 10CFR50 [10

CFR Part 50], Appendix G. These

enhancements and revisions have no impact

on the operation of the plant since they are

editorial in nature and do not change the

technical content of the curves.

Therefore, the proposed change does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

(C) The proposed change does not involve

a significant reduction in a margin of safety.

The proposed curves were developed using

the methodology of Regulatory Guide 1.99,

Revision 2, ‘‘Radiation Embrittlement of

Reactor Vessel Materials.’’ This methodology

includes an allowance for margin that is to

be included in the upper-bound values of the

adjusted reference temperature (ART). The

proposed changes maintain the existing

margins of safety by modifying the operating

limits based on the most limiting of the

actual reference temperature shifts. These

new limits consider the most limiting

pressure vessel material. The revised analysis

demonstrates that the existing Technical

Specification [TS] pressure-temperature limit

curves are applicable for periods of 16, 20,

24, 28, and 32 EFPY. Using the methodology

in NRC Regulatory Guide 1.99 Revision 2 and

fluence based on actual exposure provides

for additional conservatism, and therefore [,]

further assures the existence of current

margins of safety. The proposed pressuretemperature limit curves are inherently

conservative and provide sufficient margin to

ensure the integrity of the reactor vessel.

The curves have also been editorially

enhanced by removal of phrases used for

validation of the curves. Having the phrases

on the TS curves distracts from the intended

purpose which is to maintain operation of

the reactor to the right of the curves.

Operators, in performance of their job

function, do not need this information to

comply with TS Limiting Condition for

Operation (LCO) 3.4.11. This change also

revises the curve labeling consistent with the

terminology used in Table 1 of 10CFR50 [10

CFR Part 50], Appendix G. These

enhancements and revisions have no impact

on the operation of the plant since they are

editorial in nature and do not change the

technical content of the curves.

Continuing commitment to the

methodology contained in NRC Regulatory

Guide 1.99, Rev. 2, will ensure that the most

limiting plate or beltline weld material will

be utilized in the determination of the

pressure-temperature limits for any future

curve changes.
8798 Federal Register / Vol. 62, No. 38 / Wednesday, February 26, 1997 / Notices

Therefore, the proposed change does not

result in a significant reduction in a margin

of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 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, NW., 12th Floor,

Washington, DC 20005–3502.

NRC Project Director: William D.

Beckner.

Entergy Gulf States, Inc., 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: January

10, 1997.

Description of amendment request:

The proposed amendment would revise

the Technical Specifications (TSs) for

reactor pressure vessel pressure and

temperature (P–T) limits to replace the

curves for 2 effective full power years

(EFPY) with curves for 12 EFPY. The P–

T curves are used for heatup, cooldown,

and inservice leak and hydrostatic

testing.

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 changes do not involve a

significant increase in the probability or

consequences of an accident previously

evaluated.

Pressure-temperature (P–T) limits (RBS

Technical Specifications Figure 3.4.11–1) are

imposed on the reactor coolant system to

ensure that adequate safety margins against

nonductile or rapidly propagating failure

exist during normal operation, anticipated

operational occurrences, and system

hydrostatic tests. The P–T limits are related

to the nil-ductility reference temperature,

RTNDT, as described in ASME Section III,

Appendix G. Changes in the fracture

toughness properties of [Reactor Pressure

Vessel] RPV beltline materials, resulting from

the neutron irradiation and the thermal

environment, are monitored by a surveillance

program in compliance with the

requirements of 10 CFR [Part] 50, Appendix

H. The effect of neutron fluence on the nilductility reference temperature of pressure

vessel steel is predicted by methods given in

Regulatory Guide [RG] 1.99, Rev. 2.

The revised P–T limits of this amendment

request were established based on adjusted

reference temperatures developed in

accordance with the procedures prescribed in

Reg. Guide [RG] 1.99, Rev. 2, Regulatory

Position C.1. Calculation of adjusted

reference temperature by these procedures

includes a margin term to ensure

conservative, upper-bound values are used

for the calculation of the P–T limits. Stress

intensity factors used to compute the

pressures were calculated in accordance

with, and include the required safety factors

given in ASME Section III, Appendix G. The

limits established by the lower portion of the

P–T curves, which cover the discontinuity

(non-beltline) regions of the vessel (e.g.,

flanges, nozzles, etc.), were retained

throughout this current analysis. The limits

established by the lower portion of these

curves do not change as they are not affected

significantly by the neutron fluence.

This change is not related to any accidents

previously evaluated. The proposed change

will provide for approved P–T limit curves

which are valid through 12 EFPY. This

change will not affect any Safety Limits,

Power Distribution Limits, or Limiting

Conditions for Operation. The proposed

change will not affect reactor pressure vessel

[RPV] performance as no physical changes

are involved and RBS vessel P–T limits will

remain conservative in accordance with Reg.

Guide [RG] 1.99, Rev. 2 and ASME Section

III, Appendix G requirements. The proposed

change will not cause the reactor pressure

vessel [RPV] or interfacing systems to be

operated outside of their design or testing

limits. Also, the proposed change will not

alter any assumptions previously made in

evaluating the radiological consequences of

accidents. The proposed change ensures that

adequate margins against brittle fracture of

the vessel are maintained through 12 EFPY

of reactor operations. Therefore, the

probability or consequences of accidents

previously evaluated will not be increased by

the proposed change.

2. The proposed changes do not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

The proposed change is a revision of

Technical Specification Figure 3.4.11–1 to

show P–T limit curves valid through 12

EFPY. The revised P–T limits have been

established in accordance with applicable

NRC regulations and the ASME Code. This

proposed change does not involve a

modification of the design of plant structures,

systems, or components. The proposed

change will not impact the manner in which

the plant is operated as plant operating and

testing procedures will not be affected by the

change. The proposed change will not

degrade the reliability of structures, systems,

or components important to safety (ITS) as

equipment protection features will not be

deleted or modified, equipment redundancy

or independence will not be reduced,

supporting system performance will not be

downgraded, the frequency of operation of

ITS equipment will not be imposed. No new

accident types or failure modes will be

introduced as a result of the proposed

change. Therefore, the proposed change does

not create the possibility of a new or different

kind of accident from that previously

evaluated.

3. The proposed changes do not involve a

significant reduction in a margin of safety.

As stated in the River Bend SER,

‘‘Appendices G and H of 10 CFR 50 describe

the conditions that require pressuretemperature [P–T] limits and provide the

general bases for these limits. These

appendices specifically require that pressuretemperature [P–T] limits must provide safety

margins at least as great as those

recommended in the ASME Code, Section III,

Appendix G. * * * Until the results from the

reactor vessel surveillance program become

available, the staff will use RG 1.99, Revision

1 [now Revision 2] to predict the amount of

neutron irradiation damage. * * * The use of

operating limits based on these criteria—as

defined by applicable regulations, codes, and

standards—will provide reasonable

assurance that nonductile or rapidly

propagating failure will not occur, and will

constitute an acceptable basis for satisfying

the applicable requirements of GDC 31.’’

Bases for RBS Technical Specification

3.4.11 states: ‘‘The P/T [P–T] limits are not

derived from Design Basis Accident (DBA)

analyses. They are prescribed during normal

operation to avoid encountering pressure,

temperature, and temperature rate of change

conditions that might cause undetected flaws

to propagate and cause nonductile failure of

the RCPB, a condition that is unanalyzed.

* * * Since the P/T [P–T] limits are not

derived from any DBA, there are no

acceptance limits related to the P/T [P–T]

limits. Rather, the P/T [P–T] limits are

acceptance limits themselves since they

preclude operation in an unanalyzed

condition.’’

This amendment request proposes P–T

limit curves which will be valid through 12

EFPY. The proposed P–T limits were

established based on adjusted reference

temperatures for vessel beltline material

calculated in accordance with Regulatory

Position 1 of Reg. Guide [RG] 1.99, Rev. 2 and

pressures calculated in accordance with

ASME Section III, Appendix G requirements.

Required margins and safety factors were

included to ensure that conservative, upperbound values were used in calculation of the

P–T limits. The proposed change will not

affect any Safety Limits, Power Distribution

Limits, or Limiting Conditions for Operation.

The proposed change does not represent a

change in initial conditions, or in a system

response time, or in any other parameter

affecting the course of an accident analysis

supporting the Bases of any Technical

Specification. The proposed P–T limits

provide adequate safety margins against

brittle failure of the reactor vessel through 12

EFPY of power operations. For these reasons,

the proposed changes do not involve a

reduction in any margins 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.
Federal Register / Vol. 62, No. 38 / Wednesday, February 26, 1997 / Notices 8799

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, NW., Washington, DC

20005.

NRC Project Director: William D.

Beckner.

Entergy Gulf States, Inc., 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: January

20, 1997.

Description of amendment request:

The proposed amendment would revise

the Technical Specifications (TSs) to

allow the use of flow control spectral

shift strategies to increase cycle energy;

an estimated additional 30 days at full

power. The request is based on a

General Electric (GE) Maximum

Extended Load Line Limit (MELLL)

analysis for the River Bend Station.

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 changes do not

significantly increase the probability or

consequences of an accident previously

evaluated.

Abnormal operational transients or

accidents analyzed in the SAR have been

examined for any impact caused by MELLL

operation. The limiting abnormal operation

transients, including the Generator Load

Rejection with No Bypass (LRNBP) event and

the Feedwater Controller Failure (FWCF)

maximum demand event, have been

evaluated in detail. The LOCA [Loss-ofCoolant Accident], Fuel Loading Error (FLE),

rod drop accident, rod withdrawal error, and

the Anticipated Transient Without Scram

(ATWS) analyses have also been evaluated

for the effects of MELLL operation. The flow

and power dependent [Minimum Critical

Power Ratio] MCPR curves for off-rated and

rated conditions and the [Maximum Average

Planar Linear Heat Generation Rate]

MAPLHGR criteria establish limits on power

operation. These limits ensure that the core

is operated within the assumptions and

initial conditions of the transient or accident

analyses. Operation within these limits will

ensure that the consequences of a transient

or accident remain within the acceptable

limits of the analyses.

The [Average Power Range Monitor] APRM

scram in the Technical Specifications [TSs]

and affected rod block setpoints are revised

to ensure that operation remains within the

analyzed MELLL region. This restriction

ensures the consequences of abnormal

operation and accidents are acceptable. The

probability of an accident is not affected by

the proposed Technical Specification [TS]

changes since no systems or equipment

which could initiate an accident are affected.

Therefore, the proposed changes do not

significantly increase the probability or

consequences of any previously evaluated

accident.

2. The request does not create the

possibility of occurrence of a new or different

kind of accident from any accident

previously evaluated.

Operation in the MELLL domain expands

the current power/flow along the 121% rod

line to 100% power at 75% rated core flow

and improves flexibility and capacity factor.

Abnormal operation transients or accidents

have been evaluated and the most limiting

cases have been analyzed for applicability for

operation in the MELLL region. The

proposed Technical Specification [TS]

changes prohibit power operation outside the

MELLL region and do not constitute or

require any system or equipment changes

that might create an accident of a different

type then previously evaluated. The

MAPLHGR, the power and flow dependent

MCPR and [Liner Heat Generation Rate]

LHGR and the revised Technical

Specifications [TSs] will continue to assure

that plant operation is consistent with the

assumptions, initial conditions and assumed

power distribution and therefore will not

create a new type of accident. The proposed

Technical Specification [TS] changes do not

introduce any new modes of plant operation

nor involve new system interactions.

Therefore, the proposed changes do not

create the possibility of a new or different

kind of accident from any previous analyzed.

3. The request does not involve a

significant reduction in a margin of safety.

The proposed Technical Specifications

[TSs] prohibit power operation outside the

allowable MELLL region. The transients and

accidents described in the SAR are evaluated

for operation in the MELLL region. NEDC–

32611, ‘‘MELLL Analysis for River Bend

Station Reload 6 Cycle 7,’’ shows that the

OLMCPR for operation in the MELLL region

is bounded by the OLMCPR established for

current conditions (100% power/107% flow).

The thermal limits MCPR and LHGR curves

and the MAPLHGR limits establish limits on

power operation and thereby ensure that the

core is operated within the assumptions and

initial conditions of the transient and

accident analyses.

As demonstrated in the analysis provided

in Attachment 4, [the proposed amendment

request] operation within these limits, using

the MCPR limits, LHGH limits and

MAPLHGR criteria, will ensure that the

margin of safety will be maintained to the

same level described in the Technical

Specifications Bases and the SAR and the

consequences of the postulated transient or

accidents are not increased. The MCPR safety

limit, mechanical performance limits and

overpressure limit are not exceeded during

any transient or postulated accident.

Therefore, the proposed Technical

Specifications [TSs] to allow operation in the

MELLL region do not involve a significant

reduction in 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.

Maine Yankee Atomic Power

Company, Docket No. 50–309, Maine

Yankee Atomic Power Station, Lincoln

County, Maine.

Date of amendment request: February

7, 1997.

Description of amendment request:

The proposed amendment would

modify Technical Specification 3.12 to

require both 115 kV incoming lines to

be operable when the reactor is critical;

allow continued operations for up to 72

hours with one 115 kV incoming line

inoperable; allow continued operations

for up to 24 hours with both 115 kV

incoming lines inoperable; apply the

increased operability requirements

described above to another affected

remedial action; incorporate minor

editorial changes to uniformly apply the

usage of the term ‘‘operable;’’ and

change the basis section to be consistent

with the 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:

1. The proposed change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

The proposed changes to Specification

3.12.B do not involve a physical change to

the plant or the maintenance of the plant.

The proposed changes increase the operating

requirements associated with the operability

of the 115 kV incoming lines beyond that

currently required by Technical

Specifications. For those accidents

previously evaluated, the more restrictive

operability requirements associated with

maintaining both 115 kV incoming lines

operable and the more restrictive remedial

action times result in increased assurance

that station service power will be available

when required. This increased availability

will be achieved because elective

maintenance on the offsite power system will

be significantly restricted and the restoration

of inoperable 115 kV incoming lines will be

treated with greater urgency. The increased
8800 Federal Register / Vol. 62, No. 38 / Wednesday, February 26, 1997 / Notices

assurance of availability will result in a

decrease in the probability or consequences

of these postulated accidents.

However, the more restrictive remedial

action times decrease the restoration period

and consequently increase the possibility

that successful restoration may not be

achieved, given an outage of the 115 kV

power system. A unit shutdown without

offsite power would then be commenced.

This evolution would involve a unit

shutdown without the availability of

equipment such as the reactor coolant

pumps, condensate pumps and main

feedwater pumps. Although none of these

components are credited as available for the

mitigation of the consequences of accidents

previously evaluated, the probability of the

occurrence of certain accidents is increased

without them.

Although the combination of these

considerations could involve an increase in

the probability of accidents previously

evaluated, the increase would not be

significant due to the low probability of

independent failures or common cause

failures of both of the 115 kV incoming lines.

There is no increase in the consequences of

any accident previously evaluated as a result

of these proposed Technical Specification

changes. The proposed Technical

Specification changes are consistent with the

Standard Technical Specifications approved

by the NRC. The proposed changes, therefore,

will not involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. The proposed change does not create the

possibility of a new or different kind of

accident from any previously evaluated.

The proposed Technical Specification

change does not involve a change to the

physical plant or to the physical

configuration of the offsite power system.

The effect of the proposed change will be to

increase the availability of the offsite power

system when required. In addition, the

proposed change will increase the possibility

of a unit shutdown without offsite power

operable. However, the accidents previously

evaluated assume a simultaneous loss of

offsite power, design basis accident and

worst case single failure as part of the design

basis. The proposed changes do not result in

the creation of a unique operating condition

or a configuration that has not been

previously evaluated. 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 the margin of

safety.

This proposed change modifies Technical

Specification 3.12 to be consistent with the

Standard Technical Specifications. The

proposed Technical Specification change

maintains the current margin of safety which

is based upon supplying power to engineered

safeguards. Adequate sources of power

remain available for the operation of the

engineered safeguards equipment. Therefore,

the proposed change 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 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: Wiscasset Public Library, High

Street, P.O. Box 367, Wiscasset, ME

04578.

Attorney for licensee: Mary Ann

Lynch, Esquire, Maine Yankee Atomic

Power Company, 329 Bath Road,

Brunswick, ME 04011.

NRC Project Director: Patrick D.

Milano, Acting.

Northeast Nuclear Energy Company,

et al., Docket Nos. 50–245, 50–336, and

50–423, Millstone Nuclear Power

Station, Unit Nos. 1, 2, and 3, New

London, Connecticut.

Date of amendment request: February

3, 1997.

Description of amendment request:

The licensee has proposed to revise

Section 6, ‘‘Administrative Controls,’’ of

the Millstone Unit Nos. 1, 2, and 3

Technical Specifications to reflect

organizational changes that have been

implemented in the Nuclear Division.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

* * * The proposed changes do not involve

a [significant hazards consideration] because

the changes would not:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

No design basis accidents are affected by

these proposed changes. The proposed

changes are administrative in nature and are

being proposed to reflect the organizational

changes which become effective on February

3, 1997. The unit level responsibilities of the

Executive Vice President—Nuclear are

assigned to the Officers for the individual

Millstone units. The site level

responsibilities of the Executive Vice

President—Nuclear are shared by the Senior

Vice President and CNO [Chief Nuclear

Officer]—Millstone and the President and

Chief Executive Officer. The changes to the

SORC [Site Operations Review Committee]

and the three unit[s’] PORC [Plant Operations

Review Committee] reflect changes in job

function or job position titles only.

No safety systems are adversely affected by

the proposed changes, and no failure modes

are associated with the changes.

2. Create the possibility of a new or

different kind of accident from any accident

previously evaluated.

Because there are no changes in the way

plants are operated due to this administrative

change, the potential for an unanalyzed

accident is not created. There is no impact

on plant response, and no new failure modes

are introduced. These proposed

administrative and editorial changes have no

impact on safety limits or design basis

accidents, and they have no potential to

create a new or unanalyzed event. The

changes to the SORC and the three unit[s’]

PORC reflect changes in job function or job

position titles only.

3. Involve a significant reduction in a

margin of safety.

The changes do not directly affect any

protective boundaries nor do they impact the

safety limits for the protective boundaries.

These proposed changes are administrative

and editorial in nature. Therefore, there is no

reduction in the margin of safety. These

changes do not reduce the margin of safety

provided by the PORC and the SORC review

and approval of changes to the operations of

the Millstone Unit Nos. 1, 2, and 3.

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 the 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 Deputy Director: Phillip F.

McKee.

Northeast Nuclear Energy Company

(NNECO), et al., Docket No. 50–423,

Millstone Nuclear Power Station, Unit

No. 3, New London County,

Connecticut.

Date of amendment request: February

5, 1996.

Description of amendment request:

The amendment would delete a clause

from Technical Specification 4.0.5.a.

Specifically, this change would delete

the clause ‘‘(g), except where specific

written relief has been granted by the

Commission pursuant to 10 CFR Part 50,

Section 50.55a(g)(6)(i).’’ The

amendment would also make the

appropriate changes to the Bases

section. In addition, NNECO made

changes to Bases Section 3/4.7.7 and 3/

4.7.8 to add design basis information

and provide clarification of system

design and operation.

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
Federal Register / Vol. 62, No. 38 / Wednesday, February 26, 1997 / Notices 8801

consideration, which is presented

below:

Pursuant to 10 CFR 50.92, NNECO has

reviewed the proposed changes to Technical

Specification 4.0.5a and Bases Section 3/

4.4.10 and has concluded that the changes do

not involve a significant hazards

consideration (SHC). The basis for this

conclusion is that the three criteria of 10 CFR

50.92(c) are not compromised. The proposed

changes do not involve an SHC because the

changes would not:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed changes would remove the

wording ‘‘* * * (g), except where specific

written relief has been granted by the

Commission pursuant to 10 CFR Part 50,

Section 50.55a(g)(6)(i).’’ The Inservice

Inspection and Testing Programs are

described in the technical specifications

pursuant to 10 CFR 50.55a. In addition, the

proposed changes, in accordance with

NUREG–1431 and NUREG–1482, would

provide relief to the ASME Code requirement

in the interim between the time of submittal

of a relief request until the NRC has issued

a safety evaluation and granted the relief. The

changes being proposed are administrative in

nature and do not affect assumptions

contained in plant safety analyses, the

physical design and/or operation of the plant,

nor do they affect any technical specification

that preserves safety analysis assumptions.

Any relief from the approved ASME Section

XI Code requirements will require a 10 CFR

50.59 evaluation to ensure no technical

specification changes or unreviewed safety

questions exist. Therefore, operation of the

facility in accordance with the proposed

changes would not affect the probability or

consequences of an accident previously

analyzed.

2. Create the possibility of a new or

different kind of accident from any

previously evaluated.

The proposed changes would remove the

wording ‘‘* * * (g), except where specific

written relief has been granted by the

Commission pursuant to 10 CFR Part 50,

Section 50.55a(g)(6)(i).’’ The Inservice

Inspection and Testing Programs are

described in the technical specifications

pursuant to 10 CFR 50.55a. In addition, the

proposed changes, in accordance with

NUREG–1431 and NUREG–1482, would

provide relief to the ASME Code requirement

in the interim between the time of submittal

of a relief request until the NRC has issued

a safety evaluation and granted relief. The

changes being proposed are administrative in

nature and will not change the physical plant

or the modes of operation defined in the

facility license. The changes do not involve

the addition or modification of equipment

nor do they alter the design or operation of

plant systems. Any relief from the approved

ASME Section XI Code requirements will

require a 10 CFR 50.59 evaluation to ensure

no technical specification changes or

unreviewed safety questions exist. Therefore,

operation of the facility in accordance with

the proposed changes would not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

3. Involve a significant reduction in the

margin of safety.

The proposed changes would remove the

wording ‘‘* * * (g), except where specific

written relief has been granted by the

Commission pursuant to 10 CFR Part 50,

Section 50.55a(g)(6)(i).’’ The Inservice

Inspection and Testing Programs are

described in the technical specifications

pursuant to 10 CFR 50.55a. In addition, the

proposed changes, in accordance with

NUREG–1431 and NUREG–1482, would

provide relief to the ASME Code requirement

in the interim between the time of submittal

of a relief request until the NRC has issued

a safety evaluation and granted relief. The

changes being proposed are administrative in

nature and will not alter the bases for

assurance that safety-related activities are

performed correctly or the basis for any

technical specification that is related to the

establishment or maintenance of a safety

margin. Any relief from the approved ASME

Section XI Code requirements will require a

10 CFR 50.59 evaluation to ensure no

technical specification changes or

unreviewed safety questions exist. Therefore,

operation of the facility in accordance with

the proposed changes 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 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Learning Resources Center,

Three Rivers Community-Technical

College, 574 New London Turnpike,

Norwich, Connecticut, and the

Waterford Library, ATTN: Vince

Juliano, 49 Rope Ferry Road, Waterford,

Connecticut.

Attorney for licensee: Lillian M.

Cuoco, Esq., Senior Nuclear Counsel,

Northeast Utilities Service Company,

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

NRC Deputy Director: Phillip F.

McKee.

United States Department of

Commerce, National Institute of

Standards and Technology, Docket No.

50–184, NIST (formerly known as

National Bureau of Standards) Test

Reactor or NBSR.

Date of amendment request: January

17, 1997.

Description of amendment request:

The National Institute of Standards and

Technology (NIST) is planning to

change the name of the Reactor

Radiation Division to the NIST Center

for Neutron Research to be headed by a

Director. The requested amendment

involves a name change only. All

functions, responsibilities, and

personnel remain the same. The

Technical Specification references to

the ‘‘Chief, Reactor Radiation Division’’

will be changed to Director, NIST Center

for Neutron Research in Sections 7.1,

7.2, and 7.3. The Organization Chart in

Figure 7.1 will also reflect this change.

The Technical Specification references

to the ‘‘Reactor Radiation Division’’ will

be changed to ‘‘NIST Center for Neutron

Research’’ in Section 7.2.

Basis for proposed no significant

hazards consideration determination:

The Commission has provided

standards for determining whether a

significant hazards consideration exists

(10 CFR 50.92(c)). A 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 change being proposed is a

change in the title of the organization

and the title of the head of the

organization that directs the operation

of the reactor. As noted previously, all

functions, responsibilities and

personnel remain the same. The staff

agrees with the licensee’s no significant

hazards consideration and finds that the

mere title changes render a negative

response to the three criteria outlined in

10 CFR 50.92(c).

Local Public Document Room

location: N/A.

Attorney for licensee: N/A

NRC Project Director: Seymour H.

Weiss.

Vermont Yankee Nuclear Power

Corporation, Docket No. 50–271,

Vermont Yankee Nuclear Power Station,

Vernon, Vermont.

Date of amendment request:

December 10, 1996.

Description of amendment request:

The proposed amendment would move

fire protection requirements from the

Vermont Yankee Technical

Specifications to the Fire Protection

Plan and the final safety analysis report

(FSAR), in accordance with the

guidance in NRC Generic Letters 86–10

and 88–12.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. The proposed amendment will not

involve a significant increase in the

probability or consequences of an accident

previously evaluated:
8802 Federal Register / Vol. 62, No. 38 / Wednesday, February 26, 1997 / Notices

The proposed changes are administrative

in nature and are consistent with the

guidance provided in NRC Generic Letters

86–10 and 88–12. These changes do not

affect the initial conditions or precursors

assumed in the FSAR safety analyses. These

proposed changes also do not decrease the

effectiveness of equipment relied upon to

mitigate the previously evaluated accidents.

Programmatic controls will continue to

assure that fire protection program changes

do not reduce the effectiveness of the

program to achieve and maintain safe

shutdown in the event of a fire.

2. The proposed amendment will not

create the possibility of a new or different

kind of accident from an accident previously

evaluated:

The proposed changes do not modify any

plant equipment, there is no reduction in fire

protection requirements, there is no change

in operating procedure and surveillance

requirements and no reduction in

administrative control or equipment

reliability. Therefore, implementation of the

proposed change will not affect the design

function or configuration of any component,

introduce any new operating scenarios,

failure modes or accident initiators.

3. The proposed amendment will not

involve a significant reduction in a margin of

safety:

The proposed amendment does not involve

a reduction to the Fire Protection Program.

The fire protection requirements are simply

being relocated to other controlled

documents. There are no equipment

modifications being proposed, only the

location of fire protection requirements,

which is administrative in nature.

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: Brooks Memorial Library, 224

Main Street, Brattleboro, VT 05301.

Attorney for licensee: R. K. Gad, III,

Ropes and Gray, One International

Place, Boston, MA 02110–2624.

NRC Project Director: Patrick D.

Milano, Acting Director.

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.

Carolina Power & Light Company, et

al., Docket No. 50–400, Shearon Harris

Nuclear Power Plant, Unit 1, Wake and

Chatham Counties, North Carolina.

Date of amendment request: January

10, 1997.

Brief description of amendment: The

proposed change would revise

Technical Specification 4.8.1.1.2 to

clarify pressure testing requirements for

the isolable and non-isolable portions of

the diesel fuel oil piping.

Date of publication of individual

notice in Federal Register: February 5,

1997 (62 FR 5490).

Expiration date of individual notice:

March 6, 1997.

Local Public Document Room

location: Cameron Village Regional

Library, 1930 Clark Avenue, Raleigh,

North Carolina 27605.

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 amendment request:

November 6, 1996.

Description of amendment request:

The proposed amendments would

revise the Technical Specifications

governing the cooling water system. The

changes are proposed to improve plant

operation based on operational

experience with the vertical motordriven cooling water pump. The

changes are also proposed to

incorporate information gathered by the

licensee during its self-assessment

Service Water System Operational

Performance Inspection (SWSOPI)

completed in late 1995.

Date of individual notice in the

Federal Register: January 29, 1997 (62

FR 4338).

Expiration date of individual notice:

February 28, 1997.

Local Public Document Room

location: Minneapolis Public Library,

Technology and Science Department,

300 Nicollet Mall, Minneapolis,

Minnesota 55401.

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 amendment requests: January

29, 1997.

Description of amendment requests:

The proposed amendments would

change the Bases for Technical

Specifications and the licensing basis

for the Operating Licenses relating to

the cooling water system emergency

intake line flow capacity. The licensee

determined through testing that the

emergency intake line flow capacity was

less than the design value stated in the

Updated Final Safety Analysis Report.

The proposed changes reflect the use of

operator actions to control cooling water

system flow following a seismic event.

The proposed changes also reclassify

the intake canal for use during a seismic

event, which would be an additional

source of cooling water during a seismic

event.

Date of individual notice in the

Federal Register: February 7, 1997 (62

FR 5857).

Expiration date of individual notice:

March 10, 1997. NSHC comments:

February 24, 1997.

Attorney for licensee: Jay Silberg, Esq.,

Shaw, Pittman, Potts, and Trowbridge,

2300 N Street, NW, Washington, DC

20037.

Local Public Document Room

location: Minneapolis Public Library,

Technology and Science Department,

300 Nicollet Mall, Minneapolis,

Minnesota 55401.

Public Service Electric & Gas

Company, Docket Nos. 50–272 and 50–

311, Salem Nuclear Generating Station,

Unit Nos. 1 and 2, Salem County, New

Jersey Date of amendment request:

January 31, 1997.

Brief description of amendment

request: The amendment would make

changes to Technical Specification (TS)

3.4.3, ‘‘Relief Valves,’’ for Salem Unit 1,

and TS 3.4.5, ‘‘Relief Valves,’’ for Salem

Unit 2, to ensure that the automatic

capability of the power operated relief

valves to relieve pressure is maintained

when these valves are isolated by

closure of the block valves.

Date of publication of individual

notice in Federal Register: February 7,

1997 (62 FR 5861).

Expiration date of individual notice:

March 10, 1997.

Local Public Document Room

location: Salem Free Public Library, 112

West Broadway, Salem, NJ 08079.

Tennessee Valley Authority, Docket

Nos. 50–327 and 50–328, Sequoyah

Nuclear Plant, Units 1 and 2: Hamilton

County, Tennessee.

Date of application for amendments:

October 18, 1996.

Description of amendments request:

Amend Technical Specifications to

permanently incorporate new

requirements associated with steam

generator tube inspections and repair.

The requirements provide alternate
Federal Register / Vol. 62, No. 38 / Wednesday, February 26, 1997 / Notices 8803

steam generator tube plugging criteria at

the tube support plate intersections.

Date of publication of individual

notice in the Federal Register: February

11, 1997 (62 FR 6276).

Expiration date of individual notice:

March 13, 1997.

Local Public Document Room

location: Chattanooga-Hamilton County

Library, 1001 Broad Street, Chattanooga,

Tennessee 37402.

Wisconsin Electric Power Company,

Docket Nos. 50–266 and 50–301, Point

Beach Nuclear Power Plant, Unit Nos. 1

and 2, Town of Two Creeks: Manitowoc

County, Wisconsin.

Date of amendment requests:

September 19, 1996, as supplemented

November 18, 1996, and revised January

13 and January 27, 1997.

Description of amendment requests:

The proposed amendments would

change Technical Specification

requirements related to the low

temperature overpressure protection

(LTOP) system. Specifically, the reactor

coolant system (RCS) temperature below

which LTOP is required to be enabled

and the temperature below which one

high pressure safety injection pump is

required to be rendered inoperable

would be changed from less than 275

degrees Fahrenheit to less than 355

degrees Fahrenheit. Additionally, the

restriction of ‘‘less than the minimum

pressurization temperature for the

inservice pressure test as defined in

Figure 15.3.1–1’’ would be deleted and

the specific temperature limit of less

than 355 degrees Fahrenheit would be

specified. The setpoint for the

pressurizer power-operated relief valves

(PORVs) would be changed from less

than or equal to 425 pounds per square

inch gage (psig) to less than or equal to

440 psig to allow for instrument

inaccuracies and increased margin

allowed by the use of American Society

of Mechanical Engineers (ASME) Boiler

and Pressure Vessel Code Case N–514.

These modified requirements for LTOP

ensure that RCS materials meet the

requirements of Title 10 of the Code of

Federal Regulations, § 50.60,

‘‘Acceptance Criteria for Fracture

Prevention Measures for Lightwater

Nuclear Power Reactors for Normal

Operation,’’ (10 CFR 50.60) in

accordance with 10 CFR Part 50,

Appendices G and H, and in accordance

with the exemption granted on January

27, 1997, which allows the use of ASME

Code Case N–514 as an acceptable

alternative. Finally, editorial changes

would be made to rename the

‘‘Overpressure Mitigating System’’ to

the ‘‘Low Temperature Overpressure

Protection System.’’ The September 19,

1996, application was previously

noticed in the Federal Register on

October 1, 1996 (61 FR 51308).

Date of individual notice in the

Federal Register: February 4, 1997 (62

FR 5256).

Expiration date of individual notice:

March 6, 1997. NSHC comments

February 19, 1997.

Local Public Document Room

location: Joseph P. Mann Library, 1516

Sixteenth Street, Two Rivers, Wisconsin

54241.

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.

Baltimore Gas and Electric Company,

Docket Nos. 50–317 and 50–318, Calvert

Cliffs Nuclear Power Plant, Unit Nos. 1

and 2, Calvert County, Maryland.

Date of application for amendments:

November 26, 1996.

Brief description of amendments: The

amendments adopt Option B of 10 CFR

Part 50, Appendix J to require Type B

and Type C containment leakage testing

to be performed on a performance-based

testing schedule.

Date of issuance: February 11, 1997.

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment Nos.: 219 and 196.

Facility Operating License Nos. DPR–

53 and DPR–69: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: January 2, 1997 (62 FR 123).

The Commission’s related evaluation

of these amendments is contained in a

Safety Evaluation dated February 11,

1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Calvert County Library, Prince

Frederick, Maryland 20678.

Boston Edison Company, Docket No.

50–293, Pilgrim Nuclear Power Station,

Plymouth County, Massachusetts.

Date of application for amendment:

April 25, 1996, as supplemented

December 23, 1996.

Brief description of amendment: The

amendment will revise the definition of

Operable-Operability, revise Technical

Specifications (TSs) and associated

Bases Section for TS 3.9.B.2 and 3.9.B.3,

‘‘Auxiliary Electrical System,’’ TS

3.4.B.1, ‘‘Standby Liquid Control

System,’’ 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 TSs. 4.5.F.1, ‘‘Core and

Containment Cooling Systems,’’ and

delete TS 3.7.b.1.f, ‘‘Standby Gas

Treatment System and Control Room

High Efficiency Air Filtration System.’’

Date of issuance: February 10, 1997.

Effective date: February 10, 1997.

Amendment No.: 170.

Facility Operating License No. DPR–

35: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: June 19, 1996 (61 FR 31172).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated February 10,

1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Plymouth Public Library, 11

North Street, Plymouth, Massachusetts

02360.

Commonwealth Edison Company,

Docket Nos. STN 50–454 and STN 50–
8804 Federal Register / Vol. 62, No. 38 / Wednesday, February 26, 1997 / Notices

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

Ogle County, Illinois.

Docket Nos. STN 50–456 and STN 50–

457, Braidwood Station, Unit Nos. 1 and

2, Will County, Illinois.

Date of application for amendments:

August 2, 1996.

Brief description of amendments: The

amendments eliminate License

Condition 2.C.(16) from Facility

Operating License NPF–37; License

Condition 2.C.(5) from Facility

Operating License NPF–66; License

Condition 2.C.(6) from Facility

Operating License NPF–72 and License

Condition 2.C.(5) from Facility

Operating License NPF–77 that require

the licensee to conduct additional

corrosion testing of sleeved steam

generator tubes.

Date of issuance: February 12, 1997.

Effective date: Immediately, to be

implemented within 30 days.

Amendment Nos.: 85 to NPF–37, 85 to

NPF–66, 77 to NPF–72, and 77 to NPF–

77.

Facility Operating License Nos. NPF–

37, NPF–66, NPF–72 and NPF–77: The

amendments revise the licenses.

Date of initial notice in Federal

Register: September 25, 1996 (61 FR

50340).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated February 12,

1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: For Byron, the Byron Public

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

Box 434, Byron, Illinois 61010; for

Braidwood, the Wilmington Public

Library, 201 S. Kankakee Street,

Wilmington, Illinois 60481.

Duke Power Company, Docket Nos.

50–369 and 50–370, McGuire Nuclear

Station, Units 1 and 2, Mecklenburg

County, North Carolina.

Date of application for amendments:

November 26, 1996, as supplemented

December 17, 1996

Brief description of amendments: The

amendments revise Technical

Specification 3.8.2.1 to allow a one-time

change to replace the existing 125-volt

AT&T high specific gravity round cell

battery banks with the conventional low

specific gravity cell battery banks.

Date of issuance: February 7, 1997.

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment Nos.: 172 and 154.

Facility Operating License Nos. NPF–

9 and NPF–17: Amendments revised the

Technical Specifications.

Date of initial notice in Federal

Register: December 13, 1996 (61 FR

65605).

The December 17, 1996, letter did not

change the scope of the November 26,

1996, 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 February 7,

1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: J. Murrey Atkins Library,

University of North Carolina at

Charlotte, 9201 University City

Boulevard, Charlotte, North Carolina

28223–0001.

Entergy Gulf States, Inc., 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: August

29, 1996.

Brief description of amendment: The

amendment revises the Technical

Requirements Manual (TRM) to change

the reactor pressure vessel surveillance

capsule withdrawal schedule for the

River Bend Station. The first capsule

will be withdrawn at 10.4 effective full

power years (EFPY) rather than at 6

EFPY.

Date of issuance: February 13, 1997.

Effective date: February 13, 1997.

Amendment No.: 92.

Facility Operating License No. NPF–

47. The amendment revised the

Technical Requirements Manual.

Date of initial notice in Federal

Register: October 23, 1996 (61 FR

55034) The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

February 13, 1997.

No significant hazards consideration

comments received. No.

Local Public Document Room

location: Government Documents

Department, Louisiana State University,

Baton Rouge, LA 70803.

Entergy Operations, Inc., Docket No.

50–382, Waterford Steam Electric

Station, Unit 3, St. Charles Parish,

Louisiana.

Date of amendment request: June 27,

1996.

Brief description of amendment: The

amendment modifies TS 3/4.3.3.6,

‘‘Accident Monitoring Instrumentation,’’

to reflect the Combution Engineering

improved Standard Technical

Specification (STS) approved and

issued as NUREG–1432. This

amendment revises the TS to include

Accident Monitoring Instrumentation

recommended in Regulatory Guide (RG)

1.97, ‘‘Instrumentation for Light-WaterCooled Nuclear Plants to Assess Plant

Conditions During and Following an

Accident,’’ Revision 3.

Date of issuance: February 12, 1997.

Effective date: February 12, 1997, to

be implemented within 90 days.

Amendment No.: 122.

Facility Operating License No. NPF–

38. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: July 3, 1996 (61 FR 40017).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated February 12,

1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: University of New Orleans

Library, Louisiana Collection, Lakefront,

New Orleans, LA 70122.

Entergy Operations, Inc., Docket No.

50–382, Waterford Steam Electric

Station, Unit 3, St. Charles Parish,

Louisiana.

Date of amendment request: July 25,

1996, as supplemented by letter dated

January 27, 1997.

Brief description of amendment: The

amendment changes the Appendix A

Technical Specifications by modifying

TS 3/4.7.4, ‘‘Ultimate Heat Sink,’’ to

incorporate more restrictive fan

operability requirements and lower the

maximum allowed basin temperature.

Date of issuance: February 13, 1997.

Effective date: February 13, 1997.

Amendment No.: 123.

Facility Operating License No. NPF–

38. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: November 19, 1996 (61 FR

58903).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated February 13,

1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: University of New Orleans

Library, Louisiana Collection, Lakefront,

New Orleans, LA 70122.

Florida Power 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:

October 30, 1996.

Brief description of amendments:

These amendments revise the St. Lucie

Technical Specifications to remove

inconsistencies between the definition

of Core Alterations and the

Applicability, Action and Surveillance

requirements of two specifications

relating to water level and containment
Federal Register / Vol. 62, No. 38 / Wednesday, February 26, 1997 / Notices 8805

isolation systems during refueling

operations.

Date of Issuance: February 10, 1997.

Effective Date: February 10, 1997.

Amendment Nos.: 148 and 87.

Facility Operating License Nos. DPR–

67 and NPF–16: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: December 4, 1996 (61 FR

64386).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated February 10,

1997.

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 amendment:

October 28, 1996.

Brief description of amendment: The

amendments consist of changes to the

Technical Specifications (TS) in

response to your applications, both

dated October 28, 1996, regarding

containment leakage tests and removal

of certain component lists from the TS.

Date of Issuance: February 10, 1997.

Effective Date: February 10, 1997.

Amendment Nos.: 149 and 88.

Facility Operating License No. NPF–

16: Amendments revised the Technical

Specifications.

Date of initial notice in Federal

Register: (61 FR 64386) December 4,

1996. The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

February 10, 1997.

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,

Docket Nos. 50–250 and 50–251, Turkey

Point Plant Units 3 and 4, Dade County,

Florida.

Date of application for amendments:

December 17, 1996.

Brief description of amendments:

Revision to Technical Specification (TS)

4.4.10 regarding reactor coolant pump

flywheel inspection intervals.

Date of issuance: February 11, 1997.

Effective date: February 11, 1997.

Amendment Nos.: 193 and 187.

Facility Operating Licenses Nos. DPR–

31 and DPR–41: Amendments revise the

Technical Specifications.

Date of initial notice in Federal

Register: January 10, 1997 (62 FR 1476).

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated February 11, 1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Florida International

University, University Park, Miami,

Florida 33199.

Niagara Mohawk Power Corporation,

Docket No. 50–220, Nine Mile Point

Nuclear Station Unit No. 1, Oswego

County, New York.

Date of application for amendment:

July 16, 1996.

Brief description of amendment: The

amendment changes the Technical

Specifications to permit the use of 10

CFR Part 50, Appendix J, Option B,

Performance-Based Containment

Leakage Rate Testing in accordance with

the implementation guidance in NRC’s

Regulatory Guide 1.163 dated

September 1995.

Date of issuance: February 10, 1997.

Effective date: As of the date of

issuance to be implemented within 30

days.

Amendment No.: 159.

Facility Operating License No. DPR–

63: Amendment revises the Technical

Specifications.

Date of initial notice in Federal

Register: October 9, 1996 (61 FR 52965).

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated February 10, 1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Reference and Documents

Department, Penfield Library, State

University of New York, Oswego, New

York 13126.

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:

August 15, 1996.

Brief description of amendments: The

amendments revise the containment

cooling systems limiting conditions for

operation technical specifications to

bring them into conformance with

recently completed system analyses by

no longer permitting both containment

spray pumps to be inoperable at the

same time.

Date of issuance: February 10, 1997.

Effective date: February 10, 1997,

with full implementation within 30

days.

Amendment Nos.: 125 and 117.

Facility Operating License Nos. DPR–

42 and DPR–60: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: December 4, 1996 (61 FR

64388).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated February 10,

1997.

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.

Philadelphia Electric Company,

Docket Nos. 50–352 and 50–353,

Limerick Generating Station, Units 1

and 2, Montgomery County,

Pennsylvania.

Date of application for amendments:

November 25, 1996.

Brief description of amendments:

These amendments revise the wording

in TS Section 4.8.1.1.2.e.2 and the

associated TS Bases Section 3/4.8, to

remove the specific reference to the

Residual Heat Removal (RHR) pump

motor and its corresponding kW rating

value, and replace it with wording

consistent with that specified in the

Improved TS (i.e., NUREG–1433,

Revision 1, ‘‘Standard Technical

Specifications General Electric Plants,’’

dated April 1995).

Date of issuance: February 4, 1997.

Effective date: Both units, as of date

of issuance, to be implemented within

30 days.

Amendment Nos.: 121 and 85.

Facility Operating License Nos. NPF–

39 and NPF–85. The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: December 18, 1996 (61 FR

66716).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated February 4,

1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Pottstown Public Library, 500

High Street, Pottstown, PA 19464.

Philadelphia Electric Company,

Docket Nos. 50–352 and 50–353,

Limerick Generating Station, Units 1

and 2, Montgomery County,

Pennsylvania.

Date of application for amendments:

September 27, 1996.

Brief description of amendments:

These amendments increase the reactor

enclosure secondary containment

maximum inleakage rate, and also

impact secondary containment

drawdown time and system flow rate

assumptions, thereby, affecting charcoal

filter bed efficiency and post accident

dose analysis.
8806 Federal Register / Vol. 62, No. 38 / Wednesday, February 26, 1997 / Notices

Date of issuance: February 11, 1997.

Effective date: Both units, as of the

date of issuance, to be implemented

within 30 days.

Amendment Nos.: 122 and 86.

Facility Operating License Nos. NPF–

39 and NPF–85. The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: December 4, 1996 (61 FR

64392).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated February 11,

1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Pottstown Public Library, 500

High Street, Pottstown, PA 19464.

Philadelphia Electric Company,

Docket No. 50–353, Limerick

GeneratingStation, Unit 2, Montgomery

County, Pennsylvania.

Date of application for amendment:

December 6, 1996, as supplemented by

letters dated January 15, and 28, 1997.

Brief description of amendment: This

amendment modifies Technical

Specification (TS) Section 2.1 and its

associated TS Bases to reflect the change

in the Minimum Critical Power Ratio

safety limit due to the use of GE13 fuel

product line and the cycle-specific

analysis performed by General Electric

Company (GE), for LGS, Unit 2, Cycle 5.

Date of issuance: February 12, 1997.

Effective date: As of date of issuance,

to be implemented within 30 days.

Amendment No.: 87.

Facility Operating License No. NPF–

85. This amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: December 23, 1996 (61 FR

67582).

The January 15, and 28, 1997, letters

provided clarifying information that did

not change the initial proposed no

significant hazards consideration

determination.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated February 12,

1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Pottstown Public Library, 500

High Street, Pottstown, PA 19464.

Public Service Electric & Gas

Company, Docket Nos. 50–272 and 50–

311, Salem Nuclear Generating Station,

Unit Nos. 1 and 2, Salem County, New

Jersey.

Date of application for amendments:

June 10, 1996, as supplemented June 24,

July 1, August 13, September 20, and

October 17, 1996.

Brief description of amendments: The

amendments change Technical

Specifications 3/4.3.3.1, ‘‘Radiation

Monitoring Instrumentation,’’ and 3/

4.7.6, ‘‘Control Room Emergency Air

Conditioning System,’’ to reflect a

control room design in which the

common Unit 1 and Unit 2 control room

envelope is supplied by 2 one hundred

percent capable Control Room

Emergency Air Conditioning System

trains.

Date of issuance: February 6, 1997.

Effective date: Both units, as of date

of issuance, to be implemented within

30 days.

Amendment Nos.: 190 and 173.

Facility Operating License Nos. DPR–

70 and DPR–75. The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: June 24, 1996 (61 FR 32468)

The June 24, July 1, August 13,

September 20, and October 17, 1996,

letters provided clarifying information

that did not change the initial proposed

no significant hazards consideration

determination nor expand the scope of

the initial submittal as described in the

initial notice.

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated February 6,

1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Salem Free Public Library, 112

West Broadway, Salem, NJ 08079.

Public Service Electric & Gas

Company, Docket Nos. 50–272 and 50–

311, Salem Nuclear Generating Station,

Unit Nos. 1 and 2, Salem County, New

Jersey.

Date of application for amendments:

May 31, 1996, as supplemented

December 23, 1996.

Brief description of amendments: The

amendments change the Technical

Specification to (1) Revise the reactor

vessel level indication system action

statements, (2) revise the channel

calibration definition, and (3) delete a

requirement to install la jumper in the

auxiliary feedwater actuation logic.

Date of issuance: February 6, 1997.

Effective date: Both units, as of its

date of issuance, to be implemented

within 60 days.

Amendment Nos.: 191 and 174.

Facility Operating License Nos. DPR–

70 and DPR–75. The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: June 17, 1996 (61 FR 30641).

The December 23, 1996, letter

provided clarifying information that did

not change the initial proposed no

significant hazards consideration

determination.

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated February 6,

1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Salem Free Public Library, 112

West Broadway, Salem, NJ 08079.

Southern Nuclear Operating

Company, Inc., Docket Nos. 50–348 and

50–364, Joseph M. Farley Nuclear Plant,

Units 1 and 2, Houston County,

Alabama.

Date of amendments request:

November 15, 1996.

Brief Description of amendments: The

amendments replace Containment

Systems TS 3.6.2.2 for the Spray

Additive System, with a new Emergency

Core Cooling Systems (ECCS) TS 3.5.6

for the ECCS Recirculation Fluid pH

Control System.

Date of issuance: February 3, 1997.

Effective date: As of the date of

issuance to be implemented prior to

Mode 4 for Unit 1 following the spring

1997 refueling outage; for Unit 2

following the spring 1998 refueling

outage.

Amendment Nos.: 123 and 118.

Facility Operating License Nos. NPF–

2 and NPF–8: Amendments revise the

Technical Specifications.

Date of initial notice in Federal

Register: December 18, 1996 (61 FR

66718).

The Commission’s related evaluation

of the amendments is contained in a

Safety Evaluation dated February 3,

1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: Houston-Love Memorial

Library, 212 W. Burdeshaw Street, Post

Office Box 1369, Dothan, Alabama

36302.

Toledo Edison Company, Centerior

Service Company, and The Cleveland

Electric Illuminating Company, Docket

No. 50–346, Davis-Besse Nuclear Power

Station, Unit No. 1, Ottawa County,

Ohio.

Date of application for amendment:

August 7, 1996.

Brief description of amendment: The

amendment revises Technical

Specification (TS) 1.0, ‘‘Definitions,’’ by

defining a refueling interval to be [less

than or equal to] 730 days; and revises

TS 3/4.0, ‘‘Applicability,’’ TS 3/4.6.2.1,

‘‘Containment Systems—

Depressurization and Cooling Systems—

Containment Spray System,’’ and TS 3/

4.6.3.1, ‘‘Containment Systems—
Federal Register / Vol. 62, No. 38 / Wednesday, February 26, 1997 / Notices 8807

Containment Isolation Valves,’’ to

reflect performing surveillance tests

during a refueling interval rather than

every 18 months.

Date of issuance: February 10, 1997.

Effective date: February 10, 1997, to

be implemented not later than 120 days

after issuance.

Amendment No.: 213.

Facility Operating License No. NPF–3:

Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: October 9, 1996 (61 FR 52970).

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated February 10,

1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: University of Toledo, William

Carlson Library, Government

Documents Collection, 2801 West

Bancroft Avenue, Toledo, Ohio 43606.

Toledo Edison Company, Centerior

Service Company, and The Cleveland

Electric Illuminating Company, Docket

No. 50–346, Davis-Besse Nuclear Power

Station, Unit No. 1, Ottawa County,

Ohio.

Date of application for amendment:

September 12, 1996.

Brief description of amendment: The

amendment revised Technical

Specifications (TS) 3/4.1.3.4,

‘‘Reactivity Control Systems—Rod Drop

Time,’’ and TS 3/4.5.2, ‘‘Emergency

Core Cooling Systems—Tavg [greater

than or equal to] 280°F,’’ to change the

surveillance test interval from every 18

months to each refueling interval ([less

than or equal to] 730 days, nominally 24

months). Additionally, the amendment

removed a footnote for TS 4.5.2.b that is

no longer applicable.

Date of issuance: February 11, 1997.

Effective date: February 11, 1997, to

be implemented not later than 120 days

over issuance.

Amendment No.: 214.

Facility Operating License No. NPF–3:

Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: October 9, 1996.

The Commission’s related evaluation

of the amendment is contained in a

Safety Evaluation dated February 11,

1997.

No significant hazards consideration

comments received: No.

Local Public Document Room

location: University of Toledo, William

Carlson Library, Government

Documents Collection, 2801 West

Bancroft Avenue, Toledo, Ohio 43606.

Dated at Rockville, Maryland, this 19th day

of February 1997.

For the Nuclear Regulatory Commission.

Jack W. Roe,

Director, Division of Reactor Projects—III/IV,

Office of Nuclear Reactor Regulation.

[FR Doc. 97–4573 Filed 2–25–97; 8:45 am]

BILLING CODE 7590–01–P

SECURITIES AND EXCHANGE

COMMISSION

Request For Public Comment

Upon Written Request, Copies

Available From: Securities and

Exchange Commission, Office of Filings

and Information Services, Washington,

DC 20549.

Reapproval:

Rule 24b–2

SEC File No. 270–153

OMB Control No. 3235–0127

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 24b–2 (17 CFR 240.25b–2)

provides a procedure, whereby persons

filing documents with the Commission

may request confidential treatment of

information contained in such

documents, and may request

Commission review of adverse staff

determinations regarding the

confidential treatment request.

Approximately 630 requests for

confidential treatment are made per

year. Applications pursuant to the rule

are generally prepared in conjunction

with the document for which

confidential treatment is being

requested. Based upon our review of the

applications we have received, we

believe that not more than 30 minutes

of the time spent in preparing the entire

filing may be attributed to the

application required under Rule 24b–2.

Thus, the total compliance burden is

315 hours. The approximate cost per

hour is $100, resulting in a total cost of

compliance for respondents of $31,500

per year (315 hours @$100).

Written comments are invited on: (a)

whether the proposed collection of

information is necessary for the proper

performance of the functions of the

agency, including whether the

information shall have practical utility;

(b) the accuracy of the agency’s estimate

of the burden of the proposed collection

of information; (c) ways to enhance the

quality, utility, and clarity of the

information to be collected; and (d)

ways to minimize the burden of the

collection of information on

respondents, including through the use

of automated collection techniques or

other forms of information technology.

Consideration will be given to

comments and suggestions submitted in

writing within 60 days of this

publication.

Direct your written comments to

Michael E. Bartell, Associate Executive

Director, Office of Information

Technology, Securities and Exchange

Commission, 450 5th Street, NW.,

Washington, DC 20549.

Dated: February 19, 1997.

Margaret H. McFarland,

Deputy Secretary.

[FR Doc. 97–4748 Filed 2–25–97; 8:45 am]

BILLING CODE 8010–01–M

[Rel. No. IC–22521; 813–152]

Partners Income Fund; Notice of

Application

February 20, 1997.

AGENCY: Securities and Exchange

Commission (‘‘SEC’’).

ACTION: Notice of application for

exemption under the Investment

Company Act of 1940 (the ‘‘Act’’).

APPLICANT: Partners Income Fund (the

‘‘Initial Partnership’’).

RELEVANT ACT SECTIONS: Order requested

under section 6(b).

SUMMARY OF APPLICATION: Applicant

requests an order that would amend a

prior order to permit the employer of

certain employees’ securities companies

to invest in those companies on terms

no more favorable than those available

to eligible employees.

FILING DATES: The application was filed

on August 6, 1996 and amended on

November 26, 1996.

HEARING OR NOTIFICATION OF HEARING: An

order granting the application will be

issued unless the SEC orders a hearing.

Interested persons may request a

hearing by writing to the SEC’s

Secretary and serving applicant with a

copy of the request, personally or by

mail. Hearing requests should be

received by the SEC by 5:30 p.m. on

March 17, 1997 and should be

accompanied by proof of service on the

applicant, in the form of an affidavit or,

for lawyers, a certificate of service.

Hearing requests should state the nature

of the writer’s interest, the reason for the

request, and the issues contested.

Persons may request notification of a

hearing by writing to the SEC’s

Secretary.

ADDRESSES: Secretary, SEC, 450 5th

Street, N.W., Washington, D.C. 20549.

Applicant, c/o McKinsey & Company,