174 Federal Register / Vol. 61, No. 2 / Wednesday, January 3, 1996 / Notices

present, may exchange preliminary

views regarding matters to be

considered during the balance of the

meeting.

The Subcommittee will then hear

presentations by and hold discussions

with representatives of the NRC staff,

the Westinghouse Electric Corporation,

their consultants, and other interested

persons regarding this review.

Further information regarding topics

to be discussed, whether the meeting

has been cancelled or rescheduled, the

scheduling of sessions which are open

to the public, the Chairman’s ruling on

requests for the opportunity to present

oral statements and the time allotted

therefor can be obtained by contacting

the cognizant ACRS staff engineer, Mr.

Paul A. Boehnert (telephone 301/415–

8065) between 7:30 a.m. and 4:15 p.m.

(EST). Persons planning to attend this

meeting are urged to contact the above

named individual one or two working

days prior to the meeting to be advised

of any potential changes in the proposed

agenda, etc., that may have occurred.

Dated: December 27, 1995.

Sam Duraiswamy,

Chief, Nuclear Reactors Branch.

[FR Doc. 96–00048 Filed 1–2–96; 8:45 am]

BILLING CODE 7590–01–P

Biweekly Notice

Applications and Amendments to

Facility Operating Licenses Involving

No Significant Hazards Considerations

I. Background

Pursuant to Public Law 97-415, the

U.S. Nuclear Regulatory Commission

(the Commission or NRC staff) is

publishing this regular biweekly notice.

Public Law 97-415 revised section 189

of the Atomic Energy Act of 1954, as

amended (the Act), to require the

Commission to publish notice of any

amendments issued, or proposed to be

issued, under a new provision of section

189 of the Act. This provision grants the

Commission the authority to issue and

make immediately effective any

amendment to an operating license

upon a determination by the

Commission that such amendment

involves no significant hazards

consideration, notwithstanding the

pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all

notices of amendments issued, or

proposed to be issued from December

11, 1995, through December 20, 1995.

The last biweekly notice was published

on December 20, 1995 (60 FR 65672).

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 Rules Review and

Directives Branch, Division of Freedom

of Information and Publications

Services, Office of Administration, U.S.

Nuclear Regulatory Commission,

Washington, DC 20555, 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 February 2, 1996, the licensee may

file a request for a hearing with respect

to issuance of the amendment to the

subject facility operating license and

any person whose interest may be

affected by this proceeding and who

wishes to participate as a party in the

proceeding must file a written request

for a hearing and a petition for leave to

intervene. Requests for a hearing and a

petition for leave to intervene shall be

filed in accordance with the

Commission’s ‘‘Rules of Practice for

Domestic Licensing Proceedings’’ in 10

CFR Part 2. Interested persons should

consult a current copy of 10 CFR 2.714

which is available at the Commission’s

Public Document Room, the Gelman

Building, 2120 L Street, NW.,

Washington, DC and at the local public

document room for the particular

facility involved. If a request for a

hearing or petition for leave to intervene

is filed by the above date, the

Commission or an Atomic Safety and

Licensing Board, designated by the

Commission or by the Chairman of the

Atomic Safety and Licensing Board

Panel, will rule on the request and/or

petition; and the Secretary or the

designated Atomic Safety and Licensing

Board will issue a notice of a hearing or

an appropriate order.

As required by 10 CFR 2.714, a

petition for leave to intervene shall set

forth with particularity the interest of

the petitioner in the proceeding, and

how that interest may be affected by the

results of the proceeding. The petition

should specifically explain the reasons

why intervention should be permitted

with particular reference to the

following factors: (1) the nature of the

petitioner’s right under the Act to be

made a party to the proceeding; (2) the

nature and extent of the petitioner’s

property, financial, or other interest in

the proceeding; and (3) the possible

effect of any order which may be

entered in the proceeding on the

petitioner’s interest. The petition should

also identify the specific aspect(s) of the

subject matter of the proceeding as to

which petitioner wishes to intervene.

Any person who has filed a petition for

leave to intervene or who has been

admitted as a party may amend the

petition without requesting leave of the

Board up to 15 days prior to the first

prehearing conference scheduled in the

proceeding, but such an amended

petition must satisfy the specificity

requirements described above.
Federal Register / Vol. 61, No. 2 / Wednesday, January 3, 1996 / Notices 175

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, Attention:

Docketing and Services Branch, or may

be delivered to the Commission’s Public

Document Room, the Gelman Building,

2120 L Street, NW., Washington DC, by

the above date. Where petitions are filed

during the last 10 days of the notice

period, it is requested that the petitioner

promptly so inform the Commission by

a toll-free telephone call to Western

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

1-(800) 342-6700). The Western Union

operator should be given Datagram

Identification Number N1023 and the

following message addressed to (Project

Director): petitioner’s name and

telephone number, date petition was

mailed, plant name, and publication

date and page number of this Federal

Register notice. A copy of the petition

should also be sent to the Office of the

General Counsel, U.S. Nuclear

Regulatory Commission, Washington,

DC 20555, 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.

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 amendments request:

November 1, 1995, as supplemented on

December 1, 1995

Description of amendments request:

The proposed amendments would

revise the Calvert Cliffs Nuclear Power

Plant, Unit Nos. 2 and 3, Technical

Specifications (TSs) and supporting TS

Bases relating to the electrical

distribution system. The changes are

necessary to accommodate the

installation of a new safety-related

emergency diesel generator (EDG) and a

non-safety EDG. The non-safety EDG

will be used as an alternate air

conditioning source of power in case of

a station blackout. In addition to

reflecting the new plant configuration,

the proposed TSs also reflect the

upgraded electrical capacities of the

existing EDGs, increased fuel oil storage,

and fire protection system for the new

EDG building.

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. Would not involve a significant increase

in the probability or consequences of an

accident previously evaluated.

The Engineered Safety Features (ESF)

electrical system provides a reliable source of

electrical power to the 4.16 kV ESF busses to

operate the necessary accident mitigation

equipment, should offsite power be lost. The

proposed change to Units 1 and 2 Technical

Specifications was prompted by two

significant modifications to this system - the

addition of No. 1A Emergency Diesel

Generator (EDG) and the upgrade of the

electrical capacity of two of the three existing

Fairbanks Morse EDGs. The addition of No.

1A EDG provides the plant with an ESF

electrical system configuration consisting of

two EDGs dedicated to each unit, thereby

eliminating reliance upon a ‘‘swing’’ diesel

capable of being aligned to either unit. The

four-EDG configuration provides a greater

degree of flexibility when an EDG is being

overhauled or tested during refueling

outages. The increased electrical capacity of

the existing Fairbanks Morse EDGs will give

the operators greater flexibility in the choice

of discretionary loads for the mitigation of

accidents. Both modifications necessitate

changes to the Technical Specifications.

The ESF electrical system, including the

four EDGs, is used to mitigate the

consequences of an accident. The design of

the new No. 1A EDG is such that

incorporation of this EDG into the existing

ESF electrical system does not result in this

system becoming an accident initiator.

Furthermore, the modification to upgrade the

capacity of the existing EDGs will enhance

the plant operators’ ability to mitigate

accidents by allowing greater flexibility in

the choice of discretionary loads, but will not

change the configuration of the ESF electrical

system or any support systems such that the

EDGs would become an accident initiator.

Therefore, the proposed change would not

increase the probability of an accident

previously evaluated.

The addition of the safety-related No. 1A

EDG to the ESF electrical system will

enhance the ability to provide reliable

electric power during all modes of operation

and shutdown conditions of the plant.

Number 1A EDG and its support systems are

designed such that failure of a single

component will not prevent the capability to

safely shut down the plant and to maintain

the plant in a safe shutdown condition.

Furthermore, non-safety-related systems

associated with No. 1A EDG are designed so

that their failure will not result in the loss of

function of any safety-related system. The

design of the Fire Protection System in the

Diesel Generator Building meets the Codes

and Standards specified in the mechanical

and instrumentation and controls design

reports, previously approved by the
176 Federal Register / Vol. 61, No. 2 / Wednesday, January 3, 1996 / Notices

Commission. Inclusion of components from

these systems into the Technical

Specifications is consistent with Calvert

Cliff’s current licensing basis. The proposed

Technical Specifications will demonstrate

the reliability and capability of No. 1A EDG

and the upgraded Fairbanks Morse EDGs to

perform their accident mitigation function.

Implementation of the proposed Technical

Specifications will not reduce the ability of

the EDGs to perform their safety functions.

The increased volume of fuel oil necessary to

support operation of No. 1A EDG and the

upgraded Fairbanks Morse EDGs will not

adversely impact the ability of any systems

to perform their safety functions. The

auxiliary systems which required

modification or analysis to support the

upgraded ratings of the Fairbanks Morse

EDGs will not adversely impact operation of

any other plant systems necessary to mitigate

the consequences of an accident. Based on

the above, the proposed change would not

increase the consequences of an accident

previously evaluated.

Therefore, the proposed change does not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

2. Would not create the possibility of a new

or different type of accident from any

accident previously evaluated.

The proposed change adds Surveillance

Requirements, Limiting Conditions for

Operation, and Action Statements to reflect

the addition of a new EDG to the ESF

electrical system, and upgrades the electrical

capacity of the existing Fairbanks Morse

EDGs. This change does not add any new

equipment, modify any interfaces with any

existing equipment, or change the

equipment’s function, or the method of

operating the equipment to be modified. The

system will continue to operate in the same

manner as before the capacity upgrades were

implemented. The additional fuel oil

required to support the capacity upgrades

will be stored in the existing Seismic

Category I fuel oil storage tanks. The

modified EDGs will continue to serve a

function as accident mitigators, and will not

become an initiator of any accident.

The NRC has reviewed the design of the

new EDG, its attendant support systems and

the new EDG Building, and concurs with

Baltimore Gas and Electric Company’s

determination that the design satisfies the

design requirements for a safety-related EDG.

Number 1A EDG is a tandem engine-single

generator set, and is physically very different

from the existing single engine-generator

Fairbanks Morse EDGs. However, the 4.16 kV

three-phase rated electrical output is the

same as that provided by the Fairbanks

Morse EDGs to the other ESF busses. The

excess capacity of No. 1A EDG will allow the

operators greater flexibility in choosing postaccident discretionary loads, but will not

cause any detrimental effects to the ESF

busses or the equipment served by those

busses. Operation of No. 1A EDG in

accordance with these proposed Technical

Specifications will not jeopardize the

operation of any other plant systems. The

design of the Fire Protection System in the

Diesel Generator Building meets the Codes

and Standards specified in the mechanical,

and instrumentation and controls design

reports, previously approved by the

Commission. Inclusion of components from

these systems into the Technical

Specifications is consistent with Calvert

Cliffs current licensing basis. Furthermore,

locating No. 1A EDG and its fuel oil supply

in a separate Category I building provides

additional assurance that this equipment will

not become an initiator of any accident.

Therefore, the proposed change does not

create the possibility of a new or different

type of accident from any accident

previously evaluated.

3. Would not involve a significant

reduction in a margin of safety.

The safety function of the EDGs and the

ESF electrical system is to provide a reliable

source of electrical power to the safetyrelated busses to operate the necessary

accident mitigation equipment, should offsite

power be lost. The margin of safety

associated with this safety function is twofold: (1) a level of redundancy must be

designed into the EDGs and the ESF

electrical system such that the single failure

criteria is met; and (2) the power supplied to

the ESF electrical system by the EDGs must

be sufficient to power the necessary accident

mitigation equipment, should offsite power

be lost.

The addition of No. 1A EDG provides the

plant with an ESF electrical system

configuration consisting of two EDGs

dedicated to each unit, thereby eliminating

reliance upon a swing diesel capable of being

aligned to either unit. In the current

configuration, the facility meets the single

failure criteria on a ‘‘per site’’ basis.

However, as a result of the new four-EDG

configuration, each unit will have redundant

diesel generators to supply power to

redundant safety-related equipment required

for safe shutdown or accident mitigation. The

revised Fuel Oil System configuration and

the minimum fuel oil volume to be

maintained in the fuel oil tanks supports the

safety function of the EDGs, while

maintaining the margin of safety associated

with this equipment. Altogether, the new

four-EDG configuration may be considered an

increase in the margin of safety.

Inclusion of Surveillances for the Fire

Protection System components into the

Technical Specifications is consistent with

Calvert Cliffs current licensing basis, and

ensures that adequate fire detection and

suppression capability is available to identify

and extinguish fires in the Diesel Generator

Building, thereby reducing the potential for

damage to No. 1A EDG and its auxiliaries.

The Diesel Generator Building and its Fire

Protection System is designed so that smoke

and heat from a fire in that building will not

impact the redundant safety-related

Emergency Diesel Generator in the Auxiliary

Building.

At the completion of the modifications to

increase the capacities of the Unit 2 EDGs

and to install the new No. 1A EDG, we will

have diesel generators with more available

margin than currently exists. This will

provide the operators with more flexibility

during conditions where the diesel

generators are providing onsite power. The

higher electrical capacities will result in an

increase in the margin between the EDGs’

electrical capacities and the electrical power

required to operate safety-related equipment

required for safe shutdown or accident

mitigation. Therefore, these modifications

may be considered an increase in the margin

of safety.

Therefore, the proposed change does not

involve a significant reduction in a margin of

safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendments request

involves no significant hazards

consideration.

Local Public Document Room

location: Calvert County Library, Prince

Frederick, Maryland 20678.

Attorney for licensee: Jay E. Silbert,

Esquire, Shaw, Pittman, Potts and

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037.

NRC Project Director: Ledyard B.

Marsh

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 amendments request:

November 30, 1995

Description of amendments request:

The proposed amendments would

revise the Calvert Cliffs Nuclear Power

Plant, Unit Nos. 1 and 2, Technical

Specifications (TSs) to allow the

installation of tube sleeves as an

alternative to plugging for repairing

steam generator (SG) tubes. The

proposed changes to TS 3/4.4.5, ‘‘Steam

Generators,’’ and their supporting Bases

would permit tube sleeving repair

techniques developed by Westinghouse

Electric Corporation and ABB

Combustion Engineering, Inc., to be

used as a repair method for the SGs at

the Calvert Cliffs site.

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 would not

involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The implementation of the proposed steam

generator tube sleeving has been reviewed for

impact on the current CCNPP [Calvert Cliffs

Nuclear Power Plant] licensing basis.

Since the sleeve dimensions, materials,

and connecting joints to the existing tube are

designed to the applicable American Society

of Mechanical Engineers (ASME) Boiler and

Pressure Vessel Code, the proposed sleeving
Federal Register / Vol. 61, No. 2 / Wednesday, January 3, 1996 / Notices 177

repair acts as an in-kind substitution for the

original steam generator tubing. The

applicable design criteria for the sleeves

conform to the stress limits and margins of

safety of Section III of the ASME Code. Safety

factors of 3 for normal operation and 1.5 for

accident conditions were applied to the

design. Mechanical testing using the ASME

Code stress allowables has been performed in

support of the design. Based on the results

of Westinghouse and ABB-Combustion

Engineering analytical and test programs, the

sleeves fulfill their intended function as leak

tight structural members and meet or exceed

all design criteria.

Evaluation of the proposed sleeved tubes

indicates no detrimental effects on the sleeve

or sleeve-tube assembly from reactor system

flow, primary or secondary coolant

chemistries, thermal conditions or transients,

or pressure conditions or transients as may

be experienced at CCNPP. Corrosion testing

of sleeve-tube assemblies indicate no

evidence of sleeve or tube corrosion

considered detrimental under anticipated

service conditions.

The installation of the proposed sleeves is

controlled via the sleeving vendors’

proprietary processes and equipment. The

ABB Combustion Engineering process has

been in use since 1984, and has been

implemented 24 times for the installation of

over 4,200 sleeves. The Westinghouse

process has been in use since 1988, and

approximately 12,000 laser welded sleeves

have been installed between 1988 and 1994.

The CCNPP steam generator design was

reviewed and found to be compatible with

both installation processes and equipment.

The implementation of the proposed

sleeves has no significant effect on either the

configuration of the plant, or the manner in

which it is operated. The hypothetical

consequences of failure of the sleeved tube is

bounded by the current steam generator tube

rupture analysis described in Section 14.15

of the Calvert Cliffs Updated Final Safety

Analysis Report.

Therefore, BGE [Baltimore Gas and

Electric] has concluded that the proposed

change does not involve a significant

increase in the probability or consequences

of an accident previously evaluated.

(2) [The proposed amendment] would not

create the possibility of a new or different

kind of accident from any other accident

previously evaluated.

As discussed above, the structural

integrity, thermal characteristics, and

material properties of the proposed sleeves

are consistent with the existing plant steam

generators. Therefore, the functions of the

steam generators will not be significantly

affected by the installation of the proposed

sleeves. In addition, the proposed sleeves do

not interact with any other plant systems.

The continued integrity of the installed

sleeve is periodically verified by the

Technical Specification requirements. The

implementation of the proposed sleeves has

no significant effect on either the

configuration of the plant, or the manner in

which it is operated.

Therefore, BGE concludes that this

proposed change does not create the

possibility of a new or different kind of

accident from any previously evaluated.

(3) [The proposed amendment] would not

involve a significant reduction in a margin of

safety.

The repair of degraded steam generator

tubes via the use of the proposed sleeves has

been confirmed to restore the structural

integrity of the faulted tube under normal

operating and postulated accident

conditions. The design safety factors utilized

for the sleeves are consistent with the safety

factors in the ASME Boiler and Pressure

Vessel Code used in the original steam

generator design. The repair limit for the

proposed sleeves is consistent with that

established for the steam generator tubes. The

design of the sleeve to tube joints is verified

by testing to preclude significant leakage

during normal and postulated accident

conditions. Use of the previously identified

design criteria and design verification testing

assures that the margin to safety with respect

to the implementation of the proposed

sleeves is not significantly different from the

original steam generator tubes.

Therefore, BGE concludes that the

proposed change does not involve a

significant reduction in a margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendments request

involves no significant hazards

consideration.

Local Public Document Room

location: Calvert County Library, Prince

Frederick, Maryland 20678.

Attorney for licensee: Jay E. Silbert,

Esquire, Shaw, Pittman, Potts and

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037.

NRC Project Director: Ledyard B.

Marsh

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 amendments request:

December 7, 1995

Description of amendments request:

The proposed amendments would

change the Calvert Cliffs Nuclear Power

Plant, Unit Nos. 1 and 2, Technical

Specifications (TSs) by adding an

analysis technique to the list of

approved core operating limits

analytical methods. Specifically, these

amendments would add the convolution

analysis technique to the list of

approved methodologies in TSs

6.9.1.9.b. The convolution analysis

technique has already been reviewed

and approved by the NRC staff and the

supporting safety evaluation was

provided to the licensee by an NRC

letter dated May 11, 1995.

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 change has been evaluated against the

standards in 10 CFR 50.92 and has been

determined to not involve a significant

hazards consideration in that operation of the

facility in accordance with the proposed

amendment:

1. Would not involve a significant increase

in the probability or consequences of an

accident previously evaluated.

The proposed change is to add the

convolution analysis technique previously

approved by the NRC to the list of approved

methodologies in Calvert Cliffs’ Unit 1 and 2

Technical Specifications. By letter dated

November 1, 1994, Baltimore Gas and

Electric Company (BGE) requested approval

to use the ABB/Combustion Engineering

(ABB/CE) convolution technique for

determining the values in the Calvert Cliffs

Core Operating Limits Report (COLR) related

to the pre-trip main steam line break event.

Approval was given by the NRC in their letter

dated May 11, 1995. The addition of this

technique to the list of approved analytical

methods in Technical Specification 6.9.1.9.b

is simply intended to identify it as an

approved methodology. Therefore, the

change does not involve a significant

increase in the probability or consequences

of an accident previously evaluated.

2. Would not create the possibility of a new

or different type of accident from any

accident previously evaluated.

The proposed change is to add the

convolution analysis technique previously

approved by the NRC to the list of approved

methodologies in Calvert Cliffs’ Unit 1 and 2

Technical Specifications. By letter dated

November 1, 1994, BGE requested approval

to use the ABB/CE convolution technique for

determining the values in the Calvert Cliffs

COLR related to the pre-trip main steam line

break event. Approval was given by the NRC

in their letter dated May 11, 1995. The

addition of this technique to the list of

approved analytical methods in Technical

Specifications 6.9.1.9.b is simply intended to

identify it as an approved methodology.

Therefore, the change would not create the

possibility of a new or different type of

accident from any accident previously

evaluated.

3. Would not involve a significant

reduction in the margin of safety.

The proposed change is to add the

convolution analysis technique previously

approved by the NRC to the list of approved

methodologies in Calvert Cliffs’ Unit 1 and 2

Technical Specifications. By letter dated

November 1, 1994, BGE requested approval

to use the ABB/CE convolution technique for

determining the values in the Calvert Cliffs

COLR related to the pre-trip main steam line

break event. Approval was given by the NRC

in their letter dated May 11, 1995. The

addition of this technique to the list of

approved analytical methods in Technical

Specification 6.9.1.9.b is simply intended to

identify it as an approved methodology.

Therefore, operation of the facility in

accordance with the proposed amendment
178 Federal Register / Vol. 61, No. 2 / Wednesday, January 3, 1996 / Notices

does not involve a significant reduction in a

margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 50.92(c) are satisfied.

Therefore, the NRC staff proposes to

determine that the amendments request

involves no significant hazards

consideration.

Local Public Document Room

location: Calvert County Library, Prince

Frederick, Maryland 20678.

Attorney for licensee: Jay E. Silbert,

Esquire, Shaw, Pittman, Potts and

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037.

NRC Project Director: Ledyard B.

Marsh

Duquesne Light Company, et al., Docket

No. 50-334, Beaver Valley Power

Station, Unit No. 1, Shippingport,

Pennsylvania

Date of amendment request:

December 7, 1995

Description of amendment request:

The proposed amendment would revise

Technical Specifications (TSs) 3.4.5 and

3.4.6.2 and their Bases to maintain

voltage-based steam generator tube

repair criteria for the tube support plate

elevations beyond the current cycle of

operation. The proposed amendment

would implement a 2.0 volt repair limit

to replace a 1.0 volt repair limit which

was approved on an interim basis for

only the current fuel cycle by License

Amendment No. 184 [issued February 3,

1995]. The proposed amendment would

also include changes in addition to

those incorporated by License

Amendment No. 184 to reflect the

guidance provided in NRC Generic

Letter (GL) 95-05, ‘‘Voltage-Based Repair

Criteria for Westinghouse Steam

Generator Tubes Affected by Outside

Diameter Stress Corrosion Cracking.’’

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Does the change involve a significant

increase in the probability or consequences

of an accident previously evaluated?

Tube burst criteria are inherently satisfied

during normal operating conditions due to

the proximity of the tube support plate (TSP).

Test data indicates that tube burst cannot

occur within the TSP, even for tubes which

have 100% throughwall electric discharge

machining notches, 0.75 inch long, provided

that the TSP is adjacent to the notched area.

Since tube-to-TSP proximity precludes tube

burst during normal operating conditions,

use of the criteria must retain tube integrity

characteristics which maintain a margin of

safety of 1.43 times the bounding faulted

condition, main steamline break (MSLB)

pressure differential. As previously stated,

the Regulatory Guide (RG) 1.121 criterion

requiring maintenance of a safety factor of

1.43 times the MSLB pressure differential on

tube burst is satisfied by 7/8’’ diameter

tubing with bobbin coil indications with

signal amplitudes less than 8.82 volts,

regardless of the indicated depth

measurement.

The upper voltage repair limit (Vurl) will be

determined prior to each outage using the

most recently approved NRC database to

determine the tube structural limit (Vsl). The

structural limit is reduced by allowances for

nondestructive examination (NDE)

uncertainty (Vnde) and growth (Vgr) to

establish Vurl. Using Generic Letter (GL) 95-

05 and growth allowances for an example,

the NDE uncertainty component of 20% and

a voltage growth allowance of 30% per full

power year can be utilized to establish a Vurl

of 5.9 volts. The 20% NDE uncertainty

represents a square-root-sum-of-the-squares

(SRSS) combination of probe wear

uncertainty and analyst variability. The

degradation growth allowance should be an

average growth rate or 30% per effective full

power year, whichever is larger. This growth

allowance is conservative for BVPS-1 [Beaver

Valley Power Station, Unit No. 1] as the

percent voltage growth rates have decreased

for each of the last three inspections.

Relative to the expected leakage during

accident condition loadings, it has been

previously established that a postulated

MSLB outside of containment but upstream

of the main steam isolation valve (MSIV)

represents the most limiting radiological

condition relative to the plugging criteria. In

support of implementation of the revised

plugging limit, analyses will be performed to

determine whether the distribution of

cracking indications at the tube support plate

intersections during future cycles are

projected to be such that primary-tosecondary leakage would result in postulated

site boundary and control room doses

exceeding 10 CFR 100, and 10 CFR 50,

Appendix A, GDC-19 requirements,

respectively. A separate calculation has

determined the maximum allowable MSLB

leakage limit in a faulted loop. This limit was

calculated using the technical specification

reactor coolant system (RCS) Iodine-131

activity level of 1.0 microcuries per gram

dose equivalent Iodine-131 and the

recommended Iodine-131 transient spiking

values consistent with NUREG-0800. The

projected MSLB leakage rate calculation

methodology prescribed in Section 2.b of GL

95-05 will be used to calculate the end-ofcycle (EOC) leakage. Projected EOC voltage

distribution will be developed using the most

recent EOC eddy current results and

considering an appropriate voltage

measurement uncertainty. The log-logistic

probability of leakage correlation will be

used to establish the MSLB leakrate used for

comparison with the faulted loop allowable

limit. Due to the relatively low voltage levels

of indications at BVPS-1 and low voltage

growth rates, it is expected that the

calculated leakage values will not exceed this

limit. Therefore, as implementation of the 2.0

volt voltage-based plugging criteria at BVPS1 does not adversely affect steam generator

tube integrity and implementation will be

shown to result in acceptable dose

consequences, the proposed amendment does

not result in any increase in the probability

or consequences of an accident previously

evaluated in the UFSAR [Updated Final

Safety Analysis Report].

2. Does the change create the possibility of

a new or different kind of accident from any

accident previously evaluated?

Implementation of the proposed steam

generator tube 2.0 volt plugging limit does

not introduce any significant changes to the

plant design basis. Use of the 2.0 volt

plugging limit does not provide a mechanism

which could result in an accident outside of

the region of the tube support plate

elevations as no outside diameter stress

corrosion cracking (ODSCC) is occurring

outside the thickness of the tube support

plates. Neither a single or multiple tube

rupture event would be expected in a steam

generator in which the plugging limit has

been applied (during all plant conditions).

Duquesne Light Company will continue to

implement a maximum primary-to-secondary

leakage rate limit of 150 gpd [gallons per day]

per steam generator to help preclude the

potential for excessive leakage during all

plant conditions. The RG 1.121 criterion for

establishing operational leakage rate limits

that require plant shutdown are based upon

leak-before-break considerations to detect a

free span crack before potential tube rupture

during faulted plant conditions. The 150 gpd

limit provides for leakage detection and plant

shutdown in the event of the occurrence of

an unexpected single crack resulting in

leakage that is associated with the longest

permissible crack length. RG 1.121

acceptance criteria for establishing operating

leakage limits are based on leak-before-break

considerations such that plant shutdown is

initiated if the leakage associated with the

longest permissible crack is exceeded.

The single through-wall crack lengths that

result in tube burst at 1.43 times the MSLB

pressure differential and the MSLB pressure

differential alone are approximately 0.57

inch and 0.84 inch, respectively. A leak rate

of 150 gpd will provide for detection of 0.41

inch long cracks at nominal leak rates and

0.62 inch long cracks at the lower 95%

confidence level leak rates. Since tube burst

is precluded during normal operation due to

the proximity of the TSP to the tube and the

potential exists for the crevice to become

uncovered during MSLB conditions, the

leakage from the maximum permissible crack

must preclude tube burst at MSLB

conditions. Thus, the 150 gpd limit provides

for plant shutdown prior to reaching critical

crack lengths for MSLB conditions using the

lower 95% leakrate data. Additionally, this

leak-before-break evaluation assumes that the

entire crevice area is uncovered during

blowdown. Partial uncovery will provide

benefit to the burst capacity of the

intersection. Analyses have shown that only

a small percentage of the TSPs are deflected

greater than the TSP thickness during a

postulated MSLB.

As steam generator tube integrity upon

implementation of the 2.0 volt plugging limit
Federal Register / Vol. 61, No. 2 / Wednesday, January 3, 1996 / Notices 179

continues to be maintained through inservice

inspection and primary-to-secondary leakage

monitoring, the possibility of a new or

different kind of accident from any accident

previously evaluated is not created.

3. Does the change involve a significant

reduction in a margin of safety?

The use of the voltage-based bobbin probe

tube support plate elevation plugging criteria

at BVPS-1 maintains steam generator tube

integrity commensurate with the criteria of

RG 1.121. This guide describes a method

acceptable to the Commission for meeting

GDCs [General Design Criterion] 14, 15, 30,

31, and 32 by reducing the probability or the

consequences of steam generator tube

rupture. This is accomplished by

determining the limiting conditions of

degradation of steam generator tubing, as

established by inservice inspection, for

which tubes with unacceptable cracking

should be removed from service. Upon

implementation of the proposed criteria,

even under the worst case conditions, the

occurrence of ODSCC [Outside Diameter

Stress Corrosion Cracking] at the tube

support plate elevations is not expected to

lead to a steam generator tuberupture event

during normal or faulted plant conditions.

The EOC distribution of crack indications at

the tube support plate elevations will be

confirmed to result in acceptable primary-tosecondary leakage during all plant conditions

and that radiological consequences are not

adversely impacted.

In addressing the combined effects of lossof-coolant-accident (LOCA) + safe shutdown

earthquake (SEE) on the steam generator

component (as required by GDC 2), it has

been determined that tube collapse may

occur in the steam generators at some plants.

This is the case as the tube support plates

may become deformed as a result of lateral

loads at the wedge supports at the periphery

of the plate due to the combined effects of

the LOCA rarefaction wave and SSE loadings.

Then, the resulting pressure differential on

the deformed tubes may cause some of the

tubes to collapse. There are two issues

associated with steam generator tube

collapse. First, the collapse of steam

generator tubing reduces the RCS [reactor

coolant system] flow area through the tubes.

The reduction in flow area increases the

resistance to flow of steam from the core

during a LOCA which, in turn, may

potentially increase peak clad temperature.

Second, there is a potential that partial

through-wall cracks in tubes could progress

to complete through-wall cracks during tube

deformation or collapse.

The results of an analysis using the larger

break inputs show that the LOCA loads were

found to be of insufficient magnitude to

result in steam generator tube collapse or

significant deformation. Since the leakbefore-break methodology is applicable to

BVPS-1 reactor coolant loop piping, the

probability of breaks in the primary loop

piping is sufficiently low that they need not

be considered in the structural design of the

plant. The limiting LOCA event becomes

either the accumulator line break or the

pressurizer surge line break. Analysis results

provided in WCAP-14122, dated July 1994,

demonstrate that no tubes were subject to

deformation or collapse. No tubes have been

excluded from application of the subject

voltage-based steam generator plugging

criteria.

Addressing RG 1.83 considerations,

implementation of the bobbin probe voltagebased tube plugging criteria of 2.0 volts is

supplemented by: enhanced eddy current

inspection guidelines to provide consistency

in voltage normalization, a 100% eddy

current inspection sample size at the tube

support plate elevations, and rotating

pancake coil inspection requirements for the

larger indications left inservice to

characterize the principal degradation as

ODSCC.

As noted previously, implementation of

the tube support plate intersection voltagebased plugging criteria will decrease the

number of tubes which must be repaired. The

installation of steam generator tube plugs

reduces the RCS flow margin. Thus, the

implementation of the 2.0 volt plugging limit

will maintain the margin of flow that would

otherwise be reduced in the event of

increased tube plugging.

Based on the above, it is concluded that the

proposed license amendment request does

not result in a significant reduction in margin

with respect to plant safety as defined in the

UFSAR or any BASES of the plant technical

specifications.

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: B. F. Jones Memorial Library,

663 Franklin Avenue, Aliquippa, PA

15001

Attorney for licensee: Jay E. Silberg,

Esquire, Shaw, Pittman, Potts &

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037

NRC Project Director: John F. Stolz

Duquesne Light Company, et al., Docket

Nos. 50-334 and 50-412, Beaver Valley

Power Station, Unit Nos. 1 and 2,

Shippingport, Pennsylvania

Date of amendment request:

December 15, 1995

Description of amendment request:

The proposed amendments would (1)

revise Technical Specifications (TSs) 3/

4.6.1.1, 3/4.6.1.2, 3/4.6.1.3, 3/4.6.1.6,

and associated Bases, (2) delete TS

6.9.2.g, and (3) add a new TS 6.17. The

proposed changes would make the TSs

consistent with Option B of recently

revised Appendix J of 10 CFR Part 50

and the implementing guidance of

Regulatory Guide 1.163, ‘‘PerformanceBased Containment Leak Test Program,’’

dated September 1995. Option B of

Appendix J permits licensees to

implement a performance based option

rather than the previous prescriptive

requirements now contained in

Appendix J as Option A. The proposed

amendments would remove from the

TSs the prescriptive requirements of

Option A concerning test frequencies

and test methodology and would also

include minor administrative and

editorial changes to add consistency

between the Bases and the TSs and to

provide additional clarification.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Does the change involve a significant

increase in the probability or consequences

of an accident previously evaluated?

Containment leakage is not an accident

initiator. The proposed amendment does not

add or modify any existing plant equipment.

Therefore there is no increase in the

probability of an accident previously

evaluated.

The consequences of an accident

previously evaluated are not significantly

increased. The proposed changes do not

affect the assumptions, parameters or result

of any Updated Final Safety Analysis

(UFSAR) accident analyses. The containment

leakage rate will continue to be maintained

within the limit assumed in the accident

analysis for a Design Basis Accident (DBA).

The proposed changes do not modify the

response of the containment during a DBA.

The proposed amendment will continue to

ensure that the ability of the containment

structure, including the containment air

locks, to limit leakage from a DBA is

demonstrated using test methodologies and

guidance on test frequencies that have been

determined to be acceptable to meet the

requirements of 10 CFR 50, Appendix J,

Option B.

The potential increase to overall accident

risk due to the containment leak tightness

decreasing between extended testing

intervals and the resulting potential

increased radioactivity release to the

environment during a DBA has been

determined to be minimal based on the

findings of NUREG 1493 titled ‘‘PerformanceBased Containment Leak-Test Program.’’ In

addition, due to the performance based

nature of 10 CFR 50 Appendix J, Option B,

the extended test intervals are utilized only

when the component(s) have demonstrated

an acceptable performance history.

Therefore, a significant decrease in

containment leak tightness between extended

test intervals is not expected as a result of

this proposed change.

Based on the above discussion, it is

concluded that this change does not involve

a significant increase in the probability or

consequences of an accident previously

evaluated.

2. Does the change create the possibility of

a new or different kind of accident from any

accident previously evaluated?

The proposed change does not involve any

physical changes to the plant or changes in
180 Federal Register / Vol. 61, No. 2 / Wednesday, January 3, 1996 / Notices

plant operating configuration. The proposed

amendment involves changes to plant

programs and administrative requirements

used in determining acceptable containment

performance. The performance of plant

systems, including the containment

structure, during plant operation remains

unchanged.

Therefore, the proposed change does not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

3. Does the change involve a significant

reduction in a margin of safety?

The margin of safety is not significantly

reduced by this proposed change. The

acceptance criteria for ‘‘as left’’ measured

containment leakage rates is not being

increased as result of this proposed

amendment. For Beaver Valley Power Station

(BVPS) Unit No. 1 only, the ‘‘as found’’

maximum allowable overall Type A leakage

rate is being slightly increased. However, the

slight increase does not exceed the value

assumed in accident analysis for containment

leakage during a DBA due to changing the

acceptance criteria from less than to less than

or equal to. The margin between the

acceptable ‘‘as left’’ measured overall Type A

containment leakage rate and the leakage rate

assumed in the accident analysis is not being

decreased.

The maximum ‘‘as found’’ allowable

overall Type A leakage rate remains

unchanged for BVPS Unit No. 2. The margin

between the acceptable ‘‘as left’’ measured

overall Type A containment leakage rate and

the leakage rate assumed in the accident

analysis is also not being decreased.

The maximum allowable measured

combined Type B and C leakage rate is not

being increased above the current limits.

The maximum peak containment pressure

following a DBA remains unchanged. The

containment depressurization time following

a DBA remains unchanged. The calculated

offsite dose consequences of a DBA remains

unchanged.

The proposed amendment continues to

ensure reactor containment system reliability

by periodic testing in compliance with 10

CFR 50, Appendix J, Option B. The extension

of Type A, B and C test frequencies permitted

by 10 CFR 50 Appendix J, Option B, is not

expected to result in a significant decrease in

containment leak tightness between test

intervals. Due to the performance based

nature of 10 CFR 50 Appendix J, Option B,

the extended test intervals are utilized only

when the component(s) have demonstrated

an acceptable performance history.

Therefore, a significant decrease in

containment leak tightness between extended

test intervals is not expected as a result of

this proposed change.

The changes which are either

administrative or editorial in nature will not

reduce the margin of safety because they

have no impact on any safety analysis

assumptions.

Therefore, based on the above discussion,

it can be concluded that the proposed change

does not involve a significant reduction in a

margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: B. F. Jones Memorial Library,

663 Franklin Avenue, Aliquippa,

Pennsylvania 15001.

Attorney for licensee: Jay E. Silberg,

Esquire, Shaw, Pittman, Potts &

Trowbridge, 2300 N Street, NW.,

Washington, DC 20037

NRC Project Director: John F. Stolz

Entergy Operations Inc., Docket No. 50-

382, Waterford Steam Electric Station,

Unit 3, St. Charles Parish, Louisiana

Date of amendment request: May 19,

1995, as supplemented by letter dated

December 7, 1995.

Description of amendment request:

May 19, 1995, submittal requested to

modify Action Statement for Technical

Specification (TS) 3.6.4.2 for the

hydrogen recombiners. It also requested

to make the surveillance requirements

for hydrogen recombiners consistent

with NUREG-1432, ‘‘Standard Technical

Specifications Combustion Engineering

Plants.’’ The December 7, 1995, letter

withdrew the request to change the

Action Statement for TS 3.6.4.2.

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:

The response is predicated on the

following technical bases: (1) the current

licensing basis of record establishes that only

one recombiner system is required to

maintain hydrogen concentration below 4%,

(2) the proposed technical specification

changes are conservative when compared

with the recommendations of Regulatory

Guide 1.7, (3) short term post LOCA

hydrogen generation is less than 1%, (4) long

term post LOCA hydrogen generation is less

than the flame propagation limit, which

according to Regulatory Guide 1.7 would not

result in adverse effects to containment

systems, and (5) a design basis LOCA without

long term hydrogen control would produce

pressures below the containment design

pressure.... Therefore, the proposed change

will not involve a significant increase in the

probability or consequences of any accident

previously evaluated.

The proposed change will not alter the

configuration or operation of any other plant

system or component. The change does not

involve any change to the operational design

or limits of any other plant systems or

components. Thus, no new failure modes are

introduced or associated with the proposed

change. Therefore, the proposed change will

not create the possibility of a new or different

kind of accident from any accident

previously evaluated.

The proposed change will have no adverse

impact on the protective boundaries, safety

limits, or margin or safety. There are no

limits or margins of safety being revised for

any systems, components, or protective

boundaries.

Therefore, the proposed change will not

involve a significant reduction in a margin of

safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 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 New Orleans

Library, Louisiana Collection, Lakefront,

New Orleans, LA 70122

Attorney for licensee: N.S. Reynolds,

Esq., Winston & Strawn 1400 L Street

N.W., Washington, D.C. 20005-3502

NRC Project Director: William D.

Beckner

Entergy Operations Inc., Docket No. 50-

382, Waterford Steam Electric Station,

Unit 3, St. Charles Parish, Louisiana

Date of amendment request:

November 7, 1995

Description of amendment request:

Amendment to Technical Specification

(TS) 3/4.8.1 ‘‘Electrical Power Systems -

AC Sources’’ and the associated TS

BASES. The proposed amendment

would implement selected changes from

NUREG 1432, ‘‘Standard Technical

Specifications Combustion Engineering

Plants,’’ Generic Letter (GL) 94-01,

‘‘Removal of Accelerated Testing and

Special Reporting Requirements for

Emergency Diesel Generators,’’ and GL

93-05, ‘‘Line-Item Technical

Specifications Improvements to Reduce

Surveillance Requirements for Testing

During Power Operation.’’ The intent of

these changes is to increase Emergency

Diesel Generator (EDG) reliability by

reducing the stresses on the EDGs

caused by unnecessary testing. This

proposed TS amendment will also

relocate the Surveillance Requirements

for maintaining the properties of the

fuel oil to TS Section 6, ‘‘Administrative

Controls.’’ These requirements will be

implemented as part of the Fuel Oil

Testing Program. In addition, the

requirement for cleaning the diesel fuel

oil storage tanks with a sodium

hypochlorite solution or equivalent will

be changed to also allow an appropriate

mechanical method (such as pressure

washing or manual wiping) to be

utilized.

Basis for proposed no significant

hazards consideration determination:
Federal Register / Vol. 61, No. 2 / Wednesday, January 3, 1996 / Notices 181

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 Standby Diesel Generators do not

initiate any accidents, therefore the proposed

changes do not increase the probability of an

accident previously evaluated. The proposed

changes to TS 3/4.8.1 and the associated

BASES affect the required actions in

response to inoperable offsite and onsite AC

sources, Surveillance Requirements for the

EDG, and reporting requirements for EDG

failures. The majority of the proposed

changes are based on the recommendations

of NUREG 1432, GL 94-01, and GL 93-05.

These proposed changes have been

extensively reviewed by the NRC during the

preparation of these documents and by

Waterford 3 SES during the development of

this request for TS amendment. The

proposed changes are expected to result in

improvements in EDG performance and

reduce EDG aging due to excessive testing.

The proposed changes will permit the

elimination of the unnecessary mechanical

stress and wear on the EDGs while ensuring

that the EDGs will perform their design

function. The elimination of mechanical

stress and wear will improve reliability and

availability of the EDGs which will have a

positive effect on the ability of the EDGs to

perform their design function. The proposed

changes do not affect the availability or the

testing requirements of the offsite circuits.

Therefore, the proposed change will not

involve a significant increase in the

probability or consequences of any accident

previously evaluated.

The proposed changes to TS 3/4.8.1 and

the associated Bases do not introduce any

new modes of plant operation or new

accident precursors, involve any physical

alterations to plant configurations, or make

any changes to system setpoints which could

initiate a new or different kind of accident.

The proposed changes do not affect the

design or performance characteristics of any

EDG or its ability to perform its design

function. No new failure modes have been

defined and no new system interactions have

been introduced for any plant system or

component. In addition, there have not been

any new limiting failures identified as a

result of the proposed changes. The proposed

changes will eliminate unnecessary EDG

testing and will increase EDG reliability and

availability. This will have an overall

positive affect on plant safety. Accidents

concerning loss of offsite power and a single

failure (e.g., loss of an EDG) have previously

been evaluated. These changes are intended

to improve plant safety, decrease equipment

degradation, and remove an unnecessary

burden on personnel resources by reducing

the amount of testing that the TS requires

during power operation.

Relocating the diesel fuel oil testing

requirements to the Waterford 3 Fuel Oil

Testing Program outside of the Technical

Specifications is an administrative change

only and consequently has no effect on

accident probability, consequences, or

margin. Also, the proposed cleaning method

for the diesel fuel oil storage tanks meets the

intent of Regulatory Guide 1.137 and will not

result in the degradation of the fuel oil.

Therefore, the proposed change will not

create the possibility of a new or different

kind of accident from any accident

previously evaluated.

Under the proposed changes to TS 3/4.8.1

and the associated Bases, the EDGs will

remain capable of performing their safety

function. The changes do not affect the

design or performance of the EDGs, but will

increase EDG reliability and availability by

reducing the stresses and the effects of aging

on the EDG by eliminating unnecessary

testing. This will result in an overall increase

in plant safety. The ability of the EDGs to

perform their safety function will not be

degraded. Relocating the diesel fuel oil

testing requirements to the Waterford 3 Fuel

Oil Testing Program outside of the Technical

Specifications is an administrative change

only and consequently has no effect on

accident probability, consequences, or

margin. Also, the proposed cleaning method

for the diesel fuel oil storage tanks meets the

intent of Regulatory Guide 1.137 and will not

result in a reduction in the margin of safety.

Therefore, the proposed change will not

involve a significant reduction in a margin of

safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 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 New Orleans

Library, Louisiana Collection, Lakefront,

New Orleans, Louisiana 70122

Attorney for licensee: N.S. Reynolds,

Esq., Winston & Strawn 1400 L Street

N.W., Washington, D.C. 20005-3502

NRC Project Director: William D.

Beckner

Tennessee Valley Authority, Docket

Nos. 50-327 and 50-328, Sequoyah

Nuclear Plant, Units 1 and 2, Hamilton

County, Tennessee

Date of amendment request: (TS 93-

09) December 8, 1995

Description of amendment request:

The proposed change would revise the

setpoints and time delays for the

auxiliary feedwater loss-of-power and

6.9-kv shutdown board loss-of-voltage

and degraded-voltage instrumentation

setpoints in Items 6 and 7 of Technical

Specification Table 3.3-4, respectively.

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:

TVA has evaluated the proposed technical

specification (TS) change and has determined

that it does not represent a significant

hazards consideration based on criteria

established in 10 CFR 50.92(c). Operation of

Sequoyah Nuclear Plant (SQN) in accordance

with the proposed amendment will not:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed revision supports the

implementation of design logic and setpoint

changes to the loss-of-power relaying. This

relaying is designed to ensure adequate

voltage is available to safety-related loads in

order to enhance their operability and

support accident mitigation functions and to

provide for auxiliary feedwater (AFW) pump

starts. The design changes alter relay logic

and delete unnecessary relaying, but do not

change the diesel generator (D/G) start and

load-shedding actuations that result from

loss-of-power conditions. Therefore, no new

actuations or functions have been created;

and because the existing and proposed

functions provide for accident mitigation

considerations that are not the source of an

accident, the probability of an accident is not

increased. The deletion of the 6.9-kilovolt

shutdown board normal-feedwater

undervoltage relays actually reduces the

potential for inadvertent shutdown board

blackouts as a result of short-duration voltage

transients or instrument failures.

The setpoints and time delays for loss-ofpower functions have been modified based

on the guidelines developed by the Electrical

Distribution System Clearinghouse as

evaluated and determined through detailed

analysis by TVA. This design is documented

in TVA Calculations SQN-EEB-MS-T106-

0008, 27DAT, and DS-1-2 and is available for

NRC review at the SQN site. The assigned

values are conservative settings that will

ensure adequate voltage is supplied to safetyrelated loads for accident mitigation and

safety functions under normal, degraded, and

loss-of-offsite power voltage conditions with

appropriate time delays to prevent damage to

electrical loads and minimize premature or

unnecessary actuations. The identification of

loss-of-voltage conditions is enhanced by the

design changes to ensure the timely

sequencing of loads onto the D/G and the

initiation of AFW pump starts for accident

mitigation. Because there are no reductions

in safety functions resulting from the design

logic, setpoint and time-delay changes to the

loss-of-power instrumentation and offsite

dose levels for postulated accidents will not

be increased, the consequences of an

accident are not increased.

The applicable mode addition, TS 3.0.4

exclusion deletion, and response time

measurement clarification incorporated in

the proposed change do not affect plant

functions. These changes reflect the

requirements that SQN has been maintaining

and serve to clarify the requirements to

provide consistency of application and easier

understanding. The AFW footnote addition

and bases revision only clarify operability

conditions that are consistent with the plant

design for the AFW pump and loss-of-power

instrumentation. Because there are no

changes to plant functions or operations,

these revisions have no impact on accident

probabilities or consequences.
182 Federal Register / Vol. 61, No. 2 / Wednesday, January 3, 1996 / Notices

2. Create the possibility of a new or

different kind of accident from any

previously analyzed.

As described above, the loss-of-power

instrumentation ensures adequate voltage to

safety-related loads by initiating D/G starts

and load shedding and provides for AFW

pump starting, but is not considered to be the

source of an accident. Although the design

logic, setpoint, and time-delay actuation

criteria have changed, the output functions to

various plant systems that actuate for load

shedding and D/G starts remain the same.

Therefore, actuation criteria have been

affected, but not safety functions, and the

TVA evaluation has confirmed that the new

design enhances the ability to maintain

adequate voltage to support safety functions.

Since safety functions have not changed and

the new loss-of-power instrumentation

design continues to support operability of

safety-related equipment, no new or different

accident is created.

The applicable mode addition, TS 3.0.4

exclusion deletion, and response time

measurement clarification, as well as the

AFW operability clarifications, do not affect

plant functions and will not create a new

accident.

3. Involve a significant reduction in a

margin of safety.

The proposed loss-of-power TS changes

support design logic, setpoint, and timedelay requirements that have been verified by

TVA analysis to provide acceptable voltage

levels for safety-related components. In

determining the acceptability of these voltage

levels, the minimum voltage for operation as

well as detrimental component heating

resulting from sustained degraded-voltage

conditions were considered. This design

ensures that safety-related loads will be

available and operable for normal and

accident plant conditions. The applicable

mode addition, TS 3.0.4 exclusion deletion,

response time measurement clarification, and

AFW operability clarifications provide

enhancements to TS requirements and do not

affect plant functions. Therefore, no safety

functions are reduced by these changes and

there is no reduction in the margin of safety.

The NRC has reviewed the licensee’s

analysis and, based on thisreview, it

appears that the three standards of 10

CFR 50.92(c) are satisfied. Therefore, the

NRC staff proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Chattanooga-Hamilton County

Library, 1101 Broad Street, Chattanooga,

Tennessee 37402

Attorney for licensee: General

Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET llH,

Knoxville, Tennessee 37902

NRC Project Director: Frederick J.

Hebdon

Tennessee Valley Authority, Docket

Nos. 50-327 and 50-328, Sequoyah

Nuclear Plant, Units 1 and 2, Hamilton

County, Tennessee

Date of amendment request: (TS 95-

20) December 8, 1995

Description of amendment request:

The proposed change would revise

Surveillance Requirements 4.6.2.1.1.d

and 4.6.2.1.2.b to extend the

containment spray nozzle air or smoke

flow tests from the present 5-year

interval to a 10-year interval, in

accordance with Generic Letter 93-05.

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:

TVA has evaluated the proposed technical

specification (TS) change and has determined

that it does not represent a significant

hazards consideration based on criteria

established in 10 CFR 50.92(c). Operation of

Sequoyah Nuclear Plant (SQN) in accordance

with the proposed amendment will not:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The TS change is consistent with the

guidance provided in Generic Letter 93-05.

Containment spray (CS) systems’ header

piping is stainless steel; therefore, corrosion

will be negligible during the extended

surveillance interval. Since the CS systems’

headers are maintained dry, there is no

mechanism that could cause blockage of the

spray nozzles. Therefore, the nozzles in the

CS systems will remain operable, during the

10-year surveillance interval, to mitigate the

consequence of an accident previously

evaluated. Additionally, clogging or blockage

has not been observed during the 5-year

surveillance tests that have been performed

in the past at SQN. Testing the CS systems’

nozzles at the proposed reduced frequency

will not increase the probability of

occurrence of a postulated accident or the

consequences of an accident previously

evaluated.

2. Create the possibility of a new or

different kind of accident from any

previously analyzed.

The proposed reduced frequency testing of

the CS systems’ nozzles does not change the

manner in which these systems are operated.

The reduced testing frequency of the spray

nozzles does not generate any new accident

precursors. Therefore, the possibility of a

new or different kind of accident previously

evaluated is not created by the proposed

changes in surveillance frequency of the CS

system’s nozzles.

3. Involve a significant reduction in a

margin of safety.

Reduced testing of the CS systems’ nozzles

does not change the way the systems are

operated or the systems’ operability

requirements. In this application, any

additional corrosion of stainless steel piping

will be negligible during the extended

surveillance interval. Since the CS systems

are maintained dry, there is no additional

mechanism that could cause blockage of the

nozzles. Therefore, the proposed reduced

testing frequency is adequate to ensure spray

nozzle operability. The surveillance

requirements do not affect the margin of

safety since the operability requirements of

both the CS systems remains unchanged. The

existing safety analysis remains bounding.

Therefore, there is no reduction in the margin

of safety.

The NRC has reviewed the licensee’s

analysis and, based on thisreview, it

appears that the three standards of 10

CFR 50.92(c) are satisfied. Therefore, the

NRC staff proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Chattanooga-Hamilton County

Library, 1101 Broad Street, Chattanooga,

Tennessee 37402

Attorney for licensee: General

Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET 11H,

Knoxville, Tennessee 37902

NRC Project Director: Frederick J.

Hebdon

Tennessee Valley Authority, Docket

Nos. 50-327 and 50-328, Sequoyah

Nuclear Plant, Units 1 and 2, Hamilton

County, Tennessee

Date of amendment request:

December 8, 1995 (TS 95-24)

Description of amendment request:

The proposed change would modify

various Technical Specification

requirements in order to implement the

recent rule change to 10 CFR Part 50,

Appendix J. The new Appendix J rule

(Option B) provides a voluntary

performance based testing option for

containment leakage rate testing (CLRT).

Option B CLRT requirements are based

on system and component performance

in lieu of compliance with the current

prescriptive requirements. Option B

allows extension of the integrated

leakage rate test (Type A test) frequency

based on an acceptable past history. For

Type B and Type C local leak rate test,

Option B allows extension of the test

frequency based on plant-specific

experience history of each component

and establishes controls to ensure

continued performance during extended

testing intervals.

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:

TVA has evaluated the proposed technical

specification (TS) change and has determined

that it does not represent a significant

hazards consideration based on criteria
Federal Register / Vol. 61, No. 2 / Wednesday, January 3, 1996 / Notices 183

established in 10 CFR 50.92(c). Operation of

Sequoyah Nuclear Plant (SQN) in accordance

with the proposed amendment will not:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

The proposed amendment to SQN TSs is

in accordance with Option B to 10 CFR 50,

Appendix J. The proposed amendment adds

a voluntary performance based option for

containment leak rate testing. The changes

being proposed do not affect the precursor for

any accident or transient analyzed in Chapter

15 of SQN Updated Final Safety Analysis

Report. The proposed change does not

increase the total allowable primary

containment leakage rate. The proposed

change does not reflect a revision to the

physical design and/or operation of the plant.

Therefore, operation of the facility, in

accordance with the proposed change, does

not significantly affect the probability or

consequences of an accident previously

evaluated.

2. Create the possibility of a new or

different kind of accident from any

previously analyzed.

The proposed amendment to SQN TSs is

in accordance with the new performancebased option (Option B) to 10 CFR 50,

Appendix J. The changes being proposed will

not change the physical plant or the modes

of operation defined in the facility license.

The proposed changes do not increase the

total allowable primary containment leakage

rate. The changes do not involve the addition

or modification of equipment, nor do they

alter the design or operation of plant systems.

Therefore, operation of the facility in

accordance with the proposed change does

not create the possibility of a new or different

kind of accident from any previously

evaluated.

3. Involve a significant reduction in a

margin of safety.

The proposed change to SQN TSs is in

accordance with the new option to 10 CFR

50, Appendix J. The proposed option is

formulated to adopt performance-based

approaches. This option removes the current

prescriptive details from the TS. The

proposed changes do not affect plant safety

analyses or change the physical design or

operation of the plant. The proposed change

does not increase the total allowable primary

contaiment leakage rate. Therefore, operation

of the facility, in accordance with the

proposed change, does not involve a

significant reduction in the margin of safety.

The NRC has reviewed the licensee’s

analysis and, based on thisreview, it

appears that the three standards of 10

CFR 50.92(c) are satisfied. Therefore, the

NRC staff proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Chattanooga-Hamilton County

Library, 1101 Broad Street, Chattanooga,

Tennessee 37402

Attorney for licensee: General

Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET 11H,

Knoxville, Tennessee 37902

NRC Project Director: Frederick J.

Hebdon

Tennessee Valley Authority, Docket No.

50-328, Sequoyah Nuclear Plant, Unit 2,

Hamilton County, Tennessee

Date of amendment request:

December 12, 1995 (TS 95-23)

Description of amendment request:

The proposed change would incorporate

new requirements associated with steam

generator tube inspections and repair.

The new requirements would establish

alternate steam generator tube plugging

criteria at the tube support plate

intersections.

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:

TVA has evaluated the proposed technical

specification (TS) change and has determined

that it does not represent a significant

hazards consideration based on criteria

established in 10 CFR 50.92(c). Operation of

Sequoyah Nuclear Plant (SQN) in accordance

with the proposed amendment will not:

1. Involve a significant increase in the

probability or consequences of an accident

previously evaluated.

Testing of model boiler specimens for freespan tubing (no tube support plate restraint)

at room temperature conditions shows burst

pressures in excess of 5,000 pounds per

square inch (psi) for indications of outer

diameter stress corrosion cracking with

voltage measurements as high as 19 volts.

Burst testing performed on intersections

pulled from SQN with up to a 1.9-volt

indication shows measured burst pressure in

excess of 6,600 psi at room temperature.

Burst testing performed on pulled tubes from

other plants with up to 7.5-volt indications

shows burst pressures in excess of 5,200 psi

at room temperatures. Correcting for the

effects of temperature on material properties

and minimum strength levels (as the burst

testing was done at room temperature), tube

burst capability significantly exceeds the

safety-factor requirements of NRC Regulatory

Guide (RG) 1.121.

Tube burst criteria are inherently satisfied

during normal operating conditions because

of the proximity of the tube support plate

(TSP). Since tube-to-tube support plate

proximity precludes tube burst during

normal operating conditions, use of the

criteria must retain tube integrity

characteristics that maintain a margin of

safety of 1.43 times the bounding faulted

condition steam line break (SLB) pressure

differential. During a postulated SLB, the

TSP has the potential to deflect during

blowdown following a main SLB, thereby

uncovering the TSP intersections.

Based on the existing database, the RG

1.121 criterion requiring maintenance of a

safety factor of 1.43 times the SLB pressure

differential on tube burst is satisfied by 7/8-

inch-diameter tubing with bobbin coil

indications with signal amplitudes less than

8.82 volts (WCAP-13990), regardless of the

indicated depth measurement. A 2.0-volt

plugging criterion (resulting in a projected

end-of-cycle [EOC] voltage) compares

favorably with the 8.82-volt structural limit

considering the extremely slow apparent

voltage growth rates and few numbers of

indications at SQN. Using the established

methodology of RG 1.121, the structural limit

is reduced by allowances for uncertainty and

growth to develop a beginning of cycle (BOC)

repair limit that would preclude indications

at EOC conditions that exceed the structural

limit. The nondestructive examination (NDE)

uncertainty component is 20.5 percent, and

is based on the Electric Power Research

Institute (EPRI) alternate repair criteria

(ARC).

Test data indicates that tube burst cannot

occur within the TSP, even for tubes that

have 100 percent throughwall electrodischarge machining notches, 0.75 inch long,

provided that the TSP is adjacent to the

notched area. Because of the few number of

indications at SQN, the EPRI methodology of

applying a growth component of 35 percent

per effective full power year (EFPY) will be

used. Near-term operating cycles at SQN are

expected to be bounded by 1.23 years,

therefore, a 43 percent growth component is

appropriate. When these allowances are

added to the BOC alternate plugging criteria

(APC) of 2.0 volts in a deterministic

bounding EOC voltage of approximately 3.26

volts for Cycle 7, operation can be

established. A 5.56-volt deterministic safety

margin exists (8.82 structural limit - 3.26-volt

EOC equal 5.56-volt margin).

For the voltage/burst correlation, the EOC

structural limit is supported by a voltage of

8.82 volts. Using this structural limit of 8.82

volts, a BOC maximum allowable repair limit

can be established using the guidance of RG

1.121. The BOC maximum allowable repair

limit should not permit the existence of EOC

indications that exceed the 8.82-volt

structural limit. By adding NDE uncertainty

allowances and an allowance for crack

growth to the repair limit, the structural limit

can be validated. Therefore, the maximum

allowable BOC repair limit (RL) based on the

structural limit of 8.82 volts can be

represented by the expressions:

RL + (0.205 x RL) + (0.43 x RL) = 8.82

volts, or,

the maximum allowable BOC repair limit

can be expressed as,

RL = 8.82-volt structural limit/1.64 = 5.4

volts.

This RL (5.4 volts) is the appropriate limit

for APC implementation to repair bobbin

indications greater than 2.0 volts

independent of rotating pancake coil (RPC)

confirmation of the indication. This 5.4-volt

upper limit for non-confirmed RPC calls is

consistent with other recently approved APC

programs (Farley Nuclear Plant, Unit 2).

The conservatism of the growth allowance

used to develop the repair limit is shown by

the most recent SQN eddy current data. Only

seven tubes in Unit 2 required repair because

of outside diameter stress corrosion cracking

(ODSCC) at the TSP intersections.

Relative to the expected leakage during

accident condition loadings, it has been

previously established that a postulated main
184 Federal Register / Vol. 61, No. 2 / Wednesday, January 3, 1996 / Notices

SLB outside of containment, but upstream of

the main steam isolation valve (MSIV),

represents the most limiting radiological

condition relative to the APC.

Implementation of the APC will determine

whether the distribution of cracking

indications at the TSP intersections is

projected to be such that primary-tosecondary leakage would result in site

boundary doses within a small fraction of the

10 CFR 100 guidelines. A separate analysis

has determined this allowable SLB leakage

limit to be 3.7 gallons per minute (gpm) in

the faulted loop. This limit uses the TS

reactor coolant system (RCS) Iodine-131

activity level of 1.0 microcuries per gram

dose equivalent Iodine-131 and the

recommended Iodine-131 transient spiking

values consistent with NUREG-0800. The

analysis method is WCAP-14277, which is

consistent with the guidance of the NRC

generic letter (GL) [95-05] and will be used

to calculate EOC leakage. Because of the

relatively low number of indications at SQN,

it is expected that the actual leakage values

will be far less than this limit. Additionally,

the current Iodine-131 levels at SQN range

from about 25 to 100 times less than the TS

limit.

Application of the criteria requires the

projection of postulated SLB leakage, based

on the projected EOC voltage distribution for

Cycle 8 operation. Projected EOC voltage

distribution is developed using the most

recent EOC eddy current results and a voltage

measurement uncertainty. Data indicates that

a threshold voltage of 2.8 volts would result

in throughwall cracks long enough to leak at

SLB condition. The GL requires that all

indications to which the APC are applied

must be included in the leakage projection.

Tube pull results from another plant with 7/

8-inch tubing with a substantial voltage

growth database have shown that tube wall

degradation of greater than 40 percent

throughwall was readily detectable either by

the bobbin or RPC probe. The tube with

maximum throughwall penetration of 56

percent (42 average) had a voltage of 2.02

volts. The SQN Unit 1 pulled tube had a

1.93-volt indication with a maximum depth

of 91 percent and did not leak at SLB

condition. Based on the SQN pulled tube and

industry pulled tube data supporting a lower

threshold for SLB leakage of 2.8 volts,

inclusion of all APC intersections in the

leakage model is quite conservative. The

ODSCC occurring at SQN is in its earliest

stages of development. The conservative

bounding growth estimations to be applied to

the expected small number of indications for

the upcoming inspection should result in

very small levels of predicted SLB leakage.

Historically, SQN has not identified ODSCC

as a contributor to operational leakage.

In order to assess the sensitivity of an

indication’s BOC voltage to EOC leakage

potential, a Monte Carlo simulation was

performed for a 2.0-volt BOC indication.

The maximum EOC voltage (at 99.8 percent

cumulative probability) was found to be 4.8

volts. The leakage component from an

indication of this magnitude, using the EPRI

leakage model, is 0.028 gpm.

Therefore, as implementation of the 2.0-

volt APC does not adversely affect steam

generator (S/G) tube integrity and

implementation will be shown to result in

acceptable dose consequences, the proposed

amendment does not result in significant

increase in the probability or consequences

of an accident previously evaluated.

2. Create the possibility of a new or

different kind of accident from any

previously analyzed.

Implementation of the proposed S/G tube

APC does not introduce any significant

changes to the plant design basis. Use of the

criteria does not provide a mechanism that

could result in an accident outside of the

region of the TSP elevations; no ODSCC is

occurring outside the thickness of the TSP.

Neither a single or multiple tube rupture

event would be expected in a S/G in which

the plugging criteria is applied (during all

plant conditions).

TVA will implement a maximum leakage

rate limit of 150 gallon per day per S/G to

help preclude the potential for excessive

leakage during all plant conditions. The SQN

TS limits on primary-to-secondary leakage at

operating conditions include a maximum of

0.42 gpm (600 gallons per day [gpd]) for all

S/Gs, or, a maximum of 150 gpd for any one

S/G. The RG 1.121 criterion for establishing

operational leakage rate limits that require

plant shutdown is based upon leak-beforebreak considerations to detect a free-span

crack before potential tube rupture during

faulted plant conditions. The 150-gpd limit

should provide for leakage detection and

plant shutdown in the event of the

occurrence of an unexpected single crack

resulting in leakage that is associated with

the longest permissible crack length. RG

1.121 acceptance criteria for establishing

operating leakage limits are based on leakbefore-break considerations such that plant

shutdown is initiated if the leakage

associated with the longest permissible crack

is exceeded. The longest permissible crack is

the length that provides a factor of safety of

1.43 against bursting at faulted conditions

maximum pressure differential. A voltage

amplitude of 8.82 volts for typical ODSCC

corresponds to meeting this tube burst

requirement at a lower 95 percent prediction

limit on the burst correlation coupled with

95/95 lower tolerance limit material

properties. Alternate crack morphologies can

correspond to 8.82 volts so that a unique

crack length is not defined by the burst

pressure versus voltage correlation.

Consequently, typical burst pressure versus

through-wall crack length correlations are

used below to define the ‘‘longest permissible

crack’’ for evaluating operating leakage

limits.

The single through-wall crack lengths that

result in tube burst at 1.43 times the SLB

pressure differential and the SLB pressure

differential alone are approximately 0.57

inch and 0.84 inch, respectively. A leak rate

of 150 gpd will provide for detection of 0.4-

inch-long cracks at nominal leak rates and

0.6-inch-long cracks at the lower 95 percent

confidence level leak rates. Since tube burst

is precluded during normal operation

because of the proximity of the TSP to the

tube and the potential exists for the crevice

to become uncovered during SLB conditions,

the leakage from the maximum permissible

crack must preclude tube burst at SLB

conditions. Thus, the 150-gpd limit provides

for plant shutdown before reaching critical

crack lengths for SL-conditions.

Additionally, this leak-before-break

evaluation assumes that the entire crevice

area is uncovered during blowdown. Partial

uncover will provide benefit to the burst

capacity of the intersection.

As S/G tube integrity upon implementation

of the 2.0-volt APC continues to be

maintained through in-service inspection and

primary-to-secondary leakage monitoring, the

possibility of a new or different kind of

accident from any accident previously

evaluated is not created.

3. Involve a significant reduction in a

margin of safety.

The use of the voltage based APC at SQN

is demonstrated to maintain S/G tube

integrity commensurate with the criteria of

RG 1.121. RG 1.121 describes a method

acceptable to the NRC Staff for meeting

General Design Criteria (GDC) 14, 15, 31, and

32 by reducing the probability or the

consequences of S/G tube rupture. This is

accomplished by determining the limiting

conditions of degradation of S/G tubing, as

established by in-service inspection, for

which tubes with unacceptable cracking

should be removed from service. Upon

implementation of the criteria, even under

the worst-case conditions, the occurrence of

ODSCC at the TSP elevations is not expected

to lead to a S/G tube rupture event during

normal or faulted plant conditions. The EOC

distribution of crack indications at the TSP

elevations will be confirmed to result in

acceptable primary-to-secondary leakage

during all plant conditions and radiological

consequences are not adversely impacted.

In addressing the combined effects of lossof-coolant accident (LOCA), plus safe

shutdown earthquake (SSE) on the S/G

component (as required by GDC 2), it has

been determined that tube collapse may

occur in the S/Gs at some plants. This is the

case as the TSP may become deformed as a

result of lateral loads at the wedge supports

at the periphery of the plate because of the

combined effects of the LOCA rarefaction

wave and SSE loadings. Then, the resulting

pressure differential on the deformed tubes

may cause some of the tubes to collapse.

There are two issues associated with S/G

tube collapse. First, the collapse of S/G

tubing reduces the RCS flow area through the

tubes. The reduction in flow area increases

the resistance to flow of steam from the core

during a LOCA, which in turn, may

potentially increase peak clad temperature

(PCT). Second, there is a potential that partial

through-wall cracks in tubes could progress

to through-wall cracks during tube

deformation or collapse.

Consequently, since the leak-before-break

methodology is applicable to the SQN reactor

coolant loop piping, the probability of breaks

in the primary loop piping is sufficiently low

that they need not be considered in the

structural design of the plant. The limiting

LOCA event becomes either the accumulator

line break or the pressurizer surge line break.

LOCA loads for the primary pipe breaks were

used to bound the conditions at SQN for

smaller breaks. The results of the analysis
Federal Register / Vol. 61, No. 2 / Wednesday, January 3, 1996 / Notices 185

using the larger break inputs show that the

LOCA loads were found to be of insufficient

magnitude to result in S/G tube collapse or

significant deformation. The LOCA, plus SSE

tube collapse evaluation performed for

another plant with Series 51 S/Gs using

bounding input conditions (large-break

loadings), is applicable to SQN. Therefore, at

SQN, no tubes will be excluded from using

the voltage repair criteria due to deformation

of collapse of S/G tubes following a LOCA

plus an SSE. Additional supporting

information relative to NRC review of J.M.

Farley Nuclear Plant was provided in

Enclosure 5, Item 3 of TVA’s submittal dated

September 7, 1995 (TAC No. M92961).

Addressing RG 1.83 considerations,

implementation of the bobbin probe voltage

based interim tube plugging criteria of 2.0

volt is supplemented by: (1) enhanced eddy

current inspection guidelines to provide

consistency in voltage normalization, (2) a

100 percent eddy current inspection sample

size at the TSP elevations, and (3) RPC

inspection requirements for the larger

indications left in service to characterize the

principal degradation as ODSCC.

As noted previously, implementation of

the TSP elevation plugging criteria will

decrease the number of tubes that must be

repaired. The installation of S/G tube plugs

reduces the RCS flow margin. Thus,

implementation of the alternate plugging

criteria will maintain the margin of flow that

would otherwise be reduced in the event of

increased tube plugging.

Based on the above, it is concluded that the

proposed license amendment request does

not result in a significant reduction in margin

of safety.

The NRC has reviewed the licensee’s

analysis and, based on thisreview, it

appears that the three standards of 10

CFR 50.92(c) are satisfied. Therefore, the

NRC staff proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: Chattanooga-Hamilton County

Library, 1101 Broad Street, Chattanooga,

Tennessee 37402

Attorney for licensee: General

Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET 11H,

Knoxville, Tennessee 37902

NRC Project Director: Frederick J.

Hebdon

TU Electric Company, Docket Nos. 50-

445 and 50-446, Comanche Peak Steam

Electric Station, Units 1 and 2,

Somervell County, Texas

Date of amendment request:

November 21, 1995

Brief description of amendments: The

proposed amendments would modify

the Comanche Peak Steam Electric

Station (CPSES) Units 1 and 2 Technical

Specifications (TS) to allow the

containment personnel airlock (PAL)

doors to remain open during movement

of irradiated fuel and during core

alterations.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Do the proposed changes involve a

significant increase in the probability or

consequences of an accident previously

evaluated?

The proposed change allows the PAL doors

for containment to remain open during the

movement of irradiated fuel and core

alterations. Whether or not the PAL doors are

open does not effect the movement of fuel,

the strict compliance with the procedures

governing refueling operations, or the

integrity of fuel assemblies. The position of

the airlock doors cannot, in itself, be the

initiating event in any accident. The

probability of a fuel handling accident is not

changed.

The consequences of leaving the airlock

doors open during this accident are bounded

by the existing analysis, provided the fuel

handling accident assumptions are

maintained (e.g. 100 hours after reactor

shutdown and the water level remains 23 feet

above the fuel). The existing analysis

postulates the limiting fuel handling accident

to occur in the Fuel Building with no credit

taken for barrier or filtration. This accident

analysis envelopes the proposed change for

a fuel handling accident occurring in the

Containment Building.

Were a fuel handling accident to occur

with the PAL doors open, the impact would

be minimal. Pressure is expected to be

essentially equalized across the door with

little air flow either into or out of

containment. Based on transport time from

the location of the accident to the PAL, little,

if any, radioactive material is expected to

escape containment via the PAL. The amount

that might escape would not necessarily be

anymore than might escape as the door is

cycled to evacuate personnel. What does

escape will be filtered by the Primary Plant

Ventilation System, the same as if the

accident were to occur in the fuel building.

In summary, not only is the accident clearly

bounded by the existing analysis, the actual

increase in release of radioactive material

outside the plant will be insignificant if there

is any measurable increase at all.

Based on the above, allowing the PAL

doors to remain open during movement of

irradiated fuel and core alterations, has no

significant effect on the probability or

consequences of an accident previously

evaluated.

2. Do the proposed changes create the

possibility of a new or different type of

accident from any accident previously

evaluated?

The change does not add new hardware.

The only change in the operation of the plant

is that the PAL doors will remain open

during movement of irradiated fuel and core

alterations. Because the current fuel handling

accident analysis considers fuel handling

accidents in either the Fuel Building or the

Containment Building, the current fuel

handling accident analysis remains bounding

for the proposed change. Therefore, the

proposed change does not create the

possibility of a new or different type of

accident from any accident previously

evaluated.

3. Do the proposed changes involve a

significant reduction in a margin of safety?

The assumptions used to calculate the

offsite dose resulting from a fuel handling

accident in [the] Containment Building are

equivalent to assuming that the PAL remains

open for the entire accident and that no

filtration occurs. Since no credit was taken

for any containment barrier or ventilation

system filtration, the dose to the public as

calculated in the analysis is not affected by

this change. Therefore, the proposed change

does not involve a significant reduction in

the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: University of Texas at

Arlington Library, Government

Publications/Maps, 702 College, P.O.

Box 19497, Arlington, TX 76019

Attorney for licensee: George L. Edgar,

Esq., Morgan, Lewis and Bockius, 1800

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

NRC Project Director: William D.

Beckner

TU Electric Company, Docket Nos. 50-

445 and 50-446, Comanche Peak Steam

Electric Station, Units 1 and 2,

Somervell County, Texas

Date of amendment request:

November 21, 1995

Brief description of amendments: The

proposed amendment would revise the

core safety limit curves and revised N16 Overtemperature reactor trip

setpoints as a result of the reload

analyses for CPSES Unit 2, Cycle 3. In

addition, the minimum required Reactor

Coolant System (RCS) flow is increased

and an administrative enhancement is

included in the footnotes of the RCS

flow - low reactor trip function setpoint

for both Units.

Basis for proposed no significant

hazards consideration determination:

As required by 10 CFR 50.91(a), the

licensee has provided its analysis of the

issue of no significant hazards

consideration, which is presented

below:

1. Do the proposed changes involve a

significant increase in the probability or

consequences of an accident previously

evaluated?

A. Increase in Unit 2 minimum required

flow

This revision increases the Unit 2

minimum required RCS flow rate assumed in
186 Federal Register / Vol. 61, No. 2 / Wednesday, January 3, 1996 / Notices

the safety analyses by 3.6%. The actual core

flow is unchanged and is approximately

6.6% higher than the value assumed in

previous accident analyses. The remaining

3.0% flow is sufficient to account for all

uncertainties associated with the core flow

measurement.

Since this change only involves analysis

methodology and does not affect the actual

core flow, it does not increase the actual

probability or consequences of any

postulated accident.

When considered separately, increasing the

minimum required RCS flow is a

conservative change. Although there is no

impact on the initiation of any postulated

accidents, the potential severity of the

affected accidents is typically less when flow

is increased. In general, the increased ability

to remove heat from the fuel will reduce the

peak temperature seen by the fuel and reduce

the potential for undesirable boiling

conditions. Thus, the increase in the

assumed RCS flow will not increase the

probability or consequences of an accident

previously analyzed.

B. Revision to the Unit 2 Core Safety Limits

Analyses of reactor core safety limits are

required as part of reload calculations for

each cycle. TU Electric has performed inhouse analyses of the Unit 2, Cycle 3 core to

determine the reactor core safety limits. The

newer methodologies and safety analysis

values result in new operating curves which,

in general, permit plant operation over a

similar range of acceptable conditions. This

change means that if a transient were to

occur with the plant operating at the limits

of the new curve, a higher temperature and

power level might be attained than if the

plant were operating within the bounds of

the old curves. However, since the new

curves were developed using approved

methodologies which are wholly consistent

with and do not represent a change in the

Technical Specification bases for safety

limits, all applicable postulated transients

will continue to be properly mitigated. As a

result, there will be no significant increase in

the consequences, as determined by accident

analyses, of any accident previously

evaluated.

C. Revision to Unit 2 Overtemperature N16 Reactor Trip Setpoints, Parameters and

Coefficients

As a result of changes discussed, the

Overtemperature N-16 reactor trip setpoint

has been recalculated. These trip setpoints

help ensure that the core safety limits are

maintained and that all applicable limits of

the safety analysis are met.

Based on the calculations performed, the

safety analysis value for Overtemperature N16 reactor trip setpoint has changed. This

essentially means if a transient were to occur,

the actual temperature and power level could

be slightly higher. However, the analyses

performed show that, using the TU Electric

methodologies, all reactor core safety limits

are met and all applicable limits of the safety

analysis are met. This parameter has a

setpoint which allows the mitigation of

postulated accidents and has no impact on

accident initiation. Therefore, the changes in

safety analysis values do not involve an

increase in the probability of an accident

and, based on satisfying the core safety limits

and all applicable safety analysis limits, there

is no significant increase in the consequences

of any accident previously evaluated.

In addition, the changes result in setpoint

values which potentially offer safety benefits.

The risk of turbine runbacks or reactor trips

due to upper plenum flow anomalies will be

minimized with a higher overtemperature

setpoint, thus reducing potential challenges

to the plant safety systems. A final benefit is

that the new methods for considering N-16

setpoints and values will be consistent with

Unit 1, which reduces the potential for

personnel error due to unit differences.

Considering both the safety analysis impact

and the benefits described above, the changes

in N-16 setpoints and parameters will result

in slight reduction in the probability of an

accident and do not significantly increase the

consequences of an accident previously

evaluated.

D. Deletion of footnotes associated with the

RCS flow - low reactor trip setpoint

In lieu of revising the footnotes to support

the Unit 2 Cycle 3 operation, the deletion of

the footnote is proposed. Further, for

consistency with Unit 2, the same change is

proposed for Unit 1. This change will not

affect current plant practice; however, it will

impose a more restrictive RCS flow - low

setpoint than is currently required. The RCS

flow - low reactor trip setpoint is currently

specified in Technical Specification Table

2.2-1, Functional Unit 12.b, to be 90% of the

minimum measured RCS flow. The proposed

change would require the setpoint to be 90%

of the instrument span where 100% of

instrument span approximately corresponds

to the actual RCS flow. The actual RCS flow

is verified to be greater than the RCS flow

assumed in the accident analysis through

compliance with Technical Specification

3.2.5. Thus, through deletion of the footnotes,

the RCS volumetric flow corresponding to

the reactor trip setpoint will be greater than

or equal to the volumetric flow allowed by

the current specifications.

In summary, the proposed deletion of the

footnotes will have no impact on current

plant operations. A possible relaxation of the

RCS flow - low setpoint which is currently

allowed by the Technical Specifications will

be removed without creating the potential for

unnecessary plant trips.

The RCS flow - low reactor trip setpoint

can have no effect on the probability of an

accident. Because the reactor will be tripped

at or prior to the conditions assumed in the

accident analyses, there will be no effect on

the consequences of an accident previously

identified.

SUMMARY

The changes in the amendment request

applies new NRC approved methodologies,

changes in safety analysis values, new core

safety limits and new N-16 setpoint and

parameter values to assure that all applicable

safety analysis limits have been met. The

potential for an operational transient to occur

has been reduced and there has been no

significant impact on the consequences of

any accident previously evaluated.

2. Do the proposed changes create the

possibility of a new or different kind of

accident from any accident previously

evaluated?

The proposed changes involve the use of

revised safety analysis values and the

calculation of new reactor core safety limits

and reactor trip setpoints. As such, the

changes play an important role in the

analysis of postulated accidents but none of

the changes effect plant hardware or the

operation of plant systems in a way that

could initiate an accident. Therefore, the

proposed changes do not create the

possibility of a new or different kind of

accident from any accident previously

evaluated.

3. Do the proposed changes involve a

significant reduction in a margin of safety?

In reviewing and approving the methods

used for safety analyses and calculations, the

NRC has approved the safety analysis limits

which establish the margin of safety to be

maintained. While the actual impact on

safety is discussed in response to question 1,

the impact on margin of safety is discussed

below.

A. Increase in the Unit 2 minimum

required flow

In performing the DNB-related analyses,

the Reactor Coolant System flow rate

assumed in these analyses is increased by 3.6

percent to insure that all applicable limits of

the safety analysis are met. The Technical

Specification 3/4.2.5 limit for this parameter

will be changed to insure that it is

maintained within the normal steady-state

envelope of operation assumed in the

transient and accident safety analyses (i.e.,

ensuring that the RCS flow rate assumed in

the safety analyses remains valid). The

Technical Specification limits are consistent

with the initial safety analysis assumption

(plus uncertainties) and have been

analytically demonstrated to be adequate to

maintain a minimum DNBR at or above the

safety analysis DNBR limit throughout each

analyzed transient. Because the 95/95 DNBR

acceptance criteria is met with the proposed

change and assumptions of the safety

analyses are maintained valid by the

Technical Specification limits, there is no

change in a margin of safety.

B. Revision to the Unit 2 Reactor Core

Safety Limits

The TU Electric reload analysis methods

have been used to determine new reactor

core safety limits. All applicable safety

analysis limits have been met. The methods

used are wholly consistent with Technical

Specification BASES 2.1 which is the bases

for the safety limits. In particular, the curves

assure that for Unit 2, Cycle 3, the calculated

DNBR is no less than the safety analysis limit

and the average enthalpy at the vessel exit is

less than the enthalpy of saturated liquid.

In conjunction with the reactor core safety

limit methodology, the NRC approved TUE1 DNB correlation is used for performing

DNB-related analyses. This correlation will

be applied to the core configuration of CPSES

Unit 2, Cycle 3 and future core

configurations. The TUE-1 correlation DNBR

limit is established such that there is a 95

percent probability with 95 percent

confidence level that DNB will not occur

when the minimum DNBR for the limiting

fuel is greater than or equal to the TUE-1

correlation DNBR limit. This 95/95 criteria

defines the ‘‘margin of safety’’ for the DNB-
Federal Register / Vol. 61, No. 2 / Wednesday, January 3, 1996 / Notices 187

related analysis and remains valid even

though the DNB correlation and associated

correlation limit are changed. Margin is

provided in the DNB-related analysis for

known and potential effects such as

hydraulic differences between the two coresident fuel assembly designs and the

presence of the Reactor Coolant System lower

plenum flow anomaly. The TUE-1 correlation

DNBR limit plus margin constitutes the

safety analysis DNBR limit. The accident

analyses are performed to ensure that the

safety analysis DNBR limit acceptance

criteria are satisfied. Because the 95/95

DNBR acceptance criteria remains valid and

continues to be satisfied, no change in a

margin of safety occurs.

C. Revision to Unit 2 Overtemperature N16 Reactor Trip Setpoints, Parameters and

Coefficients

Because the reactor core safety limits for

CPSES Unit 2, Cycle 3 are recalculated, the

Reactor Trip System instrumentation setpoint

values for the Overtemperature N-16 reactor

trip setpoint which protect the reactor core

safety limits must also be recalculated. The

Overtemperature N-16 reactor trip setpoint

helps prevent the core and Reactor Coolant

System from exceeding their safety limits

during normal operation and design basis

anticipated operational occurrences. The

most relevant design basis analysis in

Chapter 15 of the CPSES Final Safety

Analysis Report (FSAR) which is affected by

the change in the safety analysis value for the

CPSES Unit 2 Overtemperature N-16 reactor

trip setpoint is the Uncontrolled Rod Cluster

Control Assembly Bank Withdrawal at Power

(FSAR Section 15.4.2). This event has been

re-analyzed with the revised safety analysis

value for the Overtemperature N-16 reactor

trip setpoint to demonstrate compliance with

event specific acceptance criteria. Because all

event acceptance criteria are satisfied, there

is no degradation in a margin of safety.

The nominal Reactor Trip System

instrumentation setpoints values for the

Overtemperature N-16 reactor trip setpoint

(Technical Specification Table 2.2-1) are

determined based on a statistical

combination of all of the uncertainties in the

channels to arrive at a total uncertainty. The

total uncertainty plus additional margin is

applied in a conservative direction to the

safety analysis trip setpoint value to arrive at

the nominal and allowable values presented

in Technical Specification Table 2.2-1.

Meeting the requirements of Technical

Specification Table 2.2-1 assures that the

Overtemperature N-16 reactor trip setpoint

assumed in the safety analyses remains valid.

The CPSES Unit 2, Cycle 3 Overtemperature

N-16 reactor trip setpoint is different from

previous cycles which provides more

operational flexibility to withstand mild

transients without initiating automatic

protective actions. Although the setpoint is

different, the Reactor Trip System

instrumentation setpoint values for the

Overtemperature N-16 reactor trip setpoint

are consistent with the safety analysis

assumption which has been analytically

demonstrated to be adequate to meet the

applicable event acceptance criteria. Thus,

there is no reduction in a margin of safety.

D. Deletion of footnotes associated with the

RCS flow - low reactor trip function

The deletion of the footnotes, and the

potential relaxation of the RCS flow - low

setpoint which could be used, will provide

further assurance that, in the event of a

partial loss of forced RCS flow or locked rotor

transient, a reactor trip signal would be

initiated prior to the conditions assumed in

the accident analyses. Thus, the accident

analyses are unaffected, and there is no

reduction in a margin of safety.

SUMMARY

The proposed changes to the CPSES

Technical Specifications involve using NRCapproved licensing analysis methods

developed by TU Electric to determine the

Technical Specification reactor core safety

limits and perform DNB-related analysis for

CPSES Unit 2, Cycle 3. The DNB-related

analyses are performed by TU Electric using

a qualified, state-of-the-art departure from

nucleate boiling (DNB) correlation, TUE-1,

which has also been approved by the NRC for

the CPSES Unit 2, Cycle 3 core configuration.

In performing these analyses, the minimum

required Reactor Coolant System flow rate is

increased by 3.6 percent. Because the core

safety limits for CPSES Unit 2, Cycle 3 are

recalculated, the Reactor Trip System

instrumentation setpoints values for the

Overtemperature N-16 reactor trip setpoint

which protect the core safety limits are also

recalculated.

Using the NRC approved TU Electric

methods, the reactor core safety limits are

determined such that all applicable limits of

the safety analyses are met, particularly the

95/95 DNBR limit. The Technical

Specification 3/4.2.5 limits for the DNB

Parameters insure the assumptions in the

safety analyses remain valid. Because the

applicable event acceptance criteriacontinue

to be met, there is no significant reduction in

the margin of safety.

The NRC staff has reviewed the

licensee’s analysis and, based on this

review, it appears that the three

standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

proposes to determine that the

amendment request involves no

significant hazards consideration.

Local Public Document Room

location: University of Texas at

Arlington Library, Government

Publications/Maps, 702 College, P.O.

Box 19497, Arlington, TX 76019

Attorney for licensee: George L. Edgar,

Esq., Morgan, Lewis and Bockius, 1800

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

NRC Project Director: William D.

Beckner

Virginia Electric and Power Company,

Docket Nos. 50-338 and 50-339, North

Anna Power Station, Units No. 1 and

No. 2, Louisa County, Virginia

Date of amendment request: October

17, 1995

Description of amendment request:

The proposed amendment would

modify the North Anna Power Station,

Units 1 and 2 Technical Specifications

(TS) to allow both of the containment

personnel airlock doors to remain open

during refueling operations, delete the

license condition referencing the

analyses for limiting doses to the control

room operators, and modify the TS

Bases to clarify the emergency power

system requirements relative to

mitigation of the consequences of a Fuel

Handling Accident.

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:

There is no significant change in the

probability or consequences of an accident

previously evaluated. There are no system

changes which would increase the

probability of an accident occurring.

Allowing both personnel airlock doors to

remain open during core alterations or fuel

movement inside containment will not have

any impact on the probability of a Fuel

Handling Accident either in containment or

in the fuel building. The consequences of a

Fuel Handling Accident have been

investigated by performing a reanalysis with

no credit for isolation or filtration by the Fuel

Building or containment ventilation systems.

The Exclusion Area Boundary [EAB] and

Low Population Zone [LPZ] doses for a Fuel

Handling Accident without credit for iodine

filtration remain well within (<25%) of the

NRC regulatory limits of 10 CFR [Part] 100.

The predicted control room operator doses

remain bounded by the limiting case for

control room doses and within the regulatory

limits of General Design Criterion [GDC] 19.

In addition, the action to clarify the

responses to NRC question 6.72 [of the

original Final Safety Analysis Report] will

not increase the probability or consequences

of the Fuel Handling Accident.

No new accident types or equipment

malfunction scenarios are introduced as a

result of the clarification to the Virginia

Power response to [NRC question] 6.72 or as

a result of these changes in analysis methods

or the proposed Technical Specifications

changes to allow both personnel airlock

doors to remain open during core alterations

or fuel movement inside containment.

Therefore, there is no possibility of an

accident of a different type than any

previously evaluated in the North Anna

USFAR [Updated Final Safety Analysis

Report].

There is no significnt reduction in the

margin of safety. An evaluation of the Fuel

Handling Accident doses at the EAB, the LPZ

and to control room operators has been

performedand it has been concluded that the

acceptance criteria defined by GDC-19, 10

CFR 100, and the NRC Standard Review Plan

will continue to be met.

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
188 Federal Register / Vol. 61, No. 2 / Wednesday, January 3, 1996 / Notices

involves no significant hazards

consideration.

Local Public Document Room

location: The Alderman Library, Special

Collections Department, University of

Virginia, Charlottesville, Virginia 22903-

2498.

Attorney for licensee: Michael W.

Maupin, Esq., Hunton and Williams,

Riverfront Plaza, East Tower, 951 E.

Byrd Street, Richmond, Virginia 23212.

NRC Project Director: David B.

Matthews

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 Nos. 50-325 and 50-324,

Brunswick Steam Electric Plant, Units

1 and 2, Brunswick County, North

Carolina

Date of application for amendments:

September 13, 1995, as amended

November 27, 1995

Brief description of amendments: The

proposed amendments would permit

the licensee to implement the

performance-based option provided by

10 CFR Part 50, Appendix J, which

allows leakage testing intervals to be

based on system and component testing

performance.

Date of publication of individual

notice in Federal Register: December 12,

1995 (60 FR 63739)

Expiration date of individual notice:

January 11, 1996

Local Public Document Room

location: The University of North

Carolina at Wilmington, William

Madison Randall Library, 601 S. College

Road, Wilmington, North Carolina

28403-3297

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.

Arizona Public Service Company, et al.,

Docket Nos. STN 50-529 and STN 50-

530, Palo Verde Nuclear Generating

Station, Units 2 and 3, Maricopa

County, Arizona

Date of application for amendments:

October 3, 1995

Brief description of amendments: The

amendments delete Sections 2.B.(7)(a)

and (b) of

Facility Operating License No. NPF-51

(Unit 2) and Sections 2.b.(6)(a) and (b)

of

Facility Operating License No. NPF-74

(Unit 3) relating to certain previous sale

and leaseback transactions that were

added by Amendment No. 3 for NPF-51

and Amendment No. 1 for NPF-74.

Date of issuance: December 8, 1995

Effective date: December 8, 1995

Amendment Nos.: Unit 2 -

Amendment No. 91; Unit 3 -

Amendment No. 74

Facility Operating License Nos. NPF51 and NPF-74: The amendments

revised the license.

Date of initial notice in Federal

Register: November 8, 1995 (60 FR

56363) The Commission’s related

evaluation of the amendments is

contained in a Safety Evaluation dated

December 8, 1995.No significant

hazards consideration comments

received: No.

Local Public Document Room

location: Phoenix Public Library, 1221

N. Central Avenue, Phoenix, Arizona

85004

Carolina Power & Light Company, et

al., Docket Nos. 50-325 and 50-324,

Brunswick Steam Electric Plant, Units

1 and 2, Brunswick County, North

Carolina

Date of application for amendments:

October 23, 1995

Brief description of amendments: The

amendments revised the Technical

Specifications to delete the applicability

of the primary coolant water chemistry

limits when the primary system is being

chemically decontaminated and the

reactor vessel is defueled.

Date of issuance: December 13, 1995

Effective date: December 13, 1995

Amendment Nos.: 180 and 211

Facility Operating License Nos. DPR71 and DPR-62.

Date of initial notice in Federal

Register: The Commission’s related

evaluation of the amendments is

contained in a Safety Evaluation dated

December 13, 1995. No significant

hazards consideration comments

received: No.

Local Public Document Room

location: University of North Carolina at

Wilmington, William Madison Randall

Library, 601 S. College Road,

Wilmington, North Carolina 28403-

3297.

Commonwealth Edison Company,

Docket Nos. STN 50-454 and STN 50-

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

Ogle County, Illinois Docket Nos. STN

50-456 and STN 50-457, Braidwood

Station, Unit Nos. 1 and 2, Will County,

Illinois

Date of application for amendments:

September 14, 1995, as supplemented

November 8, 1995.

Brief description of amendments: The

amendments allow the use of an

alternate zirconium based fuel cladding,
Federal Register / Vol. 61, No. 2 / Wednesday, January 3, 1996 / Notices 189

ZIRLO, and permit limited substitution

of fuel rods with ZIRLO filler rods. In

addition, a clarification and an editorial

change have been included.

Date of issuance: December 19, 1995

Effective date: December 19, 1995

Amendment Nos.: 78 and 70

Facility Operating License Nos. NPF37, NPF-66, NPF-72 and NPF-77: The

amendments revised the Technical

Specifications.

Date of initial notice in Federal

Register: October 25, 1995 (60 FR

54716) The November 8, 1995 letter,

provided clarifying information that did

not change the scope of the September

14, 1995, application and the initial

proposed no significant hazards

consideration determination.The

Commission’s related evaluation of the

amendments is contained in a Safety

Evaluation dated December 19, 1995No

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.

Commonwealth Edison Company,

Docket Nos. 50-237 and 50-249,

Dresden Nuclear Power Station, Units 2

and 3, Grundy County, Illinois Docket

Nos. 50-254 and 50-265, Quad Cities

Nuclear Power Station, Units 1 and 2,

Rock Island County, Illinois

Date of application for amendments:

September 15, 1995.

Brief description of amendments: The

amendments upgrade the current

custom Technical Specifications (TS)

for Dresden and Quad Cities to the

Standard Technical Specifications

contained in NUREG-0123, ‘‘Standard

Technical Specification General Electric

Plants BWR/4.’’ The application dated

September 15, 1995, contains some of

the TSUP open items from previous

Dresden and Quad Cities TS

amendments issued by the NRC.

Date of issuance: December 19,

1995Effective date: Immediately, to be

implemented no later than June 30,

1996.

Amendment Nos.: 145, 139, 167 and

163

Facility Operating License Nos. DPR19, DPR-25, DPR-29 and DPR-30. The

amendments revised the Technical

Specifications.

Date of initial notice in Federal

Register: October 5, 1995 (60 FR 52220)

The Commission’s related evaluation of

the amendments is contained in a Safety

Evaluation dated December 19, 1995.No

significant hazards consideration

comments received: No

Local Public Document Room

location: for Dresden, Morris Area

Public Library District, 604 Liberty

Street, Morris, Illinois 60450; for Quad

Cities, Dixon Public Library, 221

Hennepin Avenue, Dixon, Illinois

61021.

Consolidated Edison Company of New

York, Docket No. 50-247, Indian Point

Nuclear Generating Unit No. 2,

Westchester County, New York

Date of amendment request:

December 8, 1995

Brief description of amendment: The

amendment revises Technical

Specification (TS) 3.1.A.5 to revise the

wording to allow a single train of

Power-Operated Relief Valves (PORVs)/

Block Valves to be closed and

deenergized indefinitely. The proposed

change is administrative and is intended

to correct inconsistencies between the

intended operation of the PORVs/Block

Valves and the language of the TSs.

Date of issuance: December 8, 1995

Effective date: As of the date of

issuance to be implemented

immediately.

Amendment No.: 185

Facility Operating License No. DPR26: Amendment revised the Technical

Specifications.Public comments

requested as to proposed no significant

hazards consideration: NoThe

Commission’s related evaluation of the

amendment, emergency circumstances

and consultation with the State, and

final determination of no significant

hazards consideration are contained in

a Safety Evaluation dated December 8,

1995.

Local Public Document Room

location: White Plains Public Library,

100 Martine Avenue, White Plains, New

York 10610.

Attorney for licensee: Brent L.

Brandenburg, Esq., 4 Irving Place, New

York, New York 10003.

NRC Project Director: Ledyard B.

Marsh

Duke Power Company, et al., Docket

Nos. 50-413 and 50-414, Catawba

Nuclear Station, Units 1 and 2, York

County, South Carolina

Date of application for amendments:

October 31, 1994

Brief description of amendments: The

amendments remove the stroke times for

the steam generator power operated

relief valves from Technical

Specification Tables 3.6-2a and 3.6-2b.

Date of issuance: December 18, 1995

Effective date: As of the date of

issuance to be implemented within 30

days

Amendment Nos.: Unit 1 - 139 - Unit

2 - 133

Facility Operating License Nos. NPF35 and NPF-52: Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: February 15, 1995 (60 FR

8745) The Commission’s related

evaluation of the amendments is

contained in a Safety Evaluation dated

December 18, 1995.No significant

hazards consideration comments

received: No

Local Public Document Room

location: York County Library, 138 East

Black Street, Rock Hill, South Carolina

29730

Entergy Operations, Inc., Docket Nos.

50-313 and 50-368, Arkansas Nuclear

One, Unit Nos. 1 and 2, Pope County,

Arkansas

Date of amendment request:

September 4, 1993, as supplemented by

letters dated February 16, 1994, and

August 4, 1995

Brief description of amendments: The

license amendments revised the

Arkansas Nuclear One Industrial

Security Plan.

Date of issuance: December 19, 1995

Effective date: December 19, 1995

Amendment Nos.: 183 and 172

Facility Operating License Nos. DPR51 and NPF-6. Amendments revised the

licenses.

Date of initial notice in Federal

Register: November 8, 1995 (60 FR

56368) The Commission’s related

evaluation of the amendments is

contained in a Safety Evaluation dated

December 19, 1995.No significant

hazards consideration comments

received: No.

Local Public Document Room

location: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801

Entergy Operations, Inc., Docket No.

50-382, Waterford Steam Electric

Station, Unit 3, St. Charles Parish,

Louisiana

Date of amendment request:

December 6, 1993, as supplemented by

letters dated May 12, August 9, and

September 18, 1995.

Brief description of amendment: The

amendment changes the Appendix A

TSs to allow installation of steam

generator tube repair sleeves at the

Waterford Steam Electric Station, Unit

3. The sleeves are designed and

manufactured by Combustion

Engineering Incorporated.

Date of issuance: December 14, 1995

Effective date: December 14, 1995, to

be implemented within 60 days

Amendment No.: 117
190 Federal Register / Vol. 61, No. 2 / Wednesday, January 3, 1996 / Notices

Facility Operating License No. NPF38. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: January 19, 1994 (59 FR 2868)

The May 12, August 9, and September

18, 1995, letters provided additional

information that did not change the

initial proposed no significant hazards

consideration determination.The

Commission’s related evaluation of the

amendment is contained in a Safety

Evaluation dated December 14, 1995.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:

September 7, 1993, as supplemented by

letters dated February 8, 1994, and

August 9, 1995.

Brief description of amendment: The

amendment revised the license

condition on physical security and

approves the revision to Physical

Security Plan for the Waterford Steam

Electric Station, Unit 3.

Date of issuance: December 19, 1995

Effective date: December 19, 1995

Amendment No.: 118

Facility Operating License No. NPF38. Amendment revised the license. The

additional information contained in the

supplemented letter dated August 9,

1995, was clarifying in nature and thus,

within the scope of the initial notice

and did not affect the staff’s proposed

no significant hazards consideration

determination.

Date of initial notice in Federal

Register: March 30, 1994 (59 FR 14887)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated December 19, 1995.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,

Docket Nos. 50-250 and 50-251, Turkey

Point Plant Units 3 and 4, Dade County,

Florida

Date of application for amendments:

July 26, 1995

Brief description of amendments: The

amendments consist of changes to the

Technical Specifications relating to

nuclear instrumentation system

adjustments based on calorimetric

measurements at reduced power

levels.Date of issuance: December 12,

1995

Effective date: December 12, 1995

Amendment Nos. 180 and 174Facility

Operating Licenses Nos. DPR-31 and

DPR-41: Amendments revised the

Technical Specifications.

Date of initial notice in Federal

Register: September 13, 1995 (60 FR

47617) The Commission’s related

evaluation of the amendments is

contained in a Safety Evaluation dated

December 12, 1995.No significant

hazards consideration comments

received: No

Local Public Document Room

location: Florida International

University, University Park, Miami,

Florida 33199.

Florida Power Corporation, et al.,

Docket No. 50-302, Crystal River Unit

No. 3 Nuclear Generating Plant, Citrus

County, Florida

Date of application for amendment:

January 26, 1995, as supplemented

March 9 and May 24, 1995

Brief description of amendment: This

amendment increases the allowable U235 enrichment of fuel to be stored in

the new and spent fuel storage facilities.

Date of issuance: December 15, 1995

Effective date: December 15, 1995

Amendment No.: 151

Facility Operating License No. DPR72. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: April 26, 1995 (60 FR 20517)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated December 15, 1995. No

significant hazards consideration

comments received: No.

Local Public Document Room

location: Coastal Region Library, 8619

W. Crystal Street, Crystal River, Florida

32629

Gulf States Utilities Company, Cajun

Electric Power Cooperative, and

Entergy Operations, Inc., Docket No.

50-458, River Bend Station, Unit 1,

West Feliciana Parish, Louisiana

Date of amendment request: October

24, 1995

Brief description of amendment: The

amendment revised the Technical

Specifications to reflect the approval for

the River Bend Station to use 10 CFR

Part 50, Appendix J, Option B for the

containment leak rate testing.

Date of issuance: December 19, 1995

Effective date: December 19, 1995

Amendment No.: 84

Facility Operating License No. NPF47. The amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: November 8, 1995 (60 FR

56368) The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

December 19, 1995.No significant

hazards consideration comments

received. No.

Local Public Document Room

location: Government Documents

Department, Louisiana State University,

Baton Rouge, LA 70803

Indiana Michigan Power Company,

Docket Nos. 50-315 and 50-316, Donald

C. Cook Nuclear Plant, Unit Nos. 1 and

2, Berrien County, Michigan

Date of application for amendments:

May 25, 1995 (AEP:NRC:1200B)

Brief description of amendments: The

amendments change the surveillance

frequency for the manual actuation

function for main steam line isolation

from monthly to quarterly and delete

obsolete footnotes associated with

previous surveillance interval

extensions from Unit 2 Table 4.3-2.

Date of issuance: December 13, 1995

Effective date: December 13, 1995,

with full implementation within 45

days

Amendment Nos.: 204 and 189

Facility Operating License Nos. DPR58 and DPR-74. Amendments revised

the Technical Specifications.

Date of initial notice in Federal

Register: July 5, 1995 (60 FR 35081)The

Commission’s related evaluation of the

amendments is contained in a Safety

Evaluation dated December 13, 1995. No

significant hazards consideration

comments received: No.

Local Public Document Room

location: Maud Preston Palenske

Memorial Library, 500 Market Street, St.

Joseph, Michigan 49085.

Maine Yankee Atomic Power Company,

Docket No. 50-309, Maine Yankee

Atomic Power Station, Lincoln County,

Maine

Date of application for amendment:

August 8, 1995

Brief description of amendment: This

amendment modifies the definitions of

Transthermal (Condition 4), Hot

Shutdown (Condition 5), and Hot

Standby (Condition 6) reactor operating

conditions. The Transthermal and Hot

Shutdown Conditions are modified to

establish an applicable range of

subcriticality and be consistent with

other Definitions. The wording of Hot

Standby is modified to remove reference

to control rod position, consistent with

NUREG-1432, Standard Technical

Specifications for Combustion

Engineering Plants, Revision 1, dated

April 1995.
Federal Register / Vol. 61, No. 2 / Wednesday, January 3, 1996 / Notices 191

Date of issuance: December 15, 1995

Effective date: As of the date of

issuance, to be implemented within 30

days.

Amendment No.: 154

Facility Operating License No. DPR36: Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: October 11, 1995 (60 FR

52931) The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

December 15, 1995.No significant

hazards consideration comments

received: No

Local Public Document Room

location: Wiscasset Public Library, High

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

04578.

Omaha Public Power District, Docket

No. 50-285, Fort Calhoun Station, Unit

No. 1, Washington County, Nebraska

Date of amendment request: June 27,

1995

Brief description of amendment: The

amendment revises Technical

Specification (TS) 2.2 on chemical and

volume control system (CVCS) to

reformat and clarify the requirements

and make them more consistent with

the requirements of the Combustion

Engineering Standard Technical

Specifications (STS), as presented in

NUREG-0212, Revision 2.

Date of issuance: December 12, 1995

Effective date: December 12, 1995

Amendment No.: 171

Facility Operating License No. DPR40. Amendment revised the Technical

Specifications.

Date of initial notice in Federal

Register: August 2, 1995 (60 FR 39447)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated December 12, 1995.No

significant hazards consideration

comments received: No.

Local Public Document Room

location: W. Dale Clark Library, 215

South 15th Street, Omaha, Nebraska

68102

Pennsylvania Power and Light

Company, Docket Nos. 50-387 and 50-

388 Susquehanna Steam Electric

Station, Units 1 and 2, Luzerne County,

Pennsylvania

Date of application for amendments:

February 10, 1995, as supplemented by

letter dated November 10, 1995.

Brief description of amendments:

These amendments (1) modify the

Susquehanna Steam Electric Station,

Unit 1 and 2 Technical Specifications to

extend the allowable out-of-service

times (AOTs) for maintenance and

repair and the surveillance test intervals

(STIs) between channel functional tests

for the following groups of instruments:

reactor protection systems

instrumentation (TS 3.3.1), isolation

actuation instrumentation (TS 3.3.2),

emergency core cooling system

actuation instrumentation (TS 3.3.3),

ATWS (anticipated transient without

scram) recirculation pump trip system

instrumentation (TS 3.3.4.1), end-ofcycle recirculation pump trip system

instrumentation (TS 3.3.4.2), reactor

core isolation cooling system (RCIC)

actuation instrumentation (TS 3.3.5),

control rod block instrumentation (TS

3.3.6), radiation monitoring

instrumentation (TS 3.3.7.1), and

feedwater/main turbine trip system

actuation instrumentation (TS 3.3.90);

(2) change the required actions and

AOTs for the instruments listed above to

make requirements consistent with

supporting analysis in General Electric

topical reports and change additional

actions required to prevent extended

AOTs from resulting in extended loss of

instrument function; (3) change the

required actions and AOTs for the

instruments listed above for

instrumentation associated with the

ADS (automatic depressurization

system), recirculation pump trip, and

pump suction lineup for HPCI (high

pressure core injection) and RCIC; (4)

change applicability requirements and

required actions for the reactor vessel

water level-low, level 3 function that

isolates the RHR (residual heat removal)

system shutdown cooling system so that

the function is required to be operable

in operational conditions 3,4, and 5 to

prevent inadvertent loss of reactor

coolant via the RHR shutdown cooling

system; (5) remove notes in Table 3.3.2-

1, 3.3.2-2, and 4.3.1-1 related to

maintenance on leak detection

temperature detectors and remove the

note toTS 3.3.6 for Unit 1 related to a

previous relief from TS 3.0.4; and (6)

reformat, renumber, and/or reword

existing requirements to incorporate the

changes listed above.

Date of issuance: December 18, 1995

Effective date: As of date of issuance

and to be implemented within 30 days.

Amendment Nos.: 155 and 126

Facility Operating License Nos. NPF14 and NPF-22. The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: March 29, 1995 (60 FR 16194)

The supplemental letter provided

corrected TSs and did not change the

original proposed no significant hazards

consideration nor the Federal Register

notice.The Commission’s related

evaluation of the amendments is

contained in a Safety Evaluation dated

December 18, 1995No significant

hazards consideration comments

received: No

Local Public Document Room

location: Osterhout Free Library,

Reference Department, 71 South

Franklin Street, Wilkes-Barre,

Pennsylvania 18701.

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:

September 26, 1995

Brief description of amendments: The

amendments change the containment air

lock door seal leakage rate from ‘‘no

detectable seal leakage’’ to ‘‘less than or

equal to 0.01 La’’ when the gap between

the door seals is pressurized to greater

than or equal to 10 psig for a period of

not less than 15 minutes.

Date of issuance: December 8, 1995

Effective date: As of the date of

issuance to be implemented within 30

days

Amendment Nos.: 118 and 109

Facility Operating License Nos. NPF2 and NPF-8. Amendments revise the

Technical Specifications.

Date of initial notice in Federal

Register: November 8, 1995 (60 FR

56370) The Commission’s related

evaluation of the amendments is

contained in a Safety Evaluation dated

December 8, 1995.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

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:

August 7, 1995 (TS 95-17)

Brief description of amendments: The

changes relocate the heat flux hot

channel factor penalty from

Surveillance Requirement 4.2.2.2.e.1 to

the Core Operating Limits Report and

replace the methodology (WCAP-10216-

P-A) listed in Technical Specification

6.9.1.14.a.2 with WCAP-10216-P-A,

Revision 1A.

Date of issuance: December 11, 1995

Effective date: December 11, 1995

Amendment Nos.: 216 and 206

Facility Operating License Nos. DPR77 and DPR-79: Amendments revise the

technical specifications.

Date of initial notice in Federal

Register: August 30, 1995 (60 FR 45186)

The Commission’s related evaluation of

the amendment is contained in a Safety
192 Federal Register / Vol. 61, No. 2 / Wednesday, January 3, 1996 / Notices

Evaluation dated December 11, 1995.No

significant hazards consideration

comments received: None

Local Public Document Room

location: Chattanooga-Hamilton County

Library, 1101 Broad Street, Chattanooga,

Tennessee 37402No significant hazards

consideration comments received: None

TU Electric Company, Docket Nos. 50-

445 and 50-446, Comanche Peak Steam

Electric Station, Unit Nos. 1 and 2,

Somervell County, Texas

Date of amendment request: August

15, 1995 (TXX-95215)

Brief description of amendments:

These changes relocated the Shutdown

Margin limits from the Technical

Specifications (TSs) to the Core

Operating Limits Report (COLR). The

changes were consistent with the intent

of Generic Letter 88-16 which provides

guidelines for the removal of cyclespecific parameter limits from the TSs.

Date of issuance: December 15, 1995

Effective date: December 15, 1995

Amendment Nos.: Unit 1 -

Amendment No. 44; Unit 2 -

Amendment No. 30

Facility Operating License Nos. NPF87 and NPF-89. The amendments

revised the Technical Specifications.

Date of initial notice in Federal

Register: October 11, 1995 (60 FR

52935) The Commission’s related

evaluation of the amendments is

contained in a Safety Evaluation dated

December 15, 1995.No significant

hazards consideration comments

received: No

Local Public Document Room

Location: University of Texas at

Arlington Library, Government

Publications/Maps, 702 College, P.O.

Box 19497, Arlington, TX 76019

Union Electric Company, Docket No.

50-483, Callaway Plant, Callaway

County, Missouri

Date of amendment request: April 26,

1995

Brief description of amendment: The

amendment revises Technical

Specification (TS) 3/4.7.6 to reduce the

upper limit on the flow rate through the

control room filtration subsystem and

adopts ASTM D-3803-1989 as the

laboratory testing standard for control

room filtration and control building

pressurization charcoal adsorber. The

amendment also revises the Bases for TS

3/4.7.6 to reflect the changes.

Date of issuance: December 20, 1995

Effective date: December 20, 1995, to

be implemented within 30 days from

the date of issuance.

Amendment No.: 106

Facility Operating License No. NPF30. The amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: May 23, 1995 (60 FR 27345)

The Commission’s related evaluation of

the amendment is contained in a Safety

Evaluation dated December 20, 1995.No

significant hazards consideration

comments received: No.

Local Public Document Room

location: Callaway County Public

Library, 710 Court Street, Fulton,

Missouri 65251.

Wolf Creek Nuclear Operating

Corporation, Docket No. 50-482, Wolf

Creek Generating Station, Coffey

County, Kansas

Date of amendment request: June 14,

1995, as supplemented by letters dated

July 13, 1995, and August 22,

1995.I11Brief description of

amendment: The amendment revises

Technical Specification (TS) 3.2.3,

‘‘Nuclear Enthalpy Rise Hot Channel

Factor,’’ TS 6.9.1.9, ‘‘Core Operating

Limits Report,’’ and the associated Bases

sections. The revisions incorporate

changes associated with the planned

implementation of advanced nuclear

and core thermal-hydraulic design

methodologies licensed from

Westinghouse Electric Corporation for

core reload design, starting with Cycle

9.

Date of issuance: December 8, 1995

Effective date: December 8, 1995, to

be implemented prior to restart from the

eighth refueling outage, which is

scheduled to begin in March 1996.

Amendment No.: 92

Facility Operating License No. NPF42. The amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: August 2, 1995 (60 FR 39456)

The August 22, 1995, supplemental

letter forwarded the nonproprietary

version of Wolf Creek Nuclear Operating

Corporation’s safety evaluation and

analysis provided in the June 14, 1995,

submittal and did not change the staff’s

original no significant hazards

determination.The Commission’s

related evaluation of the amendment is

contained in a Safety Evaluation dated

December 8, 1995.No significant

hazards consideration comments

received: No.

Local Public Document Room

locations: Emporia State University,

William Allen White Library, 1200

Commercial Street, Emporia, Kansas

66801 and Washburn University School

of Law Library, Topeka, Kansas 66621

Wolf Creek Nuclear Operating

Corporation, Docket No. 50-482, Wolf

Creek Generating Station, Coffey

County, Kansas

Date of amendment request: August

22, 1995

Brief description of amendment: The

amendment revises the requirements of

Technical Specification (TS) 3.3.1 and

TS 3.3.2 and relocate Tables 3.3-2 and

3.3-5 and applicable Bases, which

provide the response time limits for the

reactor trip system (RTS) and the

engineered safety features actuation

system (ESFAS) instruments, from the

TS to the Updated Safety Analysis

Report (USAR). The licensee has stated

that the next USAR change request will

include these changes.

Date of issuance: December 12, 1995

Effective date: December 12, 1995, to

be implemented within 60 days of

issuance.

Amendment No.: 93

Facility Operating License No. NPF42. The amendment revised the

Technical Specifications.

Date of initial notice in Federal

Register: September 27, 1995 (60 FR

49950) The Commission’s related

evaluation of the amendment is

contained in a Safety Evaluation dated

December 12, 1995.No significant

hazards consideration comments

received: No.

Local Public Document Room

locations: Emporia State University,

William Allen White Library, 1200

Commercial Street, Emporia, Kansas

66801 and Washburn University School

of Law Library, Topeka, Kansas 66621

Dated at Rockville, Maryland, this

21st Day of December 1995.

For the Nuclear Regulatory Commission

Steven A. Varga,

Director, Division of Reactor Projects - I/II

Office of Nuclear Reactor Regulation

[Doc. 96–1 Filed 1–2–96; 8:45 am]

BILLING CODE 7590–O1–F

[Docket Nos. 50–295 and 50–304]

Commonwealth Edison Company,

(Zion Nuclear Power Station, Unit Nos.

1 and 2); Exemption

I

The Commonwealth Edison Company

(ComEd, the licensee) is the holder of

Facility Operating License Nos. DPR–39

and DPR–48, which authorize operation

of the Zion Nuclear Power Station,

Units 1 and 2 (the facilities). The

licenses provide, among other things,

that the facilities are subject to all the

rules, regulations, and orders of the U.S.

Nuclear Regulatory Commission (the

Commission) now or hereafter in effect.