Biweekly Notice Involving No Significant Hazards Considerations; Applications and Amendments to Facility Operating Licenses

Federal RegisterFeb 14, 1996

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NUCLEAR REGULATORY COMMISSION

Biweekly Notice Involving No Significant Hazards Considerations;

Applications and Amendments to Facility Operating Licenses

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 January 22, 1996, through February 2, 1996.

The last biweekly notice was published on January 31, 1996 (61 FR

3497).

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 March 15, 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

[[Page 5810]]

which is available at the Commission's Public Document Room, the Gelman

Building, 2120 L Street, NW., Washington, DC and at the local public

document room for the particular facility involved. If a request for a

hearing or petition for leave to intervene is filed by the above date,

the Commission or an Atomic Safety and Licensing Board, designated by

the Commission or by the Chairman of the Atomic Safety and Licensing

Board Panel, will rule on the request and/or petition; and the

Secretary or the designated Atomic Safety and Licensing Board will

issue a notice of a hearing or an appropriate order.

As required by 10 CFR 2.714, a petition for leave to intervene

shall set forth with particularity the interest of the petitioner in

the proceeding, and how that interest may be affected by the results of

the proceeding. The petition should specifically explain the reasons

why intervention should be permitted with particular reference to the

following factors: (1) the nature of the petitioner's right under the

Act to be made a party to the proceeding; (2) the nature and extent of

the petitioner's property, financial, or other interest in the

proceeding; and (3) the possible effect of any order which may be

entered in the proceeding on the petitioner's interest. The petition

should also identify the specific aspect(s) of the subject matter of

the proceeding as to which petitioner wishes to intervene. Any person

who has filed a petition for leave to intervene or who has been

admitted as a party may amend the petition without requesting leave of

the Board up to 15 days prior to the first prehearing conference

scheduled in the proceeding, but such an amended petition must satisfy

the specificity requirements described above.

Not later than 15 days prior to the first prehearing conference

scheduled in the proceeding, a petitioner shall file a supplement to

the petition to intervene which must include a list of the contentions

which are sought to be litigated in the matter. Each contention must

consist of a specific statement of the issue of law or fact to be

raised or controverted. In addition, the petitioner shall provide a

brief explanation of the bases of the contention and a concise

statement of the alleged facts or expert opinion which support the

contention and on which the petitioner intends to rely in proving the

contention at the hearing. The petitioner must also provide references

to those specific sources and documents of which the petitioner is

aware and on which the petitioner intends to rely to establish those

facts or expert opinion. Petitioner must provide sufficient information

to show that a genuine dispute exists with the applicant on a material

issue of law or fact. Contentions shall be limited to matters within

the scope of the amendment under consideration. The contention must be

one which, if proven, would entitle the petitioner to relief. A

petitioner who fails to file such a supplement which satisfies these

requirements with respect to at least one contention will not be

permitted to participate as a party.

Those permitted to intervene become parties to the proceeding,

subject to any limitations in the order granting leave to intervene,

and have the opportunity to participate fully in the conduct of the

hearing, including the opportunity to present evidence and cross-

examine witnesses.

If a hearing is requested, the Commission will make a final

determination on the issue of no significant hazards consideration. The

final determination will serve to decide when the hearing is held.

If the final determination is that the amendment request involves

no significant hazards consideration, the Commission may issue the

amendment and make it immediately effective, notwithstanding the

request for a hearing. Any hearing held would take place after issuance

of the amendment.

If the final determination is that the amendment request involves a

significant hazards consideration, any hearing held would take place

before the issuance of any amendment.

A request for a hearing or a petition for leave to intervene must

be filed with the Secretary of the Commission, U.S. Nuclear Regulatory

Commission, Washington, DC 20555, 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: January 16, 1996.

Description of amendments request: The proposed amendments would

revise the Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2,

Technical Specifications (TSs) to adopt Option B of 10 CFR Part 50,

Appendix J, to require Type A containment leak rate tests to be

performed on a performance-based testing schedule. Specifically, TSs 3/

4.6.1.2 and 4.6.1.6.3 will be revised to reference a new Containment

Leakage Rate Testing Program, TS 6.0 will be revised to add the new

Containment Leakage Rate Testing Program, identify the programmatic

controls for the new program, and reference the source of the

programmatic guidelines, Regulatory Guide 1.116, ``Performance-Based

Containment Leak-Test Programs,'' dated September 1995. The TS Bases

will be revised to reflect these changes.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Would not involve a significant increase in the probability

or consequences of an accident previously evaluated.

Containment leakage rate testing is performed in accordance with

10 CFR Part 50, Appendix J, ``Primary Reactor Containment Leakage

Testing for Water-Cooled Power Reactors.'' The Appendix J

containment leakage test requirements include performance of Type A

tests, which measure the overall leakage rate of the containment,

and Type B and C tests, which measure the leakage through

containment penetrations and valves. The Commission has amended the

regulations to provide a

[[Page 5811]]

performance-based alternative, Option B, to the existing Appendix J. At

this time, Baltimore Gas and Electric Company plans to adopt Option

B for Type A testing only.

Implementation of Option B involves no physical or operational

changes to the plant structures, systems or components. Furthermore,

leakage rate testing and containment surface visual inspections do

not contribute to the initiation of any postulated accidents;

therefore, this proposed change does not involve an increase in the

probability of any previously evaluated accidents.

Type A testing is necessary to demonstrate that leakage through

the containment is within the limits assumed in the accident

analyses. The only potential effect of the proposed change to the

Type A test frequency is the possibility that containment leakage

would go undetected between tests. As described in NUREG-1493,

passive failures resulting in containment leakage in excess of that

assumed in the accident analyses are extremely unlikely to develop

between Type A tests. Additionally, the Calvert Cliffs Individual

Plant Examination considered the phenomenological effects associated

with severe accidents which could lead to containment failure. It

was concluded that adopting a performance-based testing interval

will not significantly affect the containment failure probabilities

calculated for the Individual Plant Examination. Furthermore, the

required frequency for containment surface examinations to identify

containment degradation precursors will be relocated from the

Technical Specifications to the Containment Leakage Rate Testing

Program, but will remain at three examinations every ten years as

recommended by Regulatory Guide 1.163, September 1995. Altogether,

adoption of a performance-based testing frequency, as specified in

10 CFR Part 50, Appendix J, Option B, will not significantly

decrease the confidence in the leak-tightness of the containment.

Therefore, this change will not result in a significant increase in

the probability of undetected containment degradation or in 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 Technical Specification change adopts a

performance-based approach to containment leakage rate testing. 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. The proposed change does not

affect normal plant operations or configuration, nor does it affect

leakage rate test methods. As the proposed change would not change

the design, configuration or operation of the plant, it could not

cause containment leakage rate testing to become an accident

initiator.

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 purpose of the existing schedule for Type A tests is to

ensure that the release of radioactive material will be restricted

to those leak paths and leakage rates assumed in the accident

analyses. The margin of safety associated with containment leakage

rate is not reduced if containment leakage does not exceed the

maximum allowable leakage rate defined in the Technical

Specifications. The proposed Technical Specification change

implements a performance-based Type A testing option, but does not

affect the maximum allowable containment leakage rate. The proposed

change does not affect a safety limit, a Limiting Condition for

Operation, or the way in which the plant is operated.

In NUREG-1493, the Commission included a sensitivity study to

explore the risk affect of several alternate leakage rate testing

schedules. This study concludes that decreasing the Type A testing

frequency to one test per twenty years would ``lead to an

imperceptible increase in risk.'' Additionally, it was determined

that implementation of the performance-based testing option will not

significantly affect the containment failure probability calculated

in the Calvert Cliffs Individual Plant Examination. Based upon these

studies, there is sufficient information to conclude that the risk

increase, and that the probability of exceeding the maximum

allowable containment leakage rate as a result of adopting Option B,

is low.

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.

Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam

Electric Plant, Unit No. 2, Darlington County, South Carolina

Date of amendment request: November 27, 1995.

Description of amendment request: The proposed change would revise

technical specification (TS) section 3.2 to remove requirements for the

chemical and volume control system (CVCS). The CVCS requirements would

be relocated to a licensee-controlled document and controlled by the 10

CFR 50.59 evaluation process.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The proposed change removes the Chemical and Volume Control

System (CVCS) requirements from the Technical Specifications (TS)

and relocates these requirement[s] to a licensee-controlled

document. As such, the proposed change only affects plant

documentation and does not change the operating requirements or the

plant physical or operating configuration. The CVCS requirements

will be controlled by the plant approved process for the licensee-

controlled document using the 10 CFR 50.59 evaluation process. The

proposed change relocating the CVCS requirements from the TS to

licensee control will not affect the probability of an accident

previously evaluated because the operating restrictions will remain

in effect and any change to the operating restrictions will be

performed in accordance with 10 CFR 50.59.

Examination of the H. B. Robinson Steam Electric Plant, Unit No.

2 Updated Final Safety Analysis Report (UFSAR) Chapter 15, Accident

Analysis, finds that no CVCS structure, system, or component

functions or actuates to mitigate a design basis accident or

transient. Valves at the CVCS to Reactor Coolant System (RCS)

interface perform a containment isolation function. However, the TS

Section 3.2 does not address the containment isolation aspect of the

CVCS. As such, the proposed change to remove the CVCS requirements

from the TS will not affect the consequences of an accident

previously evaluated.

2. The proposed change does not create the possibility of a new

or different kind of accident from any accident previously

evaluated.

The proposed change removes the CVCS requirements from the TS

and relocates the requirements to a licensee-controlled document. As

such, the proposed change only affects plant documentation and does

not change the operating requirements or the plant physical or

operating configuration. The CVCS requirements will be controlled by

the plant approved process for the licensee-controlled document

using the 10 CFR 50.59 evaluation process. The proposed change will

not create the possibility of a new or different kind of accident

from any accident previously evaluated because any future change to

these operating restrictions will be performed in accordance with 10

CFR 50.59.

3. The proposed change does not involve a significant reduction

in the margin of safety.

The proposed change removes the CVCS requirements from the TS

based on the criteria of 10 CFR 50.36(c)(2)(ii). The CVCS

requirements will be relocated to a licensee-

[[Page 5812]]

controlled document. As such, the proposed change only affects plant

documentation and does not change operating requirements or the

plant physical or operating configuration. The CVCS requirements

will be controlled by the plant approved process for the licensee-

controlled document using the 10 CFR 50.59 evaluation process. The

proposed change will not result in any reduction in the margin of

safety because any future change to the CVCS operating restrictions

will be performed in accordance with 10 CFR 50.59. Therefore, the

proposed change does not involve a significant reduction in a margin

of safety.

The NRC staff has reviewed the licensee's analysis and, based on

this review, it appears that the three standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendment request involves no significant hazards consideration.

Local Public Document Room location: Hartsville Memorial Library,

147 West College Avenue, Hartsville, South Carolina 29550.

Attorney for licensee: R. E. Jones, General Counsel, Carolina Power

& Light Company, Post Office Box 1551, Raleigh, North Carolina 27602.

NRC Project Director: David B. Matthews.

Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam

Electric Plant, Unit No. 2, Darlington County, South Carolina

Date of amendment request: December 10, 1995.

Description of amendment request: The proposed change would revise

technical specification (TS) section 3.5.1 and Tables 3.5-2, 3, and 4

concerning the reactor trip system (RTS), engineered safety feature

actuation system (ESFAS), and isolation function. TS would be revised

to (1) specify actions to be taken when an instrument channel becomes

inoperable, (2) add an ``Applicable Conditions'' column that defines

the applicability and/or mode of operation of each functional unit, and

(3) make editorial enhancements.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The proposed change to upgrade the RTS and ESFAS TS to more

closely agree with Westinghouse Standard TS (i.e., NUREG-0452) will

not result in any hardware changes. The RTS and ESFAS are not

assumed to be initiators of analyzed events.

The role of these systems is in mitigating and thereby limiting

the consequences of accidents. The proposed changes will ensure the

RTS and ESFAS remain capable of mitigating design basis events as

described in the Updated Final Safety Analysis Report (UFSAR) and

that the results of the analyses in the UFSAR remain bounding.

Additionally, the proposed changes do not impose any new safety

analyses limits or alter the plant's ability to detect and mitigate

events. Therefore, this change does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

2. The proposed change does not create the possibility of a new

or different kind of accident from any accident previously

evaluated.

The proposed change to upgrade the RTS and ESFAS TS to more

closely agree with Westinghouse Standard TS (i.e., NUREG-0452) does

not necessitate a physical alteration of the plant (i.e., no new or

different type of equipment will be installed) or changes in

parameters governing normal plant operation. Thus, the proposed

change does not create the possibility of a new or different kind of

accident from any accident previously evaluated.

3. The proposed change does not involve a significant reduction

in the margin of safety.

The proposed change, which upgrades the RTS and ESFAS TS to be

consistent with Westinghouse Standard TS (i.e., NUREG-0452) does not

involve a significant reduction in a margin of safety. The proposed

change has been developed to ensure the analyzed safety limits are

not exceeded and ensures the RTS and ESFAS are available when

necessary to mitigate the consequences of accidents. It also imposes

additional requirements to ensure the RTS and ESFAS remain capable

of mitigating the consequences of design basis accidents as

described in the UFSAR accident analyses. In addition, this change

provides a benefit of avoiding unnecessary plant transients when

adequate compensatory measures are available to ensure the intended

function of the instrumentation is satisfied.

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: Hartsville Memorial Library,

147 West College Avenue, Hartsville, South Carolina 29550.

Attorney for licensee: R. E. Jones, General Counsel, Carolina Power

& Light Company, Post Office Box 1551, Raleigh, North Carolina 27602.

NRC Project Director: David B. Matthews.

Connecticut Yankee Atomic Power Company, Docket No. 50-213, Haddam Neck

Plant, Middlesex County, Connecticut, and Northeast Nuclear Energy

Company, et al., Docket Nos. 50-245, 50-336, and 50-423, Millstone

Nuclear Power Station, Unit Nos. 1, 2, and 3, New London County,

Connecticut

Date of amendment request: November 22, 1995.

Description of amendment request: The amendments would revise the

Technical Specifications (TS) for Haddam Neck and Millstone Unit Nos.

1, 2, and 3 to be consistent with the guidance of Generic Letter 93-07.

The proposed changes will remove review of the emergency and security

plans from the TS list of responsibilities of the Plant Operations

Review Committee (PORC)/Site Operations Review Committee (SORC), and

will also remove the requirement for PORC/SORC to review procedures and

procedure changes necessary for the implementation of the emergency and

security plans.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

* * * The proposed changes do not involve an SHC [significant

hazards consideration] because the changes would not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed changes delete the technical specification

requirement to review the emergency plans, security plans, and their

implementing procedures by PORC/SORC. The requirement which mandates

PORC/SORC review will be maintained in the respective emergency plan

and security plan. These changes are purely administrative in

nature. These changes do not affect the configuration, operation, or

performance of any system, structure, or component. The proposed

changes are therefore not relevant to the probability of initiation

of any accident previously evaluated, and they are not related to

the prevention or mitigation of any accident previously evaluated.

Thus they do not increase the consequences of any design basis

accident.

Therefore, these proposed changes to the Technical

Specifications do not involve a significant increase in the

probability or consequences of an accident previously evaluated.

2. Create the possibility of a new or different kind of accident

from any accident previously evaluated.

The proposed changes delete from the technical specifications

the line item requiring the review of emergency plans, security

plans, and their implementing

[[Page 5813]]

procedures by PORC/SORC. Revisions to these plans will continue to be

reviewed by PORC/SORC due to commitments to contain the requirement

for PORC/SORC review in the emergency plan and security plan. These

changes are purely administrative in nature.

None of the proposed changes described above alter the

configuration, normal operation, design bases, function, or

performance of any components or systems. Thus, the proposed

administrative changes do not create the possibility of a new or

different kind of accident from any previously evaluated since these

changes do not introduce any new or different equipment, operating

mode, or design basis functions for the existing licensed

structures, systems and components. Thus, the proposed changes do

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.

None of the above proposed changes alter the configuration,

normal operation, design bases, function, or performance of any

components or systems. Therefore, the proposed changes do not affect

the margin of safety inherent in the design, analysis, function, or

operation of the relevant structures, systems or components.

These proposed changes do not alter the fuel clad barrier, fuel

integrity, reactor coolant system integrity or the containment

boundary integrity; thus no margin of safety related to these

barriers is involved.

None of the proposed administrative changes described above

alter the configuration, normal operation, design bases, function or

performance of any components, systems, or barriers to a

radiological release. Thus, the proposed administrative changes do

not affect the margin of safety inherent in the design, analysis,

function, or operation of the relevant structures, systems or

components.

Based on the above, these proposed changes do not involve a

significant reduction in a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on

this review, it appears that the three standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendment request involves no significant hazards consideration.

Local Public Document Room location: Russell Library, 123 Broad

Street, Middletown, CT 06457 for the Haddam Neck Plant, and Learning

Resources Center, Three Rivers Community-Technical College, 574 New

London Turnpike, Norwich, CT 06360 for Millstone Units 1, 2, and 3.

Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear

Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,

CT 06141-0270.

NRC Project Director: Phillip F. McKee.

Duke Power Company, et al., Docket Nos. 50-413 and 50-414, Catawba

Nuclear Station, Units 1 and 2, York County, South Carolina

Date of amendment request: January 11, 1996.

Description of amendment request: The Catawba Unit 1 and the

Catawba Unit 2 containment process penetration M308 and associated

containment isolation valves are currently not in service and serve no

function other than providing containment integrity. The licensee plans

to implement modifications for both units to remove containment

isolation valves RN-429A and RN-432B of penetration M308, remove

associated wiring and control room instrumentation, and cut and cap

tubing providing containment valve injection water to these containment

isolation valves during the forthcoming Unit 1 refueling outage,

currently scheduled to begin by June 1996, and the Unit 2 refueling

outage currently scheduled to begin in March 1997. The proposed

Technical Specifications (TS) would be revised to delete these

containment isolation valves and associated equipment to permit

implementation of these modifications. The licensee's requested

amendment removes process penetration M308 from TS Table 3.6-1 and

removes containment isolation valves RN-429A and RN-432B from TS Table

3.6-2a and Table 3.6-2b due to planned modifications which physically

remove these valves from process penetration M308.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

Criterion 1

The physical removal of containment isolation valves RN-432B and

RN-429A, associated control room instrumentation, containment valve

injection water connections to these valves and the subsequent

sealing of process penetration M308 will decrease unnecessary

challenges to containment isolation, containment valve injection

water leak-rate testing and the condition of control room

instrumentation, as opposed to the current configuration.

Since the sealing of process penetration M308 will be performed

per the requirements of the applicable ASME code piping safety class

requirements, the confidence in the pressure boundary will be

equivalent to the component as originally designed. Therefore, this

Technical Specification amendment to remove process penetration M308

from Technical Specification Table[] 3.6-1 and to remove containment

isolation valves RN-429A and 432B from Technical Specification Table

3.6-2a and Table 3.6-2b will not increase the probability or

consequences of an accident that has been previously evaluated.

Criterion 2

Since no new failure modes are created, on the basis that the

penetration is equivalent in confidence to the original design, and

the plant will operate the same way it does now, this Technical

Specification amendment to remove process penetration M308 from

Technical Specification Table[] 3.6-1 and to remove containment

isolation valves RN-429A and 432B from Technical Specification Table

3.6-2a and Table 3.6-2b does not create the possibility of a new or

different kind of accident from any accident previously evaluated.

Criterion 3

This proposed change to Technical Specifications will not cause

a significant reduction in the margin of safety. Upon completion of

the removal of containment isolation valves RN-432B and 429A and the

subsequent sealing of process penetration M308, the penetration will

be Type B leak rate tested as part of post-modification testing, and

will be retested periodically and following each use of the

penetration for temporary containment cooling purposes during

refueling outages. Therefore, the fuel, cladding, reactor coolant

pressure boundary, and containment are not negatively affected by

the proposed Technical Specification amendment. No assumptions made

in any accident analysis are compromised by this proposed Technical

Specification amendment.

The NRC staff has reviewed the licensee's analysis and, based on

this review, it appears that the three standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendment request involves no significant hazards consideration.

Local Public Document Room location: York County Library, 138 East

Black Street, Rock Hill, South Carolina 29730.

Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422

South Church Street, Charlotte, North Carolina 28242.

NRC Project Director: Herbert N. Berkow.

Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389,

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

Date of amendment request: January 4, 1996.

Description of amendment request: The proposed revisions rectify a

discrepancy in Specification 3.5.3 for each St. Lucie unit, and provide

assurance that administrative controls for High Pressure Safety

Injection pumps remain effective in the lower operational modes.

[[Page 5814]]

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

(1) Operation of the facility in accordance with the proposed

amendment would not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The amendment proposed for each St. Lucie Unit (1 and 2)

rectifies an error in the Applicability statement for Technical

Specification 3.5.3, which provides limiting conditions for

operation (LCO) for the Emergency Core Cooling System (ECCS)

subsystems during plant shutdown. The revision is administrative in

nature and does not change the technical requirements within the LCO

that are established to assure a minimum functional capability

required of the ECCS systems to mitigate analyzed transients.

Rather, the revision provides assurance that the effectiveness of

certain administrative controls, established to restrict the number

of operable HPSI [High Pressure Safety Injection] pumps during

shutdown, will not be diminished by a misinterpretation of the modes

and conditions for which the LCO must apply.

This proposal does not create any accident initiators, nor does

it change the availability or method of operation of equipment that

is assumed to function in the success path(s) for mitigating

accidents evaluated in the plant safety analyses. Therefore,

operation of either facility in accordance with its proposed

amendment would not involve a significant increase in the

probability or consequences of an accident previously evaluated.

(2) Operation of the facility in accordance with the proposed

amendment would not create the possibility of a new or different

kind of accident from any accident previously evaluated.

The proposed administrative change to the LCO 3.5.3

Applicability statement for each St. Lucie unit will not change the

physical plant or the modes of plant operation defined in the

Facility License. The revision does not involve the addition or

modification of equipment, nor does it alter the design or operation

of plant systems. Therefore, operation of either facility in

accordance with its proposed amendment would not create the

possibility of a new or different kind of accident from any accident

previously evaluated.

(3) Operation of the facility in accordance with the proposed

amendment would not involve a significant reduction in a margin of

safety.

The proposed amendment involves an administrative change to LCO

3.5.3 for each St. Lucie unit, which applies to the ECCS subsystems

during the plant shutdown modes. The revision rectifies a

discrepancy in the Applicability statement, and thereby provides

assurance that the effectiveness of administrative controls

established within the LCO to limit the number of operable High

Pressure Safety Injection pumps during the shutdown modes will not

be diminished. The changes do not alter the basis for any technical

specification that is related to the establishment of, or the

maintenance of, a nuclear safety margin. Therefore, operation of

either facility in accordance with its proposed amendment would not

involve a significant reduction in a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on

this review, it appears that the three standards of 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendment request involves no significant hazards consideration.

Local Public Document Room location: Indian River Junior College

Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003.

Attorney for licensee: Harold F. Reis, Esquire, Newman and

Holtzinger, 1615 L Street, NW., Washington, DC 20036.

NRC Project Director: David B. Matthews.

IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy Center, Linn

County, Iowa

Date of amendment request: January 18, 1996.

Description of amendment request: The proposed amendment would

lower the Reactor Water Cleanup (RWCU) isolation setpoint from reactor

low level to reactor low-low level.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

(1) The proposed [technical specification] TS amendment will not

significantly increase the probability or consequences of any

previously evaluated accidents. The RWCU vessel level isolation

occurs as a result of a [loss-of-coolant-accident] LOCA and

therefore does not affect the probability of occurrence of a LOCA or

any other previously evaluated accident.

An IES calculation demonstrates that for all RWCU breaks or

cracks considered, high ambient temperature, high differential

temperature and/or high differential flow will provide the RWCU

isolation signal prior to reaching reactor low level. Therefore, the

level setpoint acts as a backup isolation signal for a break in RWCU

piping outside primary containment.

As discussed, this change will utilize four existing reactor

level sensors. These reactor level sensors are safety related and

located in the same physical area and in the same configuration as

the four existing sensors. Therefore, the reliability of the RWCU

vessel level isolation capability is not reduced.

(2) The proposed changes will not create the possibility of a

new or different kind of accident. The configuration of the RWCU

isolation valves is unchanged. As before, the failure of any single

active component in the new logic results in, at worst, failure of

one containment isolation valve to close. Because the closure of one

of the two valves is sufficient to achieve the containment

isolation, the possibility of an accident of a different type is not

increased.

The modification to the RWCU vessel level isolation logic has

been designed to the same standards as the original logic. This

change will require the same surveillance requirements for the

reactor low-low level trip point circuitry that are currently

required for the reactor low level trip point circuitry. All other

RWCU isolation functions remain unchanged. Consequently, no new

accidents are postulated as a result of this proposed change.

(3) The proposed change will not result in a significant

reduction in any margin of safety. No margin of safety is affected

by this change. The RWCU vessel level isolation occurs to establish

primary containment and limit fluid loss. The proposed change will

preserve these functions.

It can be noted, however, that for a RWCU piping break outside

primary containment, high ambient temperature, high differential

temperature and/or high differential flow will provide the RWCU

isolation signal. In the unlikely event that these temperature and

flow sensing devices fail, isolation will be initiated upon reactor

level reaching 119.5'' above [top of active fuel] TAF. Using

blowdown rates and valve closure times, analysis shows reactor level

will not drop below 105'' above TAF. The is well above the TAF.

Additionally, lowering the RWCU isolation setpoint does not increase

the consequences of a LOCA.

The NRC staff has reviewed the licensee's analysis and, based on

this review, it appears that the three standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendment request involves no significant hazards consideration.

Local Public Document Room location: Cedar Rapids Public Library,

500 First Street, S.E., Cedar Rapids, Iowa 52401.

Attorney for licensee: Jack Newman, Kathleen H. Shea, Morgan,

Lewis, & Bockius, 1800 M Street, NW., Washington, DC 20036-5869.

NRC Project Director: Gail H. Marcus.

IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy Center, Linn

County, Iowa

Date of amendment request: January 30, 1996.

Description of amendment request: The proposed amendment would

revise certain control rod scram insertion time testing limits. The

proposed change is compatible with the limits specified in the Improved

Standard Technical Specifications (ITS), NUREG 1433, Revision 1,

``Standard Technical Specifications, General Electric Plants, BWR/4.''

[[Page 5815]]

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

(1) The proposed amendment does not involve a change in the

probability or consequences of an accident previously evaluated. The

amount of reactivity inserted at rod position 46 (approximately 5%

of rod insertion) is small and the time required to insert this

amount of reactivity is not explicitly considered in the plant

transient analysis. A generic BWR/2-5 study (Reference 3 [EAS-56-

0889, ``BWR/2-5 Scram Time Technical Specification'', dated August

1989]) performed on behalf of the [boiling water reactor] BWR

Owner's Group to support the ITS demonstrated that relaxing the 5%

rod insertion time requirement had a negligible impact on plant

transient performance provided the insertion time requirements to

the other rod positions are met. We have confirmed that this study

is applicable to the [Duane Arnold Energy Center] DAEC. Increasing

the allowable average scram insertion time to rod position 46 for

all Operable control rods in addition to increasing the allowable

average scram insertion time to rod position 46 for the three

fastest control rods in any 2X2 array would still demonstrate that

the [control rod drive] CRD system will perform its intended

function. Scram time is a measure of CRD performance for

operability. As such, it is not the initiator of any plant event.

Therefore, the proposed change will not result in an increase in the

probability of an accident occurring.

(2) The amount of reactivity inserted at rod position 46

(approximately 5% of rod insertion) is small and the time required

to insert this amount of reactivity is not explicitly considered in

the transient analysis. A generic BWR/2-5 study showed that relaxing

the 5% rod insertion time requirement had a negligible impact on

plant transient performance. Increasing the allowable average scram

insertion time to rod position 46 for all Operable control rods,

while increasing the allowable average scram insertion time to rod

position 46 for the three fastest control rods in any 2X2 array,

would still demonstrate that the CRD system will perform its

intended function. Therefore, increasing the limits proposed does

not create the possibility of a new or different kind of accident

from any previously evaluated. Scram time is a measure of CRD

performance for operability. As such, it is not the initiator of any

plant event.

(3) The safety limit most affected by an increase in scram times

is the Minimum Critical Power Ratio (MCPR). The DAEC [technical

specification] TS safety limit for MCPR is 1.07. To ensure that the

MCPR safety limit is not exceeded during design basis transients and

accidents, an operating limit is conservatively placed on the MCPR

during normal plant operation (OLMCPR). The amount of reactivity

inserted at rod position 46 (approximately 5% of rod insertion) is

small. The analysis used to establish the OLMCPR does not consider

the scram insertion time at position 46 but does consider the scram

insertion time to rod position 38 for the most limiting transient

(turbine load rejection without bypass). The required scram time to

position 38 remains unchanged by this proposed amendment. A generic

BWR/2-5 study showed that relaxing the 5% rod insertion time

requirement had a negligible impact on plant transient performance.

This change will not result in any changes to the calculated OLMCPR,

which assures that the safety limit MCPR will not be exceeded.

Therefore, this change will not reduce the margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on

this review, it appears that the three standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendment request involves no significant hazards consideration.

Local Public Document Room location: Cedar Rapids Public Library,

500 First Street, S.E., Cedar Rapids, Iowa 52401.

Attorney for licensee: Jack Newman, Kathleen H. Shea, Morgan,

Lewis, & Bockius, 1800 M Street, NW., Washington, DC 20036-5869.

NRC Project Director: Gail H. Marcus.

Northeast Nuclear Energy Company, et al., Docket No. 50-336, Millstone

Nuclear Power Station, Unit No. 2, New London, Connecticut

Date of amendment request: December 18, 1995.

Description of amendment request: The Allowable Value for the

Reactor Coolant Flow Instrumentation contained in Table 2.2-1 is

proposed to be changed to reflect the design changes implemented during

the last refueling outage. The Reactor Coolant System (RCS) Steam

Generator Differential Pressure Instrumentation Loops have been

modified to reflect a re-calibration of the differential pressure

transmitter from ``-8 to 64 psid'' to ``0 to 35 psid,'' and an

elimination of the Foxboro signal characterizer modules from the

instrument loop string.

Additionally, an editorial change is proposed for the text

associated with the allowable value. The current wording ``reactor

coolant'' is being changed to ``reactor coolant flow.''

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no significant hazards

consideration which is presented below:

Pursuant to 10 CFR 50.92, NNECO has reviewed the proposed

changes. NNECO concludes that these changes do not involve a

significant hazards consideration (SHC) since the proposed changes

satisfy the criteria in 10 CFR 50.92(c). That is, the proposed

changes do not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed change to the Allowable Value of the Reactor

Coolant Flow Instrumentation is based on design changes that reduce

the uncertainties in the overall instrument loop, as well as

improved calculation methodology for instrument uncertainty and

setpoint. The new hardware configuration results in calculated

uncertainties which are bounded by the Safety Analysis assumptions.

There is no adverse impact on any design basis analysis due to this

change, and, therefore does not affect the probability or

consequence of any previously evaluated accident.

Additionally, the proposed change to add the word ``flow'' is an

editorial correction and therefore does not affect the probability

or consequence of any previously evaluated accident.

2. Create the possibility of a new or different kind of accident

from any accident previously evaluated.

The new Allowable Value has been calculated using an improved

methodology. The new hardware configuration results in calculated

uncertainties which are bounded by the Safety Analysis assumptions.

The function of the Allowable Value is not changed. Therefore no new

accident scenarios are created.

Additionally, the proposed change to add the word ``flow'' is an

editorial correction and therefore no new accident scenarios are

created.

3. Involve a significant reduction in a margin of safety.

The change to the Allowable Value for the Reactor Coolant Flow

Instrumentation reflects the design changes implemented during the

last refueling outage. The design improvement of the loop

performance ensures that the assumptions of the Safety Analysis are

met. Since the proposed changes do not affect the consequences of

any accident previously analyzed, there is no reduction in a margin

of safety.

Additionally, the proposed change to add the word ``flow'' is an

editorial correction and has no effect on the margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on

this review, it appears that the three standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendment request involves no significant hazards consideration.

Local Public Document Room location: Learning Resources Center,

Three Rivers Community-Technical College, 574 New London Turnpike,

Norwich, CT 06360.

Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear

Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,

CT 06141-0270.

NRC Project Director: Phillip F. McKee.

[[Page 5816]]

Northeast Nuclear Energy Company, et al., Docket No. 50-336, Millstone

Nuclear Power Station, Unit No. 2, New London, Connecticut

Date of amendment request: January 5, 1996.

Description of amendment request: Northeast Nuclear Energy Company

(NNECO) is proposing to implement the guidance of Generic Letter 93-08

and relocate Tables 3.3-2, ``Reactor Protective Instrumentation

Response Times'' and 3.3-5, ``Engineered Safety Features Response

Times'' from the technical specifications to the Millstone Unit No. 2

Technical Requirements Manual (TRM). In accordance with Generic Letter

93-08, the Limiting Conditions for Operations for Technical

Specifications 3.3.1.1, 3.3.2.1, and 3.7.1.6 are also proposed to be

revised to eliminate their references to the aforementioned tables.

NNECO has also proposed to revise Bases 3/4.3.1 and 3/4.3.2 to

reference that the instrument response times are located in the TRM and

that these tables in the TRM are now controlled under 10CFR50.59. NNECO

also proposes to remove a cycle-specific note from Tables 3.3-3 and

3.3-4.

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:

In accordance with 10CFR50.92, NNECO has reviewed the attached

proposed changes and has concluded that they do not involve a

significant hazards consideration. The basis of this conclusion is

that the three criteria of 10CFR50.92(c) are not compromised. The

proposed changes do not involve a significant hazards consideration

because the changes would not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed license amendment will remove the reactor

protective system and engineered safety feature actuation response

times from the technical specifications. This proposed change will

not affect the operation of the reactor protective system and the

engineered safety feature actuation system. Operability and

surveillance requirements are still maintained in the technical

specifications and the response times will be included and

maintained in the Technical Requirements Manual (TRM). Once

relocated to the TRM, any future proposed changes will require a

safety evaluation and Plant Operations Review Committee review.

The proposed license amendment will also delete the cycle-

specific note contained in Tables 3.3-2 and 3.3-4. This is

administrative in nature and do not result in changes to plant

configuration, operation, accident mitigation, or analysis

assumptions. The notes was in effect only during Cycle 12.

Since the systems will not be affected by the proposed changes,

there is no impact on the performance of these systems or on the

probability or consequences of an accident previously analyzed.

2. Create the possibility of a new or different kind of accident

from any previously evaluated.

There are no new failure modes associated with the proposed

changes. Since the plant will continue to operate as designed, the

proposed changes will not modify plant responses to the point where

it can be considered a new or different kind of accident.

Involve a significant reduction in a margin of safety.

The proposed changes do not have any adverse impact on the

protective boundaries nor do they affect the consequences of any

accident previously analyzed. The portion of the change associated

with Generic Letter 93-08 will not affect the technical

specification operability and surveillance requirements which will

still ensure that the systems are tested and are within limits.

Changing the limits requires a safety evaluation and Plant

Operations Review Committee review. This will ensure that the

licensing basis is maintained.

The proposed changes to delete the cycle-specific notes are

administrative in nature and do not result in changes to plant

configuration, operation, accident mitigation, or analysis

assumptions. The notes were in effect only during Cycle 12.

Therefore, the proposed changes will not result in 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: Learning Resources Center,

Three Rivers Community-Technical College, 574 New London Turnpike,

Norwich, CT 06360.

Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear

Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,

CT 06141-0270.

NRC Project Director: Phillip F. McKee.

Northeast Nuclear Energy Company, et al., Docket No. 50-336, Millstone

Nuclear Power Station, Unit No. 2, New London, Connecticut

Date of amendment request: January 26, 1996.

Description of amendment request: The licensee proposes to modify

the Technical Specifications for Millstone Unit No. 2 as follows:

1. Limiting Condition for Operation 3.6.1.2.a-c: Replace the less

than or equal to sign with a ``h remote

shutdown indication. The minimum number of channels required is being

revised from one per RCS Loop for each RCS Loop to one per RCS Loop for

three of the four RCS Loops. These changes are only applicable to CPSES

Unit 2 and are being submitted on the CPSES Unit 1 docket for

administrative purposes only because the CPSES TSs is a single document

which applies to both units.

Date of issuance: February 2, 1996.

Effective date: February 2, 1996.

Amendment Nos.: Unit 1--Amendment No. 45; Unit 2--Amendment No. 31.

Facility Operating License Nos. NPF-87 and NPF-89: The amendments

revised the Technical Specifications.

Public comments requested as to proposed significant hazards

consideration: Yes (61 FR 1651, dated January 22, 1996). The notice

provided an opportunity to submit comments on the Commission's proposed

no significant hazards consideration determination. No comments have

been received. The notice also provided for an opportunity to request a

hearing by February 21, 1996, but stated that any such hearing would

take place after issuance of the amendment. The Commission's related

evaluation of the amendments, finding of exigent circumstances, and

final determination of no significant hazards consideration is

contained in a Safety Evaluation dated February 2, 1996.

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.

Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee

Nuclear Power Plant, Kewaunee County, Wisconsin

Date of application for amendment: October 18, 1995.

Brief description of amendment: The amendment revises Kewaunee

Nuclear Power Plant Technical Specification (TS) 3.4, ``Steam and Power

Conversion System,'' by modifying and clarifying the operability

requirements for the main steam safety valves (MSSVs), the auxiliary

feedwater (AFW) System, and the condensate storage tank system. The

amendment also eliminates inconsistencies within TS Section 3.4 and

provides the basis for acceptable operation of the Auxiliary Feedwater

System below 15% reactor power.

Date of issuance: January 3, 1996.

Effective date: January 3, 1996.

Amendment No.: 123.

Facility Operating License No. DPR-43: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: November 27, 1995 (60

FR 58407).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated January 3, 1996.

No significant hazards consideration comments received: No.

Local Public Document Room location: University of Wisconsin,

Cofrin Library, 2420 Nicolet Drive, Green Bay, Wisconsin 54311-7001.

Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek

Generating Station, Coffey County, Kansas

Date of amendment request: October 18, 1995.

Brief description of amendment: This amendment replaces the current

fuel oil volume requirement in the emergency diesel generator (EDG) day

tank in Technical Specifications 3.8.1.1.b.1) and 3.8.1.2.b.1) with a

fuel oil level requirement. Associated Surveillance Requirement

4.8.1.1.2.a.1) is also changed to replace the visual check requirement

on fuel oil level in the day tank with a requirement to verify that the

fuel oil transfer pump starts on low level in the day tank standpipe.

The associated Bases section is also revised to reflect the above

changes.

Date of issuance: January 19, 1996.

Effective date: January 19, 1996, to be implemented prior to

startup from the eighth refueling outage currently scheduled to begin

in March 1996.

Amendment No.: 94.

Facility Operating License No. NPF-42. The amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: November 27, 1995 (60

FR 58049).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated January 19, 1996.

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.

[[Page 5826]]

Notice of Issuance of Amendments to Facility Operating Licenses and

Final Determination of No Significant Hazards Consideration and

Opportunity for a Hearing (Exigent Public Announcement or Emergency

Circumstances)

During the period since publication of the last biweekly notice,

the Commission has issued the following amendments. The Commission has

determined for each of these amendments that the application for the

amendment complies with the standards and requirements of the Atomic

Energy Act of 1954, as amended (the Act), and the Commission's rules

and regulations. The Commission has made appropriate findings as

required by the Act and the Commission's rules and regulations in 10

CFR Chapter I, which are set forth in the license amendment.

Because of exigent or emergency circumstances associated with the

date the amendment was needed, there was not time for the Commission to

publish, for public comment before issuance, its usual 30-day Notice of

Consideration of Issuance of Amendment, Proposed No Significant Hazards

Consideration Determination, and Opportunity for a Hearing.

For exigent circumstances, the Commission has either issued a

Federal Register notice providing opportunity for public comment or has

used local media to provide notice to the public in the area

surrounding a licensee's facility of the licensee's application and of

the Commission's proposed determination of no significant hazards

consideration. The Commission has provided a reasonable opportunity for

the public to comment, using its best efforts to make available to the

public means of communication for the public to respond quickly, and in

the case of telephone comments, the comments have been recorded or

transcribed as appropriate and the licensee has been informed of the

public comments.

In circumstances where failure to act in a timely way would have

resulted, for example, in derating or shutdown of a nuclear power plant

or in prevention of either resumption of operation or of increase in

power output up to the plant's licensed power level, the Commission may

not have had an opportunity to provide for public comment on its no

significant hazards consideration determination. In such case, the

license amendment has been issued without opportunity for comment. If

there has been some time for public comment but less than 30 days, the

Commission may provide an opportunity for public comment. If comments

have been requested, it is so stated. In either event, the State has

been consulted by telephone whenever possible.

Under its regulations, the Commission may issue and make an

amendment immediately effective, notwithstanding the pendency before it

of a request for a hearing from any person, in advance of the holding

and completion of any required hearing, where it has determined that no

significant hazards consideration is involved.

The Commission has applied the standards of 10 CFR 50.92 and has

made a final determination that the amendment involves no significant

hazards consideration. The basis for this determination is contained in

the documents related to this action. Accordingly, the amendments have

been issued and made effective as indicated.

Unless otherwise indicated, the Commission has determined that

these amendments satisfy the criteria for categorical exclusion in

accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),

no environmental impact statement or environmental assessment need be

prepared for these amendments. If the Commission has prepared an

environmental assessment under the special circumstances provision in

10 CFR 51.12(b) and has made a determination based on that assessment,

it is so indicated.

For further details with respect to the action see (1) the

application for amendment, (2) the amendment to Facility Operating

License, and (3) the Commission's related letter, Safety Evaluation

and/or Environmental Assessment, as indicated. All of these items are

available for public inspection at the Commission's Public Document

Room, the Gelman Building, 2120 L Street, NW., Washington, DC, and at

the local public document room for the particular facility involved.

The Commission is also offering an opportunity for a hearing with

respect to the issuance of the amendment. By March 15, 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