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

NUCLEAR REGULATORY COMMISSION

Biweekly Notice

Applications and Amendments to Facility Operating Licenses

Involving No Significant Hazards Considerations

I. Background

Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory

Commission (the Commission or NRC staff) is publishing this regular

biweekly notice. Public Law 97-415 revised section 189 of the Atomic

Energy Act of 1954, as amended (the Act), to require the Commission to

publish notice of any amendments issued, or proposed to be issued,

under a new provision of section 189 of the Act. This provision grants

the Commission the authority to issue and make immediately effective

any amendment to an operating license upon a determination by the

Commission that such amendment involves no significant hazards

consideration, notwithstanding the pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all notices of amendments issued, or

proposed to be issued from May 11, 1996, through May 23, 1996. The last

biweekly notice was published on May 22, 1996 (61 FR 25696).

Notice Of Consideration Of Issuance Of Amendments To Facility

Operating Licenses, Proposed No Significant Hazards onsideration

Determination, And Opportunity For A Hearing

The Commission has made a proposed determination that the following

amendment requests involve no significant hazards consideration. Under

the Commission's regulations in 10 CFR 50.92, this means that operation

of the facility in accordance with the proposed amendment would not (1)

involve a significant increase in the probability or consequences of an

accident previously evaluated; or (2) create the possibility of a new

or different kind of accident from any accident previously evaluated;

or (3) involve a significant reduction in a margin of safety. The basis

for this proposed determination for each amendment request is shown

below.

The Commission is seeking public comments on this proposed

determination. Any comments received within 30 days after the date of

publication of this notice will be considered in making any final

determination.

Normally, the Commission will not issue the amendment until the

expiration of the 30-day notice period. However, should circumstances

change during the notice period such that failure to act in a timely

way would result, for example, in derating or shutdown of the facility,

the Commission may issue the license amendment before the expiration of

the 30-day notice period, provided that its final determination is that

the amendment involves no significant hazards consideration. The final

determination will consider all public and State comments received

before action is taken. Should the Commission take this action, it will

publish in the Federal Register a notice of issuance and provide for

opportunity for a hearing after issuance. The Commission expects that

the need to take this action will occur very infrequently.

Written comments may be submitted by mail to the Chief, Rules

Review and Directives Branch, Division of Freedom of Information and

Publications Services, Office of Administration, U.S. Nuclear

Regulatory Commission, Washington, DC 20555-0001, and should cite the

publication date and page number of this Federal Register notice.

Written comments may also be delivered to Room 6D22, Two White Flint

North, 11545 Rockville Pike, Rockville, Maryland from 7:30 a.m. to 4:15

p.m. Federal workdays. Copies of written comments received may be

examined at the NRC Public Document Room, the Gelman Building, 2120 L

Street, NW., Washington, DC. The filing of requests for a hearing and

petitions for leave to intervene is discussed below.

By July 5, 1996, the licensee may file a request for a hearing with

respect to issuance of the amendment to the subject facility operating

license and any person whose interest may be affected by this

proceeding and who wishes to participate as a party in the proceeding

must file a written request for a hearing and a petition for leave to

intervene. Requests for a hearing and a petition for leave to intervene

shall be filed in accordance with the Commission's ``Rules of Practice

for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested

persons should consult a current copy of 10 CFR 2.714 which is

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

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

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

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

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

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

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

Secretary or the designated Atomic Safety and Licensing Board will

issue a notice of a hearing or an appropriate order.

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

shall set forth with particularity the interest of the petitioner in

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

the proceeding. The petition should specifically explain the reasons

why intervention should be permitted with particular reference to the

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

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

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

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

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

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

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

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

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

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

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

the specificity requirements described above.

Not later than 15 days prior to the first prehearing conference

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

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

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

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

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

brief explanation of the bases of the contention and a concise

statement of the alleged facts or expert opinion which support the

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

contention at the hearing. The petitioner must also

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provide references to those specific sources and documents of which the

petitioner is aware and on which the petitioner intends to rely to

establish those facts or expert opinion. Petitioner must provide

sufficient information to show that a genuine dispute exists with the

applicant on a material issue of law or fact. Contentions shall be

limited to matters within the scope of the amendment under

consideration. The contention must be one which, if proven, would

entitle the petitioner to relief. A petitioner who fails to file such a

supplement which satisfies these requirements with respect to at least

one contention will not be permitted to participate as a party.

Those permitted to intervene become parties to the proceeding,

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

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

hearing, including the opportunity to present evidence and cross-

examine witnesses.

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

determination on the issue of no significant hazards consideration. The

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

If the final determination is that the amendment request involves

no significant hazards consideration, the Commission may issue the

amendment and make it immediately effective, notwithstanding the

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

of the amendment.

If the final determination is that the amendment request involves a

significant hazards consideration, any hearing held would take place

before the issuance of any amendment.

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

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

Commission, Washington, DC 20555-0001, Attention: Docketing and

Services Branch, or may be delivered to the Commission's Public

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

by the above date. Where petitions are filed during the last 10 days of

the notice period, it is requested that the petitioner promptly so

inform the Commission by a toll-free telephone call to Western Union at

1-(800) 248-5100 (in Missouri 1-(800) 342-6700). The Western Union

operator should be given Datagram Identification Number N1023 and the

following message addressed to (Project Director): petitioner's name

and telephone number, date petition was mailed, plant name, and

publication date and page number of this Federal Register notice. A

copy of the petition should also be sent to the Office of the General

Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001,

and to the attorney for the licensee.

Nontimely filings of petitions for leave to intervene, amended

petitions, supplemental petitions and/or requests for a hearing will

not be entertained absent a determination by the Commission, the

presiding officer or the Atomic Safety and Licensing Board that the

petition and/or request should be granted based upon a balancing of

factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).

For further details with respect to this action, see the

application for amendment which is available for public inspection at

the Commission's Public Document Room, the Gelman Building, 2120 L

Street, NW., Washington, DC, and at the local public document room for

the particular facility involved.

Boston Edison Company, Docket No. 50-293, Pilgrim Nuclear Power

Station, Plymouth County, Massachusetts

Date of amendment request: May 1, 1996

Description of amendment request: The proposed amendment will

relocate the administrative controls related to the quality assurance

review and audit requirements of Section 6 from the Pilgrim Station

Technical Specifications to the Boston Edison Quality Assurance Manual.

This change is in accordance with the guidance contained in NRC

Administrative Letter 95-06, ``Relocation of Technical Specification

Administrative Controls Related to Quality Assurance.''

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

The change will relocate the administrative controls related to

the quality assurance review and audit requirements from the

technical specifications to the quality assurance plan. These

changes are administrative in nature and do not impact initiators of

analyzed events, accident mitigation capabilities, or transient

events. The quality assurance program is a logical candidate for

such relocation due to the controls imposed by such regulations as

Appendix B to 10 CFR [Part] 50, the existence of NRC approved

quality assurance plans and commitments to industry quality

assurance standards, and the established quality assurance program

change control process in 10 CFR 50.54(a). Therefore, the changes do

not involve a significant increase in the probability or

consequences of an accident previously evaluated.

2. The proposed amendment does not create the possibility of a

new or different kind of accident from any accident previously

evaluated.

The change will relocate the administrative controls related to

the quality assurance review and audit requirements from the

technical specifications to the quality assurance plan. The quality

assurance program is a logical candidate for such relocation due to

the controls imposed by such regulations as Appendix B to 10 CFR

[Part] 50, the existence of NRC approved quality assurance plans and

commitments to industry quality assurance standards, and the

established quality assurance program change control process in 10

CFR 50.54(a). The proposed changes do not involve a physical

alteration of the plant or changes in methods governing plant

operation. The changes will not impose or eliminate any new or

different requirements. Therefore the changes do not create the

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

previously evaluated.

3. The proposed amendment does not involve a significant

reduction in a margin of safety.

The change will relocate the administrative controls related to

the quality assurance review and audit requirements from the

technical specifications to the quality assurance plan. These

changes are administrative in nature. The quality assurance program

is a logical candidate for such relocation due to the controls

imposed by such regulations as Appendix B to 10 CFR [Part] 50, the

existence of NRC approved quality assurance plans and commitments to

industry quality assurance standards, and the established quality

assurance program change control process in 10 CFR 50.54(a). The

proposed change will not reduce a margin of safety because it has no

impact on any safety analysis assumptions. Therefore, the operation

of PNPS [Pilgrim Nuclear Power Station] in accordance with the

proposed license amendment will not involve a significant reduction

in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Plymouth Public Library, 11

North Street, Plymouth, Massachusetts 02360.

Attorney for licensee: W. S. Stowe, Esquire, Boston Edison Company,

800 Boylston Street, 36th Floor, Boston, Massachusetts 02199.

NRC Project Director: Jocelyn A. Mitchell, Acting

[[Page 28606]]

Boston Edison Company, Docket No. 50-293, Pilgrim Nuclear Power

Station, Plymouth County, Massachusetts

Date of amendment request: May 1, 1996

Description of amendment request: The proposed amendment will

reflect the implementation of 10 CFR Part 50, Appendix J, Option B at

the Pilgrim Nuclear Power Station.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The operation of Pilgrim Station in accordance with the

proposed amendment will not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The proposed changes do not involve a significant increase in

the probability or consequences of an accident previously evaluated.

The proposed changes do not involve any physical or operational

changes to structures, systems or components. The proposed changes

provide a mechanism within the TS [Technical Specifications] for

implementing a performance-based leakage rate test program which was

promulgated by the revision to 10CFR50 to incorporate Option B into

Appendix J. The TS Limiting Conditions for Operation (LCO) remain

unaffected by these changes. Thus, the safety design basis for the

accident mitigation functions of the primary containment is

maintained. Therefore, these changes will not increase the

probability or consequences of an accident previously evaluated.

2. The operation of Pilgrim Station in accordance with the

proposed amendment will not create the possibility of a new or

different kind of accident from any accident previously evaluated.

Revising surveillance requirement acceptance criteria and

frequencies does not physically modify the plant and does not modify

the operation of any existing equipment. Further, the TS LCOs remain

unaffected by these changes.

3. The operation of Pilgrim Station in accordance with the

proposed amendment will not involve a significant reduction in a

margin of safety.

The proposed changes do not involve a significant reduction in

the margin of safety, nor do they affect a safety limit, an LCO, or

the manner in which plant equipment is operated. The NRC letter

dated November 2, 1995, recognizes that changes similar to the

proposed changes are required to implement Option B of 10CFR50,

Appendix J. In NUREG-1493, ``Performance-Based Containment Leak-Test

Program,'' which forms the basis for the Appendix J revision, the

NRC concludes that adoption of performance-based test intervals for

Appendix J testing will not significantly 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 50.92(c) are

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Plymouth Public Library, 11

North Street, Plymouth, Massachusetts 02360.

Attorney for licensee: W. S. Stowe, Esquire, Boston Edison Company,

800 Boylston Street, 36th Floor, Boston, Massachusetts 02199.

NRC Project Director: Jocelyn A. Mitchell, Acting

Boston Edison Company, Docket No. 50-293, Pilgrim Nuclear Power

Station, Plymouth County, Massachusetts

Date of amendment request: May 1, 1996

Description of amendment request: The proposed amendment would

modify the definition of ``Core Alteration,'' and the Limiting

Condition for Operation, Surveillance conditions and Bases section

associated with Technical Specification (TS) 3.7.C, ``Secondary

Containment.''

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

Operation of PNPS [Pilgrim Nuclear Power Station] in accordance

with the proposed license amendment will not involve a significant

increase in the probability or consequences of an accident

previously evaluated because of the following:

Proposed Change 1: Definition of ``Alteration of the

Reactor Core

The definition, ``Alteration of the Reactor Core'', is being

revised so that the term will apply only to those activities that

create the potential for a reactivity excursion and, therefore,

warrant special precautions or controls in the TS. The proposed

definition includes normal control rod movement in the definition,

but excludes control rod drive movement (such as rod removal from

the core) when all four fuel bundles surrounding a control rod are

removed. The proposed change does not increase the probability or

consequences of an accident because the proposed definition, by

identifying activities with the potential for causing a reactivity

excursion, ensures that the additional precautions and controls in

the TS are implemented at all appropriate times. In addition, the

movement of components excluded by this definition is not assumed in

the initiation of any analyzed event. Therefore, the proposed change

does not involve a significant increase in the probability or

consequences of an accident previously evaluated.

Proposed Change 2: Secondary Containment

The current specifications are revised to specify more clearly

when secondary containment is required, what actions to take if

secondary containment is inoperable, and time frames for completing

the actions. These revisions enhance the existing specification and

serve to make it more definitive by encompassing the conditions

currently specified by TS and supplementing them to specify other

conditions when secondary containment is required.

Surveillances 4.7.C.1.a and b were only necessary during initial

and Cycle 1 operations. Removing obsolete information from the

existing specifications, re-numbering and re-arranging the wording

is an administrative change.

These changes are administrative in nature and do not impact

initiators of analyzed events, accident mitigation capabilities, or

transient events. Therefore, the changes do not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

2. The proposed amendment does not create the possibility of a

new or different kind of accident from any accident previously

evaluated.

The operation of PNPS in accordance with the proposed license

amendment will not create the possibility of a new or different kind

of accident from any accident previously evaluated because of the

following:

Proposed Change 1: Definition of ``Alteration of the

Reactor Core

The definition change specifies more accurately which component

movements constitute a ``Core Alteration''. This change does not

involve a physical alteration of the plant (no new or different type

of equipment will be installed) or changes in methods governing

normal plant operation. The proposed changes will allow movement of

some components (camera, lights, etc.) during times when ``Core

Alterations'' have been halted since these components will not

affect core reactivity. Removal of a control rod involves unlatching

and withdrawal/insertion from over-vessel handling equipment. These

activities necessitate, by design, the removal of the adjacent four

fuel assemblies. With this configuration (no fuel in the cell;

handling the associated control rod), the proposed change will allow

movement of a ``reactivity control component'' while not imposing

requirements unique to ``Core Alterations'' (note: other

requirements, such as those for handling loads over irradiated fuel,

will remain applicable). The reactivity effects of this control rod

movement are more than compensated for by the initial removal of the

fuel assemblies. Therefore, this change will not create the

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

previously evaluated.

Proposed Change 2: Secondary Containment

The proposed change does not eliminate or relax any existing TS

condition. Rather, it better defines when secondary containment is

required, provides action statements for inoperability and removes

obsolete

[[Page 28607]]

requirements (from first operating cycle). This change does not

involve a physical change to structures, systems or components, and

the safety design bases for the accident mitigating function of the

secondary containment is maintained. Therefore, these changes will

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

from any accident previously evaluated.

3. The proposed amendment does not involve a significant

reduction in a margin of safety.

The operation of PNPS in accordance with the proposed license

amendment will not involve a significant reduction in a margin of

safety because of the following:

Proposed Change 1: Definition of ``Alteration of the

Reactor Core

The proposed definition more accurately identifies those

activities with the potential for causing a reactivity excursion.

The more accurate identification of ``Core Alterations'' will ensure

that when there is a potential for reactivity excursions,

appropriate precautions are applied. The components now excluded

from the proposed definition are those that do not have the

capability for adversely impacting core reactivity. The proposed

change has no impact on safety analysis assumptions. Therefore, the

change will not involve a significant reduction in a margin of

safety.

Proposed Change 2: Secondary Containment

The proposed additions of applicability conditions provide a

more precise understanding of when secondary containment integrity

is required and what actions to take if it becomes inoperable. The

change does not eliminate any existing conditions. The deletion of

surveillances applicable only for the first operating cycle and re-

numbering and re-arranging the remaining surveillance wording is an

administrative change and has no impact on the operation of the

plant or mitigation of accidents. Therefore, the operation of the

facility in accordance with this 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: Plymouth Public Library, 11

North Street, Plymouth, Massachusetts 02360.

Attorney for licensee: W. S. Stowe, Esquire, Boston Edison Company,

800 Boylston Street, 36th Floor, Boston, Massachusetts 02199.

NRC Project Director: Jocelyn A. Mitchell, Acting

Carolina Power & Light Company, et al., Docket No. 50-325,

Brunswick Steam Electric Plant, Unit 1, Brunswick County, North

Carolina

Date of amendment request: April 8, 1996

Description of amendment request: The licensee has proposed to

revise the Technical Specifications (TS) to include the following

changes: 1. The Minimum Critical Power Ratio (MCPR) Safety Limit

specified in TS 2.1.2 from 1.07 to 1.09 for Unit 1 Cycle 11 operation;

TS 5.3.1 to reflect the new fuel type (GE13) that will be inserted

during Unit 1 Refueling Outage 10; 2. The acceptable range of sodium

pentaborate concentration for the standby liquid control system shown

in TS Figure 3.1.5-1 to reflect changes to poison material

concentration needed to achieve reactor shutdown based on the new GE13

fuel type.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

Proposed Change 1

The proposed amendment will allow the loading and use of GE13

fuel assemblies in the Brunswick Unit 1 reactor core. The use of

GE13 fuel assemblies requires that the safety limit minimum critical

power ratio value also be revised. The safety limit minimum critical

power ratio is established to maintain fuel cladding integrity

during operational transients. The GE13 fuel assembly design has

been analyzed using methods that have been previously approved by

the Nuclear Regulatory Commission and documented in General Electric

Nuclear Energy's reload licensing methodology Topical Report (NEDE-

24011-P-A-11, ``General Electric Standard Application for Reactor

Fuel (GESTAR II)'' dated November 1995).

The proposed revision of the safety limit minimum critical power

ratio does not alter any plant safety-related equipment, safety

function, or plant operations that could change the probability of

an accident. The change does not affect the design, materials, or

construction standards applicable to the fuel bundles in a manner

that could change the probability of an accident.

A methodology that has been previously reviewed and accepted by

the Nuclear Regulatory Commission was used to derive both the

existing and updated safety limit minimum critical power ratio

value. The same methodology and criteria have been applied to derive

the existing safety limit minimum critical power ratio of 1.07 as

that used to derive the updated safety limit minimum critical power

ratio value of 1.09. The updated safety limit minimum critical power

ratio assures that fuel cladding protection equivalent to that

provided with the existing safety limit minimum critical power ratio

value is maintained. This ensures that the consequences of

previously evaluated accidents are not significantly increased.

Proposed Change 2

The standby liquid control system provides a means of reactivity

control that is independent of the normal reactivity control system.

The standby liquid control system must be capable of assuring that

the reactor core can be placed in a subcritical condition at any

time during reactor core life. Technical Specification Figure 3.1.5-

1 specifies the acceptable range of concentrations and volumes for

sodium pentaborate solution used as a neutron absorber (i.e., for

reactivity control). The portion of the sodium pentaborate

concentration range shown in Technical Specification Figure 3.1.5-1

applicable to the lower range of tank volumes is being revised to

increase the required concentration of sodium pentaborate solution.

This change is needed to account for the additional shutdown

reactivity needed based on the planned use of GE13 fuel assemblies

as reload fuel for the Unit 1 reactor core. Since the standby liquid

control system is independent from the normal means of controlling

reactor core reactivity and not used to control core reactivity

during normal plant operations, the proposed revision to the sodium

pentaborate concentration curve for the standby liquid control

system does not alter any plant safety-related equipment, safety

function, or plant operations that could change the probability of

an accident.

The current volume-concentration range of sodium pentaborate

used in the standby liquid control system will achieve a sufficient

concentration of boron in the reactor vessel to ensure reactor

shutdown. Based on the increased reactivity of the new GE13 reload

fuel assemblies, the required sodium pentaborate volume-

concentration range is being revised to ensure sufficient neutron

absorbing solution is available to achieve reactor shutdown;

therefore, the consequences of an accident previously evaluated are

not significantly increased.

2. The proposed amendment would not create the possibility of a

new or different kind of accident from any accident previously

evaluated.

Proposed Change 1

The GE13 fuel assembly has been designed and complies with the

acceptance criteria contained in General Electric Nuclear Energy's

standard application for reactor fuel (GESTAR-II), which provides

the latest acceptance criteria for new General Electric fuel

designs. The GE13 fuel assembly complies with GESTAR-II acceptance

criteria that have been previously reviewed and accepted by the

Nuclear Regulatory Commission. The similarity of the GE13 fuel

design to the previously accepted GE11 fuel design, in conjunction

with the increased critical power capability of the GE13 fuel

design, ensure that no new mode or condition of plant operation is

being authorized by the loading and use of the

[[Page 28608]]

GE13 fuel type. The proposed revision of the safety limit minimum

critical power ratio from 1.07 to 1.09 does not modify any plant

controls or equipment that will change the plant's responses to any

accident or transient as given in any current analysis. Therefore,

the proposed change to allow the loading and use of the GE13 fuel

type and the revision of the safety limit minimum critical power

ratio value from 1.07 to 1.09 will not create the possibility for a

new or different kind of accident from any accident previously

evaluated.

Proposed Change 2

As discussed above, the standby liquid control system provides a

means of reactivity control that is independent of the normal

reactivity control system and is capable of assuring that the

reactor core can be placed in a subcritical condition at any time

during reactor core life. The proposed revision to the sodium

pentaborate concentration range does not modify the standby liquid

control system or its controls, does not modify other plant systems

and equipment, and does not permit a new or different mode of plant

operation. As such, the proposed revision to the minimum pentaborate

concentration value does not create the possibility of a new or

different kind of accident from any accident previously evaluated.

3. The proposed license amendment does not involve a significant

reduction in a margin of safety.

Proposed Change 1

As previously discussed, the GE13 fuel assembly design has been

analyzed using methods that have been previously approved by the

Nuclear Regulatory Commission and documented in General Electric

Nuclear Energy's reload licensing methodology Topical Report (NEDE-

24011-P-A-11, ``General Electric Standard Application for Reactor

Fuel (GESTAR II)'' dated November 1995). The safety limit minimum

critical power ratio value is selected to maintain the fuel cladding

integrity safety limit (i.e., that 99.9 percent of all fuel rods in

the core are expected to avoid boiling transition during operational

transients). Appropriate operating limit minimum critical power

ratio values are established, based on the safety limit minimum

critical power ratio value, to ensure that the fuel cladding

integrity safety limit is maintained. The operating limit minimum

critical power ratio values are incorporated in the Core Operating

limits Report as required by Technical Specification 6.9.3.1. The

new GE13 safety limit minimum critical power ratio value of 1.09 is

based on the same fuel cladding integrity safety limit criteria [as]

that for the GE11 safety limit minimum critical power ratio value of

1.07 (i.e., that 99.9 percent of all fuel rods in the core are

expected to avoid boiling transition during operational transients);

therefore, the proposed change does not result in a significant

reduction in the margin of safety.

Proposed Change 2

As previously stated, the purpose of the standby liquid control

is to inject a neutron absorbing solution into the reactor in the

event that a sufficient number of control rods cannot be inserted to

maintain subcriticality. Sufficient solution is to be injected such

that the reactor will be brought from maximum rated power conditions

to subcritical over the entire reactor temperature range from

maximum operating to cold shutdown conditions. General Electric

methodology establishes a fuel type dependent standby liquid control

system shutdown margin to account for calculational uncertainties.

General Electric calculations show that an in-vessel concentration

of 660 ppm will provide a standby liquid control system minimum

shutdown margin in excess of the 3.2%[delta]k value required for the

GE13 fuel. To achieve an in-vessel concentration of 660 ppm, the

acceptable range of standby liquid control system tank

concentrations is being revised for the lower range of tank volumes.

Thus, the proposed revision of the standby liquid control system

sodium pentaborate volume-concentration range ensures that there

will not be a significant reduction in the amount of available

shutdown margin and, therefore, not a significant reduction in the

margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: University of North Carolina

at Wilmington, William Madison Randall Library, 601 S. College Road,

Wilmington, North Carolina 28403-3297.

Attorney for licensee: William D. Johnson, Vice President and

Senior Counsel, Carolina Power & Light Company, Post Office Box 1551,

Raleigh, North Carolina 27602

NRC Project Director: Eugene V. Imbro

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

Neck Plant, Middlesex County, Connecticut

Date of amendment request: February 27, 1996

Description of amendment request: The proposed license amendment

would modify the Action Statement of Technical Specification (TS)

3.7.1.1.1. Currently, the TS action statement requires that with the

self actuation function on one or more main steam line code safety

valves associated with an operating loop inoperable, the licensee must

restore the inoperable valve to operable status within 4 hours.

Otherwise, the plant must be in hot standby within the next 6 hours and

in hot shutdown within the following 30 hours. The proposed change will

allow continued power operation at reduced power levels with main steam

safety valves inoperable. The proposed change is consistent with the

philosophy of the Westinghouse Standard Technical Specifications.

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 the Action Statement of LCO [Limiting

Condition for Operation] 3.7.1.1.1 will allow indefinite operation

at less than or equal to 75% power in the event that the self

actuation function of no more than one safety valve per steam

generator is inoperable, and allow indefinite operation at less than

or equal to 50% power in the event that the self actuation function

of no more than two safety valves per steam generator is inoperable.

The requirement to reduce power will ensure that there is no

increase in the consequences of a loss of load accident. The

proposed change is consistent with the methodology in the

Westinghouse Standard Technical Specifications. The methodology is

conservative, since the PORVs [power operated relief valves] cannot

affect the time of reactor trip on high pressurizer pressure. Thus,

it is concluded that the change does not increase the consequences

of any previously evaluated accident.

The change only specifies a power reduction in the event that

the self actuation function of steam generator safety valves is

inoperable. It does not affect the probability of any accident. The

change by itself does not affect the likelihood of an inoperable

safety valve.

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

new or different kind of accident from any previously evaluated.

The change only specifies a power reduction in the event that

the self actuation function of steam generator safety valves is

inoperable. This does not create the potential for a new or

different kind of accident. The lower power level assures that peak

steam generator pressure and RCS [reactor coolant system] pressure

will remain below 110% of design. This provides assurance that no

equipment failure will occur due to overpressurization. Thus, the

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

of accident.

3. [The proposed change does not involve] a significant

reduction in a margin of safety.

The allowable power levels have been selected, consistent with

the Westinghouse Standard Technical Specifications, to assure that

steam generator and RCS pressure will remain below 110% of design.

Thus, there is no reduction in a margin of safety for overpressure

protection.

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

[[Page 28609]]

amendment request involves no significant hazards consideration.

Local Public Document Room location: Russell Library, 123 Broad

Street, Middletown, CT 06457.

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

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

CT 06141-0270.

NRC Project Director: Phillip F. McKee

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

Neck Plant, Middlesex County, Connecticut

Date of amendment request: March 7, 1996

Description of amendment request: The licensee will be replacing a

locally operated (manual) containment sump suction isolation valve, RH-

V-808A, with a remote manually operated (motor operated) valve, RH-MOV-

808A during the upcoming Cycle 19 refueling outage. As a result,

changes are being requested to the Haddam Neck Plant Technical

Specifications to reflect this design change.

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 technical specification change to Section 3/4.4.6.2

and its bases are the replacement of the designation RH-V-808A with

RH-MOV-808A. There are no changes to the requirements of this

specification and this change is therefore an administrative change.

The changes to Section 3/4.5.1 will make the requirements for RH-

MOV-808A identical to those of RH-MOV-22. RH-V-808A is being

converted to a motor operated valve (MOV). This MOV will make the

ability to establish a suction path from the containment to the

Residual Heat Removal (RHR) System single failure proof from the

control room. Both RH-MOV-22 and RH-MOV-808A will be opened to

establish containment sump recirculation post-loss of coolant

accident (LOCA). This will provide added assurance that core cooling

will be maintained in the switch from injection to containment sump

recirculation following a LOCA. The requirement for RH-MOV-808A to

be closed and its hand wheel locked can not cause an accident. The

credit for operation of RH-MOV-808A to ensure that the establishment

of containment sump recirculation is single failure proof is

equivalent to the current crediting of RH-V-808A with the only

difference being that operation of the valve can now be performed

from the control room. Also, since both RH-MOV-22 and RH-MOV-808A

will be procedurally opened during establishment of containment sump

recirculation, the elimination of the requirement to lock open the

breaker for RH-MOV-22 will not affect the consequences of a LOCA.

The proposed changes that reflect the conversion of RH-V-808A to a

MOV and the proposed changes in how the valve is used do not

increase the consequences of a LOCA.

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

new or different kind of accident from any previously evaluated.

The proposed changes will require RH-MOV-808A to be closed with

the hand wheel locked. This provides assurance that the valve is in

the required position. Also, RH-MOV-808A will be capable of remote

manual operation during the monthly surveillance which provides

assurance that the valve can be repositioned if necessary. The

proposed opening of RH-MOV-808A at the same time as RH-MOV-22 is

opened, provides greater assurance that a suction path is available

to the RHR pumps as well as lowering the total effective piping

resistance from the containment sump to the pump suction. Therefore,

the proposed changes do not introduce the possibility of a new or

different kind of accident.

3. [The proposed change does not involve] a significant

reduction in a margin of safety.

The proposed changes make RH-MOV-808A identical to RH-MOV-22

with the exception that RH-MOV-808A will not get a closure signal on

Safety Injection Actuation. Both RH-MOV-22 and RH-MOV-808A are

containment isolation valves in a closed system. For closed systems,

the containment isolation requirement is that the valves be either:

a) automatic, b) locked closed, or c) capable of remote manual

operation. RH-MOV-808A and RH-MOV-22 are both capable of remote

manual operation and therefore do not need automatic closure when

they are opened as part of the technical specification required

surveillance. Therefore, the proposed changes can not cause a

significant reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Russell Library, 123 Broad

Street, Middletown, CT 06457.

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

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

CT 06141-0270.

NRC Project Director: Phillip F. McKee

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

Neck Plant, Middlesex County, Connecticut

Date of amendment request: March 28, 1996

Description of amendment request: The proposed license amendment

will add an additional footnote to Limiting Condition for Operation

(LCO) 3.4.2.1 and revise an existing footnote for LCO 3.4.2.2.

Currently, the footnote for LCO 3.4.2.2 requires the pressurizer code

safety valve as-found lift setting to be within +3 percent and -1

percent of the setpoint. The proposed change will relax the negative

as-found lift tolerance to -3 percent. The as-left lift tolerance will

remain as plus or minus 1 percent. The same footnote will be added to

LCO 3.4.2.1.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. [The proposed change does not involve] a significant increase

in the probability or consequences of an accident previously

evaluated.

The proposed change will relax the pressurizer safety valve

negative as-found lift tolerance to -3 percent. The as-left lift

tolerance will remain as plus or minus 1 percent. This proposed

technical specification change will allow for the full use of the

plus or minus 3 percent as-found acceptance criterion for valve

testing consistent with 1989 ASME Section XI, Subsection IWV. The

relaxing of the as-found lift tolerance can not cause an accident.

The relaxing of the tolerance will allow the safety valve setpoint

to be closer to the Power Operated Relief Valve (PORV) setpoint and

could result in a slightly lower pressure for overheating events.

The analysis that takes credit for the increase in pressure to the

PORV setpoint is the Loss of Load analysis. The minimum departure

from nucleate boiling ratio (DNBR) was reanalyzed without taking any

credit for the transient increase in pressure. The minimum DNBR

still remains above the acceptance criterion as well as above the

limiting minimum DNBR predicted for all Updated Final Safety

Analysis Report Chapter 15 accidents. Also, the relaxed tolerance in

conjunction with a lower safety valve blowdown, yet still

conservative, results in a slightly higher average pressure for a

valve lift/reset cycle. This means that pressurizer overfill will

not be predicted for the limiting transient, loss of feedwater.

Thus, the proposed relaxation of as-found lift tolerance does not

increase the probability or consequences of the design basis

accidents previously evaluated.

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

new or different kind of accident from any previously analyzed.

The proposed relaxation of the lift tolerance still requires the

safety valve lift setpoint to be above both the PORV setpoint and

the pressurizer high pressure reactor trip setpoint. In addition,

the as-left setpoint is not being changed. The relaxed tolerance in

combination with a conservative safety valve blowdown still will

preclude the prediction of water relief from the pressurizer. This

means that the proposed change does not introduce the possibility of

a new or different kind of accident.

[[Page 28610]]

3. [The proposed change does not involve] a significant

reduction in a margin of safety.

The proposed relaxation of the as-found lift tolerance for valve

testing is consistent with 1989 ASME Section XI, Subsection IWV. The

as-left lift tolerance will remain plus or minus 1 percent. In

addition, the design basis analyses still meet their acceptance

criteria with the -3 percent lift tolerance. Therefore, the proposed

change can not cause 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.

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

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

Neck Plant, Middlesex County, Connecticut

Date of amendment request: April 16, 1996

Description of amendment request: The licensee is proposing to

revise the Technical Specifications to permit the Haddam Neck Plant to

remain in Mode 1, 2, 3, or 4 with the average water temperature of the

ultimate heat sink (UHS) greater than 90 deg. additional action has

been added which would require the plant to be placed in at least Hot

Standby within 6 hours and in Cold Shutdown within the following 30

hours upon identifying that the average water temperature of the UHS is

greater than 95 deg.F. In addition, the licensee is proposing to

include a new surveillance requirement for monitoring the average

circulating water inlet temperature to be within its limits when the

average water temperature of the UHS exceeds 89 deg.F.

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 addition to the Action Statement of LCO 3.7.12 of

an 8 hour period to monitor the average water temperature of the UHS

does not involve an increase in the probability of an accident

previously evaluated. The probability of an accident previously

evaluated is not increased by a short-term increase in the average

water temperature of the UHS. An evaluation of the service water

loads associated with the loss-of-offsite power and a coincident

worst case single failure of a diesel generator to start (resulting

in the loss of two of the four service water pumps) determined that

there is adequate margin to accomplish plant cooldown at a service

water inlet temperature of 95 deg.F. The recirculation phase of a

LOCA [loss-of-coolant accident] was evaluated to verify that

adequate flow would be available to the RHR [residual heat removal]

heat exchangers. The most limiting assumptions for the recirculation

phase are offsite power is available and one RHR heat exchanger

service water isolation valve fails to open. The injection phase of

a LOCA was evaluated to verify that adequate flow would be available

to the CAR [containment air recirculation] fan cooling coils. The

most limiting assumption for the injection phase is a loss-of-

offsite power. The results of these evaluations determined that

there is adequate service water flow to accomplish plant cooldown

with average water temperature of the UHS up to 95 deg.F. CYAPCO

[Connecticut Yankee Atomic Power Company] also proposes to include

an additional surveillance requirement to monitor the average water

temperature of the UHS at least once per hour if the average water

temperature of the UHS exceeds 89 deg.F. This additional

surveillance requirement ensures increased operator awareness as the

average water temperature of the UHS approaches the 90 deg.F LCO

limit. Based on the above, there is no significant increase in the

consequences of any accident previously evaluated.

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

new or different kind of accident from any accident previously

evaluated.

The proposed technical specification changes do not create the

possibility of a new or different kind of accident from those

previously evaluated. The addition of an 8 hour time period to

monitor the average water temperature of the UHS increases from 6 to

14 hours the amount of time that is allowed before the plant must

proceed to Hot Standby should the average water temperature of the

UHS increase above 90 deg.F. This extension of the time allowed for

the plant to be in Hot Standby does not change the plant

configuration. CYAPCO also proposes to include an additional

surveillance requirement to monitor the average water temperature of

the UHS at least once per hour if the average water temperature of

the UHS exceeds 89 deg.F. This additional surveillance requirement

ensures increased operator awareness as the average water

temperature of the UHS approaches the 90 deg.F LCO limit.

As such, the changes do not create the possibility of a new or

different kind of accident from those previously evaluated.

3. [The proposed change does not] involve a significant

reduction in a margin of safety.

The proposed technical specification changes do not involve a

significant reduction in any margin of safety. The addition of an 8

hour time period to monitor the average water temperature of the UHS

increases from 6 to 14 hours the time required before the plant must

proceed to Hot Standby should the average water temperature of the

UHS temperature [exceed] 90 deg.F. An evaluation has been performed

to demonstrate that the risk significance associated with the

increased action time is very low. In addition, safe shutdown

capability has been demonstrated for service water inlet

temperatures as high as 95 deg.F. The addition of a surveillance

requirement to monitor the average water temperature of the UHS at

least once per hour if the average water temperature of the UHS

exceeds 89 deg.F is an additional requirement, limitation, or

restriction not currently within the technical specifications.

Therefore, these 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

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

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

Neck Plant, Middlesex County, Connecticut

Date of amendment request: April 22, 1996

Description of amendment request: The proposed amendment will allow

the use of the performance-based containment leakage testing

requirements described in 10 CFR Part 50, Appendix J, Option B.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Involve a significant increase in the probability or

consequences of an accident previously analyzed.

The changes involved in this license amendment request revise

the testing criteria for the containment penetrations. The revised

criteria will be based on the guidance in Regulatory Guide 1.163,

``Performance-Based Containment Leak-Test Program.'' This guidance

allows for the use of relaxed testing frequencies for containment

penetrations that have performed satisfactorily on a historical

basis. The Containment Leakage Rate Testing Program considers the

type of service, the design of the penetration, and the safety

impact of the penetration in determining the

[[Page 28611]]

testing interval of each penetration. The NRC Staff has reviewed the

potential impact of performance-based testing frequencies for

containment penetrations during the development of the Option B

regulation. The NRC Staff review is documented in NUREG-1493,

``Performance-Based Containment Leakage-Test Program.'' The review

concluded that reducing the frequency of Type A tests (Integrated

Leakage Rate Tests) from three per 10 years to one per 10 years

leads to an imperceptible increase in risk. For Type B and C testing

(Local Leakage Rate Tests), the change in testing frequency should

not have significant impact since this leakage contributes less than

0.1 percent of the overall risk based on the existing regulations.

The use of Option B will allow the extension of testing intervals

with a minimal impact on the radiological release rates since most

penetration leakage is continually well below the specified limits.

In the accident risk evaluation, the NRC Staff noted that the

accident risk is relatively insensitive to the containment leakage

rate because the accident risk is dominated by accident sequences

that result in failure of or bypass of the containment. The use of a

performance-based testing program will continue to provide assurance

that the accident analysis assumptions remain bounding. Therefore,

the proposed change does not involve a significant increase in the

probability or consequences of an accident previously analyzed.

Removal of the surveillance accuracy requirement in Section

4.6.1.2.c will not affect the probability of an accident previously

analyzed since a similar requirement is contained in ANSI/ANS-56.8-

1994, ``Containment System Leakage Testing Requirements.'' ANSI/ANS-

56.8-1994 will be used to develop the technical methods and

techniques for the Containment Leakage Rate Test Program as stated

in Regulatory Guide 1.163. The technical methods and techniques in

ANSI/ANS-56.8-1994 have been determined to be acceptable to the NRC

Staff.

Changes to the Administrative Section describe the containment

testing program only and cannot increase the probability or

consequences of an accident previously analyzed.

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

from any previously analyzed.

The proposed license amendment does not change the operation or

equipment of the plant. The change in the test frequency is

dependent on the establishment of a Containment Leakage Test

Program. This test program will ensure the performance history of

each penetration is satisfactory prior to the changing of any test

frequency. Since the performance history of the penetration will be

known, there is no possibility of the implementation of the program

creating a new or different kind of accident than previously

analyzed. Since there is no change to the equipment or the operation

of the plant, there is no possibility of creating a new or different

kind of accident than previously analyzed. Therefore, the proposed

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

accident from any previously analyzed.

Removal of the surveillance accuracy requirement in Section

4.6.1.2.c will not create the possibility of a new or different kind

of accident from those previously analyzed since a similar

requirement is contained in ANSI/ANS-56.8-1994, ``Containment System

Leakage Testing Requirements.'' ANSI/ANS-56.8-1994 will be used to

develop the technical methods and techniques for the Containment

Leakage Rate Test Program as stated in Regulatory Guide 1.163. The

technical methods and techniques in ANSI/ANS-56.8-1994 have been

determined to be acceptable to the NRC staff.

Changes to the Administrative Section describe the containment

testing program only and cannot create a different accident from any

previously analyzed.

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

During the development of 10 CFR Part 50, Appendix J, Option B,

the NRC staff determined the reduction in safety associated with the

implementation of the performance-based testing program. The results

of this review are documented in NUREG-1493. The review concluded

that reducing the frequency of Type A tests (Integrated Leakage Rate

Tests) from three per 10 years to one per 10 years leads to an

imperceptible increase in risk. For Type B and C testing (Local

Leakage Rate Tests), the increase in testing frequency should not

have significant impact since this leakage contributes less than 0.1

percent of the overall risk based on the existing regulations. The

use of Option B will allow the extension of testing intervals with a

minimal impact on the radiological release rates since most

penetration leakage is continually well below the specified limits.

In the accident risk evaluation, the NRC Staff noted that the

accident risk is relatively insensitive to the containment leakage

rate because the accident risk is dominated by accident sequences

that result in failure of or bypass of the containment. The use of a

performance based testing program will continue to provide assurance

that the accident analysis assumptions remain bounding. Therefore,

this change does not involve a significant reduction in the margin

of safety.

Removal of the surveillance accuracy requirement in Section

4.6.1.2.c will not involve a significant reduction in the margin of

safety since a similar requirement is contained in ANSI/ANS-56.8-

1994, ``Containment System Leakage Testing Requirements.'' ANSI/ANS-

56.8-1994 will be used to develop the technical methods and

techniques for the Containment Leakage Rate Test Program as stated

in Regulatory Guide 1.163. The technical methods and techniques in

ANSI/ANS-56.8-1994 have been determined to be acceptable to the NRC

Staff.

Changes to the Administrative Section describe the containment

testing program only and do not reduce the margin of safety.

Moreover, the Commission has provided guidance concerning the

application of standards in 10 CFR 50.92 by providing certain

examples (51 FR 7751, March 6, 1986) of amendments that are

considered not likely to involve an SHC [significant hazards

consideration]. Although the proposed change is not enveloped by a

specific example, it has been shown that the proposed change is not

an SHC.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Russell Library, 123 Broad

Street, Middletown, CT 06457.

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

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

CT 06141-0270.

NRC Project Director: Phillip F. McKee

Consumers Power Company, Docket No. 50-255, Palisades Plant, Van

Buren County, Michigan

Date of amendment request: February 6, 1996

Description of amendment request: The proposed amendment would

delete the requirement to perform additional operability testing of

safety system train components when a required component in the

redundant train becomes inoperable.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Do the proposed changes involve a significant increase in the

probability or consequences of an accident previously evaluated?

The proposed changes remove the requirement for testing which is

in addition to the normal surveillance interval. The affected

equipment is subject to periodic surveillance testing required by

the Technical Specifications. Removing the requirement for

additional testing cannot alter any plant operating conditions,

operating practices, equipment settings, or equipment capabilities.

Therefore, changing an AOT [allowable outage time] or a surveillance

interval cannot increase the probability or consequences of an

accident previously evaluated.

2. Do the proposed changes create the possibility of a new or

different kind of accident from any previously evaluated?

The proposed changes remove the requirement for testing which is

in addition to the normal surveillance interval. The affected

equipment is subject to periodic surveillance testing required by

the Technical Specifications. Removing the requirement for

additional testing cannot alter any plant operating conditions,

operating practices, equipment settings, or equipment capabilities.

Therefore, changing an AOT or a surveillance interval cannot create

the possibility of a new or different

[[Page 28612]]

kind of accident from any previously evaluated.

3. Do the proposed changes involve a significant reduction in a

margin of safety?

The proposed changes remove the requirement for testing which is

in addition to the normal surveillance interval, in effect extending

the surveillance interval. An excessive surveillance interval

extension could reduce the margin of safety by reducing assurance

that required equipment will function as designed; an overly

restrictive surveillance interval could also reduce the margin of

safety by imposing unnecessary testing wear, equipment

manipulations, and system transients on the plant.

The existing requirements to perform cross-train testing were

based on the operating experience available when they were added to

the TS. Typically this was done during the initial plant licensing

in 1971. The recently published Standard Technical Specifications

(NUREG 1432) do not include cross-train testing requirements for the

Engineered Safety Features components. It has been judged by the NRC

and by the industry, that cross-train testing is unnecessary, and

that testing at normal surveillance intervals is adequate to assure

equipment operability. This recent judgment is based on a much

larger accumulation of operating experience than was available at

the time Palisades was licensed. There are no special features of

the Palisades plant which would invalidate these more recent

judgments of optimal testing requirements. Therefore, operation of

the facility in accordance with the proposed changes 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: Van Wylen Library, Hope

College, Holland, Michigan 49423

Attorney for licensee: Judd L. Bacon, Esquire, Consumers Power

Company, 212 West Michigan Avenue, Jackson, Michigan 49201

NRC Project Director: Mark Reinhart

Duke Power Company, Docket Nos. 50-369 and 50-370, McGuire Nuclear

Station, Units 1 and 2, Mecklenburg County, North Carolina

Date of amendment request: December 14, 1995, as supplemented by

letter dated May 16, 1996.

Description of amendment request: The proposed amendments would

change the Technical Specifications (TS) to improve the TS Action

Statements and Surveillance Requirements for diesel generators in

accordance with the recommendations and guidance in Generic Letter 93-

05, Generic Letter 94-01, NUREG-1366, and NUREG-1431. The proposed

amendments would also incorporate technical and administrative changes.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

Criterion 1

Operation of the facilities in accordance with the requested

amendments will not involve a significant increase in the

probability or consequences of an accident previously evaluated.

Improvements to the LCOs [limiting condition for operation] and

surveillance requirements for the emergency diesel generators do not

affect their capability to provide emergency power to plant vital

instruments and safety related equipment. In fact, these

improvements make the diesel generators more reliable since they

significantly reduce the amount of wear and stress due to excessive

and unnecessary testing. The proposed monthly testing of the diesel

generator continues to ensure that the system is ready for service

when needed. The fast starts and fast loadings continue to ensure

that the timing and loading requirements for engineered safety

features actuation are met. The proposed changes do not affect any

of the design basis accident analyses previously evaluated.

Therefore, these proposed changes do not involve any increase in the

probability or consequences of any accident previously evaluated.

The proposed changes are fully consistent with the recommendations

and guidance contained in GL [Generic Letter] 93-05, GL 94-01,

NUREG-1366, NUREG-1431, and are compatible with plant operating

experience.

Criterion 2

Operation of the facilities in accordance with the requested

amendments will not create the possibility of a new or different

kind of accident from any accident previously evaluated. The

proposed changes in fact improve the reliability of the diesel

generators by eliminating unnecessary wear and stress. Improved

reliability decreases the failure probability which also decreases

the probability of an accident not previously evaluated. None of the

requested amendments increase the common mode failure probability

thus would not increase the chance of both EDG's [emergency diesel

generators] for a particular nuclear unit being out of service

simultaneously. The proposed changes are fully consistent with the

recommendations and guidance contained in GL 93-05, GL 94-01, NUREG-

1366, NUREG-1431, and are compatible with plant operating

experience.

Criterion 3

Operation of the facilities in accordance with the requested

amendments will not involve a significant reduction in a margin of

safety. The proposed monthly testing of the diesel generators

continues to ensure that the system is ready for service when

needed. The fast starts and fast loadings continue to ensure that

the timing and loading requirements for engineered safety features

actuation are met. The proposed changes improve the reliability of

the diesel generators. Implementation of the Maintenance Rule also

ensures continued reliability of the diesel generators. No margin of

safety is decreased as a result of these TS changes.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Atkins Library, University of

North Carolina, Charlotte (UNCC Station), North Carolina 28223

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

South Church Street, Charlotte, North Carolina 28242

NRC Project Director: Herbert N. Berkow

Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One,

Unit No. 1, Pope County, Arkansas

Date of amendment request: April 29, 1996

Description of amendment request: The proposed amendment relocates

several cycle specific operating parameters from the technical

specifications to the Core Operating Limits Report per Generic Letter

88-16. The parameters being relocated by this change include the

variable low reactor coolant system pressure trip (VLPT) and the

variable low reactor coolant system pressure-temperature protective

limits.

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. Does Not Involve a Significant Increase in the

Probability or Consequences of an Accident Previously Evaluated.

The removal of the cycle-dependent variable low RCS pressure-

temperature protective limits and the VLPT setpoint from technical

speciications and placing them into the COLR has no impact on plant

safety and is considered to be administrative in nature. The

proposed change does not affect the safety analyses, physical

design, or operation of the plant. Technical specifications will

continue to require operation within the core protective and

operational limits for each reload cycle as calculated by the

approved reload design methodologies. The appropriate actions

required if limits are violated will remain in the technical

specifications. The reload report presents the results of cycle-

specific evaluations of accident analyses and transients addressed

in the ANO-1 Safety Analysis Report. The cycle-specific 10CFR50.59

evaluation of the reload

[[Page 28613]]

report demonstrates that changes in fuel cycle design and the

corresponding COLR do not involve a significant increase in the

probability or consequences of an accident previously evaluated.

Therefore, this change does not involve a significant increase

in the probability or consequences of any accident previously

evaluated.

Criterion 2. Does not Create the Possibility of a New or

Different Kind of Accident from any Previously Evaluated.

The proposed change to relocate the variable low RCS pressure-

temperature protective limits and the VLPT setpoint from the

technical specifications to the COLR is administrative in nature. No

change to the design configuration or method of operation of the

plant is made by this proposed change, and therefore, no new

transient initiator has been created. Technical specifications will

continue to require operation within the required core protective

and operating limits and appropriate actions will be taken if the

limits are exceeded. Because plant operation will continue to be

limited by the cycle-specific COLR limits that are established using

NRC-approved methodologies, these relocations will have no impact on

plant safety.

Therefore, this change does not create the possibility of a new

or different kind of accident from any previously evaluated.

Criterion 3. Does Not Involve a Significant Reduction in the

Margin of Safety.

Existing technical specification operability and surveillance

requirements are not reduced by the proposed change to relocate the

variable low RCS pressure-temperature protective limits and the VLPT

setpoint to the COLR. The proposed changes are administrative in

nature and do not relate to or modify the safety margins defined in

and maintained by the technical specifications. The cycle-specific

COLR limits for future reload fuel cycles will continue to be

developed based on NRC approved methodologies. Each future reload

undergoes a 10CFR50.59 evaluation to assure that operation of the

plant within the cycle-specific limits will not involve a

significant reduction in a margin of safety.

Therefore, this 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: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801

Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and

Strawn, 1400 L Street, N.W., Washington, DC 20005-3502

NRC Project Director: William D. Beckner

Entergy Operations, Inc., et al., Docket No. 50-416, Grand Gulf

Nuclear Station, Unit 1, Claiborne County, Mississippi

Date of amendment request: May 6, 1996

Description of amendment request: The amendment would reflect that

the name of Mississippi Power & Light Company (MP&L) has been changed

to Entergy Mississippi, Inc. The amendment revises Operating License

NPF-29 and Antitrust Conditions for the Grand Gulf Nuclear Station,

Unit 1 (GGNS) to (1) add the phrase ``(now renamed Entergy Mississippi,

Inc.)'' after the name of Mississippi Power & Light Company (MP&L), (2)

replace the name of Mississippi Power & Light Company (MP&L) by the

name Entergy Mississippi, Inc., and (3) replace a footnote by the

statement: ``Amendment ---- resulted in a name change for Mississippi

Power & Light Company (MP&L) to Entergy Mississippi, Inc.''.The

proposed amendment involves only a change in company name. It does not

involve any changes to the Technical Specifications for GGNS, or to any

requirements or limiting conditions for operation on any equipment or

any systems in the plant.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

Entergy Operations, Inc. proposes to change the current Grand

Gulf Nuclear Station Facility Operating License and Antitrust

Conditions. The specific proposed change is to reflect that the name

of one of the companies owning Grand Gulf Nuclear Station has

legally changed from Mississippi Power & Light Company to Entergy

Mississippi, Inc.

The Commission has provided standards for determining whether a

no significant hazards consideration exists as stated in 10 CFR

50.92(c). A proposed amendment to an operating license involves no

significant hazards consideration if operation of the facility in

accordance with the proposed amendment would not: (1) involve a

significant increase in the probability or consequences of an

accident previously evaluated; or (2) create the possibility of a

new or different kind of accident from any accident previously

evaluated; or (3) involve a significant reduction in a margin of

safety.

Entergy Operations, Inc. has evaluated the no significant

hazards consideration in its request for this license amendment and

determined that no significant hazards consideration results from

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

Inc. is providing the analysis of the proposed amendment against the

three standards in 10 CFR 50.92(c). A description of the no

significant hazards consideration determination follows:

I. The proposed change does not significantly increase the

probability or consequences of an accident previously evaluated.

The proposed change documents changing the legal name of the

company. The proposed change will not affect any other obligations.

The company will still own all of the same assets, serve the same

customers, and all existing obligations and commitments will

continue unaffected.

[The proposed change does not affect any of the existing

requirements or commitments on equipment or systems that are

designed for the safe operation of the plant. It does not affect the

design or operation of the plant.]

Therefore, the proposed change does not significantly increase

the probability or consequences of an accident previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

The administrative changes to the Operating License [and

Antitrust Condition] requirements [to change the name of Mississippi

Power & Light] do not involve any change in the design or operation

of the plant. The company will still own all of the same assets,

serve the same customers, and all existing obligations and

commitments will continue unaffected.

[The proposed changes do not affect equipment or systems that

could caused an accident at the plant.]

Therefore, the proposed changes do not create the possibility of

a new or different kind of accident from any accident previously

evaluated.

III. The proposed change does not involve a significant

reduction in a margin of safety.

The proposed change [in name] is administrative in nature, as

described above; therefore, this change does not reduce the level of

safety imposed by any current requirements. [The proposed changes do

not affect any equipment or systems at the plant.] The company will

still own all of the same assets, serve the same customers, and all

existing obligations and commitments will continue unaffected.

Therefore, the proposed changes do not cause a significant

reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room Location: Judge George W. Armstrong

Library, 220 S. Commerce Street, Natchez, MS 39120

Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and

Strawn, 1400 L Street, N.W., 12th Floor, Washington, DC 20005-3502

NRC Project Director: William D. Beckner

[[Page 28614]]

Entergy Operations, Inc., et al., Docket No. 50-416, Grand Gulf

Nuclear Station, Unit 1, Claiborne County, Mississippi

Date of amendment request: May 8, 1996

Description of amendment request: The amendment request would

replace the current frequency requirements in Surveillance Requirement

(SR) 3.6.1.3.5, on the leakage rate testing for each containment purge

valve with resilient seals, in the Technical Specifications for Grand

Gulf Nuclear Station, Unit 1 (GGNS). The proposed changes would place

these purge valves on a performance-based leakage testing frequency,

instead of the current once every 184 days and once within 92 days

after opening the valve.The proposed changes do not change the limiting

conditions for operation, the required actions for inoperability, or

the other surveillance requirements on these primary containment

isolation valves.

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 10 CFR 50.92, Entergy Operations, Inc. has

evaluated the proposed change to the Operating License of GGNS and

has determined that the operation of the facility in accordance with

the proposed amendment would not involve any significant hazards

considerations. In accordance with 10 CFR 50.91(a), Entergy

Operations, Inc. is providing the following analysis of the proposed

amendment against the three [following] standards of 10 CFR

50.92(c):

1) The proposed change does not significantly increase the

probability or consequences of an accident previously evaluated.

This change deletes the augmented testing requirement for these

containment isolation valves and allows the surveillance intervals

to be set in accordance with the Appendix J testing program.

[Appendix J to 10 CFR Part 50 defines primary containment leakage

testing requirements for water-cooled power reactors as GGNS and

these requirements include frequency of testing for the primary

containment isolation valves.] This change does not affect the

system function or design. The purge valves are not an initiator of

any previously analyzed accident. Leakage rates do not affect the

probability of the occurrence of any accident. Operating history has

demonstrated that these valves do not degrade and cause leakage as

previously anticipated. Because these valves have been demonstrated

to be reliable, these valves can be expected to perform the

containment isolation function as assumed in the accident analyses.

Therefore, there is no significant increase in the consequences

of any previously evaluated accident.

2) The proposed change would not create the possibility of a new

or different kind of accident from any accident previously

evaluated.

Extending the test intervals has no influence on, nor does it

contribute in any way to, the possibility of a new or different kind

of accident or malfunction from those previously analyzed. No change

has been made to the design, function or method of performing

leakage testing [or to the design and function of these valves].

Leakage acceptance criteria have not changed. No new accident modes

are created by extending the testing intervals. No safety-related

equipment or safety functions are altered as a result of this

change.

[Therefore, the proposed changes do not create the possibility

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

evaluated.]

3) The proposed change does not involve a significant reduction

in a margin of safety

The only margin of safety that has the potential of being

impacted by the proposed changes involves the offsite dose

consequences of postulated accidents which are directly related to

the containment leakage rate. The proposed change does not alter the

method of performing the tests nor does it change the leakage

acceptance criteria. Sufficient data has been collected to

demonstrate that the resilient seals do not degrade at an

accelerated rate.

[Also, the proposed change would test these valves in accordance

with the Appendix J testing program at the plant. Appendix J to 10

CFR Part 50 defines primary containment leakage testing requirements

for water-cooled power reactors as GGNS and these requirements

include frequency of testing for the primary containment isolation

valves.]

Because of this demonstrated reliability, this change will

provide sufficient surveillance to determine an increase in the

unfiltered leakage prior to the leakage exceeding that assumed in

the accident analysis.

Therefore, the proposed change does not result in a significant

reduction in a margin of safety.

Based on the above evaluation, Entergy Operation, Inc. has

concluded that operation in accordance with the proposed amendment

involves no significant hazards considerations.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room Location: Judge George W. Armstrong

Library, 220 S. Commerce Street, Natchez, MS 39120

Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and

Strawn, 1400 L Street, N.W., 12th Floor, Washington, DC 20005-3502

NRC Project Director: William D. Beckner

Entergy Operations, Inc., et al., Docket No. 50-416, Grand Gulf

Nuclear Station, Unit 1, Claiborne County, Mississippi

Date of amendment request: May 9, 1996

Description of amendment request: The amendment request would (1)

increase the safety limit minimum critical power ratio (MCPR) for two

loop operation and single loop operation to 1.10 and 1.11,

respectively, and (2) add a General Electric topical report to the list

of documents describing the analytical methods used to determine the

core operating limits. The proposed changes are to Section 2.1.1,

Reactor Core Safety Limits, and Section 5.6.5, Core Operating Limits

Report (COLR), respectively, of the Technical Specifications (TSs).

The licensee also proposed changes to the Bases of the TSs

associated with the above proposed changes.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

Entergy Operations, Inc. proposes to change the current Grand

Gulf Nuclear Station [GGNS] Technical Specifications. The specific

change is to modify the Minimum Critical Power Ratio (MCPR) safety

limits reported in Technical Specification 2.1.1.2, the list of

references in Technical Specification 5.6.5, and associated Bases

changes. The proposed change is necessary in order to switch reload

fuel vendors. [General Electric GE11 fuel is being added to the core

in place of Siemens Power Corporation (SPC) fuel.]

The Commission has provided standards for determining whether no

significant hazards considerations exists as stated in 10 CFR 50.92

(c). A proposed amendment to an operating license involves no

significant hazards if operation of the facility in accordance with

the proposed amendment would not: (1) involve a significant increase

in the probability or consequences of an accident previously

evaluated; (2) create the possibility of a new or different kind of

accident from any accident previously evaluated; or (3) involve a

significant reduction in a margin of safety.

Entergy Operations, Inc. has evaluated the no significant

hazards consideration in its request for this license amendment and

determined that no significant hazards considerations result from

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

Inc. is providing the analysis of the proposed amendment against the

three standards in 10 CFR 50.92(c). A description of the no

significant hazards consideration determination follows:

I. The proposed change does not significantly increase the

probability or consequences of an accident previously evaluated.

[[Page 28615]]

The Minimum Critical Power Ratio (MCPR) safety limit is defined

in the Bases to Technical Specification 2.1.1 as that limit which

``ensures that during normal operation and during Anticipated

Operational Occurrences (AOOs), at least 99.9% of the fuel rods in

the core do not experience transition boiling.'' The MCPR safety

limit is re-evaluated for each reload and, for GGNS [Operating]

Cycle 9, the analyses have concluded that a two-loop MCPR safety

limit of 1.10 based on the application of the generic GE MCPR

methodology is necessary to ensure that this acceptance criterion is

satisfied. For single-loop operation, a MCPR safety limit of 1.11

based on the generic GE MCPR methodology was determined to be

necessary. Core MCPR operating limits are developed to support the

Technical Specification 3.2 requirements and ensure these safety

limits are maintained in the event of the worst-case transient.

Since the MCPR safety limit will be maintained at all times,

operation under the proposed changes will ensure at least 99.9% of

the fuel rods in the core do not experience transition boiling.

Therefore, The Minimum Critical Power Ratio (MCPR) safety limit

change does not affect the probability or consequences of an

accident.

The implementation of GE's GESTAR-II approved methodology has no

effect on the probability or consequences of any accidents

previously evaluated. One exception to GESTAR is that the mis-

oriented and mis-located bundle events will continue to be analyzed

as accidents subject to the acceptance criteria in the current

licensing basis. The design of the GE11 fuel bundles is such that

the bundles are not likely to be mis-oriented or mis-located and the

normal administrative controls will be in effect for assuring proper

orientation and location. Therefore, the probability of a fuel

loading error is not increased. This analysis ensures that

postulated dose releases will not exceed a small fraction (10

percent) of 10CFR100 limits.

Therefore, the consequences of accidents previously evaluated

are unchanged.

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

or different kind of accident from any accident previously

evaluated.

The GE11 fuel to be used in [Operating] Cycle 9 is of a design

compatible with fuel present in the core and used in the previous

cycle. Therefore, the GE11 fuel will not create the possibility of a

new or different kind of accident. The proposed changes do not

involve any new modes of operation, any changes to setpoints, or any

plant modifications. They introduce revised MCPR safety limits that

have been proved to be acceptable for Cycle 9 operation. Compliance

with the applicable criterion for incipient boiling transition

continues to be ensured. The proposed MCPR safety limits do not

result in the creation of any new precursors to an accident.

Therefore, the proposed changes do not create the possibility of

a new or different type of accident from any accident previously

evaluated.

III. The proposed change does not involve a significant

reduction in a margin of safety.

The MCPR safety limits have been evaluated to ensure that during

normal operation and during AOOs [abnormal operating occurrences],

at least 99.9% of the fuel rods in the core do not experience

transition boiling. Therefore, the implementation of the proposed

changes in the MCPR safety limit ensure there is no reduction in the

margin of safety.

As with the current SPC methodology, GGNS will implement only

the NRC-approved revisions to GE's GESTAR methodology. This GE

methodology is similar to those SPC reports currently listed in TS

5.6.5 and it will be applied in a similar, conservative fashion. One

exception to GESTAR is that the mis-oriented and mis-located bundle

events will continue to be analyzed as accidents subject to the

acceptance criteria in the current licensing basis. This analysis

ensures that postulated dose releases will not exceed a small

fraction (10 percent) of 10CFR100 limits. On this basis, the

implementation of this GE methodology 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: Judge George W. Armstrong

Library, 220 S. Commerce Street, Natchez, MS 39120

Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and

Strawn, 1400 L Street, N.W., 12th Floor, Washington, DC 20005-3502

NRC Project Director: William D. Beckner

Entergy Operations Inc., Docket No. 50-382, Waterford Steam

Electric Station, Unit 3, St. Charles Parish, Louisiana

Date of amendment request: February 6, 1996

Description of amendment request: The proposed change will amend

the Allowable Values of parameters in Table 3.3-4 of Waterford Steam

Electric Station, Unit 3, (Waterford 3) Technical Specifications (TSs)

to make it consistent with the identical parameters in Table 2.2-1 of

TSs for Waterford 3. The proposed change will add Mode 4 to the

surveillance requirements of Table 4.3-2, Item 5.c (Safety Injection

System Automatic Actuation Logic) that was inadvertently removed.

Finally, the proposed change removes a reference to TS 3.3.3.2 in

Surveillance Requirements TS 4.10.2.2 and 4.10.4.2 since Incore

Detectors has been removed from the TSs.

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 described herein are administrative changes

necessary to correct administrative errors. The proposed changes

will have no affect on any design basis accidents nor will these

changes affect any material condition of the plant. Therefore, the

proposed changes will not involve a significant increase in the

probability or consequences of any accident previously evaluated.

The proposed changes are purely administrative. There are no new

system or design changes associated with this proposal. 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 impact on any protective

boundary, safety limit, or margin to safety. The proposed change

corrects inconsistencies in the TS and is purely administrative in

nature. 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

GPU Nuclear Corporation, et al., Docket No. 50-219, Oyster Creek

Nuclear Generating Station, Ocean County, New Jersey

Date of amendment request: May 7, 1996 (TSCR 247)

Description of amendment request: The proposed change to the

technical specifications would adopt the provisions of the Standard

Technical Specifications (STS), NUREG-1433, Rev. 1, which clarify

surveillance requirement applicability and allow a maximum period of 24

hours to complete a surveillance requirement upon discovery that the

surveillance has been missed.

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

[[Page 28616]]

involve a significant increase in the probability of occurrence or

consequence of an accident previously evaluated. The proposed

changes only affect administrative requirements regarding the

applicability and performance of surveillances. This change

clarifies surveillance requirement applicability and allows a

maximum 24 hour delay period for the performance of a surveillance

when it is discovered that the surveillance has not been performed

within the required frequency, consistent with the STS. There is

minimal safety significance associated with a delay of 24 hours in

completing the required surveillance, particularly due to the fact

that the most probable result of any particular surveillance

performed is the successful verification of conformance with the

requirements.

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 previously evaluated. The proposed changes

only affect administrative requirements regarding the applicability

of surveillance requirements and the performance of surveillances to

allow a maximum 24 hour delay period when it is discovered that a

surveillance has been missed. No changes to plant equipment or

operation are affected.

3. Operation of the facility in accordance with the proposed

amendment would not involve a significant reduction in the margin of

safety since the change contained in the proposed amendment does not

change any existing safety margins.

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: Ocean County Library,

Reference Department, 101 Washington Street, Toms River, NJ 08753

Attorney for licensee: Ernest L. Blake, Jr., Esquire. Shaw,

Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.

NRC Project Director: John F. Stolz

GPU Nuclear Corporation, Docket No. 50-320, Three Mile Island

Nuclear Station, Unit No. 2 (TMI-2), Dauphin County, Pennsylvania

Date of amendment request: February 16, 1995

Description of amendment request: The proposed amendment would

revise TMI-2 Operating License No. DPR-73 by modifying sections 4.02,

4.04, and 4.1.1.3 of the unit technical specifications. The revisions

to sections 4.02 and 4.04 would add flexibility to the scheduling of

surveillance activities and would allow for a 24 hour period to perform

missed surveillances before declaration of a limiting condition for

operation, respectively. These changes would make the TMI-2 technical

specifications consistent with the Standard Technical Specifications

for B&W Plants (NUREG-1430). The revision to section 4.1.1.3 would

allow extension of the time interval for surveillance of the

containment airlock doors from quarterly to annually. The proposed

changes to the TMI-2 technical specifications section 4.1.1.3 would

allow a decrease in worker exposure to radiation while maintaining an

adequate level of environmental protection at the facility.

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:

10 CFR 50.92 provides the criteria which the Commission uses to

perform a no significant hazards consideration. 10 CFR 50.92 states

that an amendment to a facility license involves no significant

hazards if operation of the facility in accordance with the proposed

amendment would not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated, or

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

from any accident previously evaluated, or

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

The proposed changes to the technical specifications sections

4.02 and 4.04 are administrative and do not involve any physical

changes to the facility. No changes are made to operating limits or

parameters, nor to any surveillance activities. The changes to

section 4.1.1.3 extends the interval between surveillance of the

containment airlocks; it does not change the operability

requirements, test methodology or acceptance criteria. Based on

this, GPU Nuclear has concluded that the proposed changes to

sections 4.02 and 4.04 do not:

1. Involve a significant increase in the probability of

occurrence or the consequences of an accident previously evaluated.

The changes do not modify any operating parameters or the release of

radioactive materials. The clarification of maximum time extensions

for surveillance is consistent with the NRC's Standard Technical

Specifications for Babcock and Wilcox Plants (NUREG-1430).

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

since these change are administrative and no plant configuration or

operational changes are involved.

3. Involve a change in the margin of safety. These changes are

administrative in nature, compatible with standard technical

specifications, and do not affect any safety settings or operational

limits.

GPU Nuclear has also concluded that the proposed changes to

section 4.1.1.3 do not:

1. Involve a significant increase in the probability of

occurrence of or consequences of an accident previously evaluated.

The change to this section does not change operating parameters,

equipment operability requirements, surveillance test methodology,

or acceptance criteria.

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

since the change does not affect plant equipment, plant

configuration, or plant operating parameters.

3. Involve a change in the margin of safety since the change

does not affect any operational limits.

Based on the above analysis the licensee concluded that the

proposed changes involve no significant safety hazards

considerations as defined by 10 CFR 50.92.

The NRC staff has reviewed the analysis of the licensee 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: Government Publications

Section, State Library of Pennsylvania, Walnut Street and Commonwealth

Avenue, Box 1601, Harrisburg, Pennsylvania 17105

Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw,

Pittman, Potts & Trowbridge, 2300 N Street, NW, Washington, DC 20037

NRC Project Director: Seymour H. Weiss

Houston Lighting & Power Company, City Public Service Board of San

Antonio, Central Power and Light Company, City of Austin, Texas,

Docket Nos. 50-498 and 50-499, South Texas Project, Units 1 and 2,

Matagorda County, Texas

Date of amendment request: May 1, 1996

Description of amendment request: The proposed amendments would

change the Technical Specifications to implement 10 CFR Part 50,

Appendix J, Option B, by referring to Regulatory Guide 1.163,

``Performance-Based Containment Leakage-Test Program.''

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:

[South Texas Project] STP has evaluated the proposed Technical

Specification Amendment and determined that it does not represent a

significant hazards consideration. Based on the criteria for

defining a significant hazards consideration established in 10 CFR

50.92, operation of STP in accordance with the proposed amendment

will not:

1) Involve a significant increase in the probability or

consequences of an accident previously evaluated because of the

following:

10 CFR [Part] 50, Appendix J has been amended to include

provisions regarding

[[Page 28617]]

performance based leakage testing requirements (Option B). Option B

allows plants with satisfactory Integrated Leak Rate Testing (ILRT)

performance history to extend the Type A testing interval from three

tests in ten years to one test in ten years. For Type B and Type C

tests, Option B allows extended testing interval[s] based on the

leak rate test history of each component. To be consistent with the

requirements of 10 CFR [Part] 50, Appendix J, Option B, STP proposes

to include appropriate changes to the Technical Specifications that

incorporate the necessary revisions associated with 10 CFR [Part]

50, Appendix J, Option B.

The proposed amendment represents the conversion of current

Technical Specification requirements to maintain consistency with

those requirements specified by 10 CFR [Part] 50, Appendix J, Option

B. The proposed changes are consistent with the current safety

analyses. Implementation of these changes will provide continued

assurance that specified parameters associated with containment

integrity will remain within acceptance limits, and will not

significantly increase the probability or consequences of a

previously evaluated accident.

Some proposed changes represent minor relaxations in current

Technical Specification requirements, but are based on the

requirements specified by Option B of 10 CFR [Part] 50, Appendix J.

Changes are consistent with the current safety analyses and

determined to represent sufficient requirements for the assurance

and reliability of equipment assumed to operate in the safety

analyses, and provide continued assurance that specified parameters

associated with containment integrity remain within their acceptance

limits. These changes will not significantly increase the

probability or consequences of a previously evaluated accident.

The systems affecting containment integrity related to this

proposed amendment request are not assumed in any safety analyses to

initiate any accident sequence. The probability of any accident

previously evaluated is not increased by this proposed amendment.

The proposed changes to Technical Specification LCOs or SRs maintain

an equivalent level of reliability and availability for all affected

systems. The proposed amendment does not increase the consequences

of any accident previously evaluated.

There is no change to the consequences of an accident previously

evaluated because maintaining leakage within the analyzed limit

assumed for any associated accident analyses does not adversely

affect either the on-site or off-site dose consequences resulting

from an accident. There is no adverse impact on the probability of

accident initiators. There is no significant increase in the

probability of any previously analyzed accident. A plant specific

risk-based analysis of Appendix J performed for STP indicates the

containment penetration leakage dose rate contribution to the total

dose rate in person-rem is insignificant.

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

from any accident previously evaluated because:

10 CFR [Part] 50, Appendix J, Option B specifies, in part, that

a Type A test which measures both the containment system overall

integrated leakage rate at containment pressure and system

alignments assumed during a large break LOCA [loss-of-coolant

accident], and demonstrates the capability of primary containment to

withstand an internal pressure load, may be conducted at an interval

based on the performance of the overall containment system. The

acceptable leakage rates are specified in the plant's Technical

Specifications. For Type B and Type C tests, intervals are proposed

based on the performance history of each component. Acceptance

criteria for each component is based upon demonstration that the sum

leakage rates at design basis pressure conditions for applicable

penetrations, is within the limit specified in the Technical

Specifications.

The proposed amendment represents the conversion of current

Technical Specification requirements to maintain consistency with

those requirements specified in 10 CFR [Part] 50, Appendix J, Option

B. The proposed changes are consistent with the current safety

analyses. Some minor relaxations in current Technical Specification

requirements, associated with containment integrity are based on

generic guidance provided in Option B, NEI 94-01 and ANSI/ANS 56.8,

1994. These changes do not involve revisions to the design of the

station. Some of the changes may involve revision in the testing of

components; however, these are in accordance with the STP current

safety analyses and provide for appropriate testing or surveillance

that are consistent with 10 CFR [Part] 50, Appendix J, Option B. The

proposed changes will not introduce new failure mechanisms beyond

those already considered in the current safety analyses.

The proposed amendment has been reviewed for acceptability

considering similarity of system or component design affecting

containment integrity. No new modes of operation are introduced by

the proposed changes. Surveillance requirements are changed to

reflect corresponding changes associated with Option B of 10 CFR

[Part] 50, Appendix J and improvements in technique or interval of

leak rate testing performance. The proposed changes maintain, at

minimum, the present level of operability of any system that affects

containment integrity. The proposed changes do not create the

possibility of a new or different kind of accident from any

previously evaluated.

The associated systems that affect leak rate integrity related

to the proposed amendment, are not assumed in any safety analysis to

initiate any accident sequence. The proposed surveillance

requirements for any affected systems are consistent with the

current requirements specified within the Technical Specifications

and are consistent with the requirements of Option B of 10 CFR

[Part] 50, Appendix J. The proposed surveillance requirements

maintain an equivalent level of reliability and availability of all

affected systems and therefore, does not increase the consequences

of any previously evaluated accident.

3) Involve a significant reduction in the margin of safety

because:

The provisions specified in Option B of 10 CFR [Part] 50

Appendix J allow changes to Type A, Type B, and Type C test

intervals based upon the performance of past leak rate tests. The

effect of extending containment leakage rate testing intervals has a

corresponding increase in the likelihood of containment leakage. The

degree to which intervals can be extended is a direct function of

the potential effect to existing safety margins and the public

health and safety that can occur due to an increased likelihood of

containment leakage. 10 CFR [Part] 50 Appendix J, Option B allows

longer intervals between leakage tests based on performance trends

but does not increase the leakage acceptance criteria. La [maximum

allowable leakage rate] is still limited to 0.3 wt%/day. By

referencing the Containment Leakage Rate Testing Program in LCO

3.6.1.2 ACTION, the point at which ACTION is required is increased

from .75 La to 1.0 La. This makes the specification consistent with

the intent of having margin between an AS-LEFT leakage of less than

or equal to .75 La and maintaining operability with less than or

equal to 1.0 La.

Changing Appendix J test intervals from those currently provided

in the Technical Specification to those provided in 10 CFR [Part]

50, Appendix J, Option B, slightly increases the risk associated

with Type A, Type B, and Type C specified accident sequences.

Historical data suggests that increasing the Type C test interval

can slightly increase the associated risk; however, this is

compensated by the corresponding risk reduction benefits associated

with reduction in component cycling, stress, and wear associated

with increased test intervals. When considering the total integrated

risk which includes all analyzed accident sequences, the risk

associated with increasing test intervals is negligible. A plant

specific risk-based analysis of Appendix J performed for STP

indicates the containment penetration leakage dose rate contribution

to total dose rate in person-rem is insignificant.

STP proposes to revise the Technical Specifications to be

consistent with those provisions specified in Option B of 10 CFR,

Appendix J. The proposed changes are consistent with the STP current

safety analyses. These proposed changes do not involve revisions to

the design of the station. The proposed changes will maintain the

same level of reliability of equipment associated with containment

integrity assumed to operate in the safety analysis, and provide

continued assurance that specified parameters affecting plant leak

rate integrity will remain within acceptance limits. The proposed

changes provide continued assurance of leakage integrity of

containment without adversely affecting the public health and safety

and will not significantly reduce existing safety margins. Plant

specific risk-based analysis indicates sufficient technical

justification exists to further extend the limits beyond those

allowed by Option B.

The proposed amendment to the Technical Specifications

implements present requirements, or the requirements in accordance

with the guidelines set forth in Option B of 10 CFR [Part] 50,

Appendix J. NUREG-1493, ``Performance-Based Containment Leak-Test

Program,'' served as the technical basis for Option B. STP

[[Page 28618]]

performed a plant specific risk-based analysis of containment

penetration leakage dose utilizing the same methodology used in

NUREG-1493. The analysis indicates the containment penetration

leakage dose rate contribution to the total dose rate in person-rem

is insignificant. This plant specific analysis serves to validate

the applicability of the proposed changes for STP. The proposed

changes have been approved by the NRC, are applicable to STP,

maintain necessary levels of system or component reliability

affecting containment integrity, and do not involve a significant

reduction in the margin of safety.

The performance-based approach to leakage rate testing concludes

the impact on public health and safety due to revised testing

intervals is negligible. The proposed amendment will not reduce

availability of systems associated with containment integrity when

required to mitigate accident conditions; therefore, the proposed

changes do not involve a significant reduction in the margin of

safety.

Guidance has been provided in ``Final Procedures and Standards

on No Significant Hazards Considerations,'' Final Rule, 51 FR 7744,

for the application of standards to license change requests for

determination of the existence of significant hazards

considerations. This document provides examples of amendments which

are and are not considered likely to involve significant hazards

considerations.

This proposed amendment does not involve a significant

relaxation of the criteria used to establish safety limits, a

significant relaxation of the bases for limiting safety system

settings or a significant relaxation of the bases for LCOs.

Therefore, based on the guidance provided in the Federal Register

and criteria established in 10 CFR 50.92(c), the proposed change

does not constitute a significant hazards consideration.

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

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

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

request for amendments involves no significant hazards consideration.

Local Public Document Room location: Wharton County Junior College,

J. M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488

Attorney for licensee: Jack R. Newman, Esq., Morgan, Lewis &

Bockius, 1800 M Street, N.W., Washington, DC 20036-5869

NRC Project Director: William D. Beckner

Northern States Power Company, Docket Nos. 50-282 and 50-306,

Prairie Island Nuclear Generating Plant, Unit Nos. 1 and 2, Goodhue

County, Minnesota

Date of amendment requests: December 14, 1995

Description of amendment requests: The proposed amendments would

revise the Administrative Control (Chapter 6) Section and other

affected Sections of the Prairie Island Technical Specifications to

generally conform with NUREG-1431, Standard Technical Specifications,

Westinghouse Plants, Revision 1, dated April 7, 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:

1. The proposed amendment will not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

Operation of the Prairie Island plant in accordance with the

proposed changes does not involve a significant increase in the

probability or consequences of an accident previously evaluated.

None of the proposed changes involve a physical modification to the

plant, a new mode of operation or a change to the Updated Safety

Analysis Report transient analyses. These proposed amendments

generally conform to the guidance of NUREG-1431, Revision 1, Section

5.0 which was previously reviewed, accepted and issued by the NRC.

Some Section 5.0 Specifications in NUREG-1431 were not

incorporated in this License Amendment Request. These Specifications

were not proposed because they 1) specify requirements not currently

in the Prairie Island Technical Specifications or otherwise

committed to, 2) are addressed elsewhere in the current Technical

Specifications, or 3) the current Technical Specifications level of

commitment is maintained. In all these instances, the NRC has

previously reviewed and approved the proposed level of commitment

through the issuance of the current Prairie Island Technical

specifications.

The proposed changes, in themselves, do not reduce the level of

qualification or training such that personnel requirements would be

decreased.

In total these changes do not involve a significant increase in

the probability or consequences of an accident previously evaluated.

2. The proposed amendment will not create the possibility of a

new or different kind of accident from any accident previously

analyzed.

The proposed changes do not create the possibility of a new or

different kind of accident from any accident previously evaluated

because the proposed changes, in themselves, do not introduce a new

mode of plant operation, surveillance requirement or involve a

physical modification to the plant. These proposed amendments

generally conform to the guidance of NUREG-1431, Revision 1, Section

5.0 which was previously reviewed, accepted and issued by the NRC.

Some Section 5.0 Specifications in NUREG-1431 were not

incorporated in this License Amendment Request. These Specifications

were not proposed because they 1) specify requirements not currently

in the Prairie Island Technical Specifications or otherwise

committed to, or 2) are addressed elsewhere in the current Technical

Specifications. Other features are not fully implemented but rather,

the current Technical Specification level of commitment is

maintained. In all these instances, the NRC has previously reviewed

and approved the proposed level of commitment through the issuance

of the current Prairie Island Technical Specifications.

In general, the proposed changes are administrative in nature.

The changes propose to revise, delete or relocate Specifications

within the Technical Specifications or from the Technical

Specifications to the Updated Safety Analysis Report, plant

procedures or the Operational Quality Assurance Plan through which

adequate control is maintained. The proposed changes do not alter

the design, function, or operation of any plant components and

therefore, no new accident scenarios are created.

Therefore, the possibility of a new or different kind of

accident from any accident previously evaluated would not be created

[by] these amendments.

3. The proposed amendment will not involve a significant

reduction in the margin of safety.

The proposed changes do not involve a significant reduction in a

margin of safety because the Current Technical Specifications

requirements for safe operation of the Prairie Island plant are

maintained or increased. The proposed changes are administrative in

nature and do not involve a physical modification to the plant, a

new mode of operation or a change to the Updated Safety Analysis

Report transient analyses. The proposed changes do not alter the

scope of equipment currently required to be operable or subject to

surveillance testing nor does the proposed change affect any

instrument setpoints or equipment safety functions.

Therefore, a significant reduction in the margin of safety would

not be involved with these amendments.

Based on the evaluation describe above, and pursuant to 10 CFR

Part 50, Section 50.91, Northern States Power Company has determined

that operation [of] the Prairie Nuclear Generating Plant in

accordance with the proposed license amendment request does not

involve any significant hazards considerations as defined by Nuclear

Regulatory Commission regulations in 10 CFR Part 50, Section 50.92.

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

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

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

amendment requests involve no significant hazards consideration.

Local Public Document Room location: Minneapolis Public Library,

Technology and Science Department, 300 Nicollet Mall, Minneapolis,

Minnesota 55401

Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts, and

Trowbridge, 2300 N Street, NW, Washington, DC 20037

[[Page 28619]]

NRC Project Director: Mark Reinhart (Acting Director)

Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323,

Diablo Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis

Obispo County, California

Date of amendment requests: February 15, 1996

Description of amendment requests: The proposed amendments would

revise the combined Technical Specifications (TS) for the Diablo Canyon

Nuclear Power Plant, Unit Nos. 1 and 2 to revise Technical

Specification 3.5.2, ``ECCS Subsystems - Tavg Greater Than or Equal to

350 deg.F,'' to change the allowed outage time for any one safety

injection pump from 72 hours to 7 days. The specific TS change proposes

to add a new footnote that increases the allowed outage time (AOT) for

one safety injection (SI) pump from 72 hours to 7 days for performance

of non-routine, emergent maintenance and requires review by the Plant

Staff Review Committee (PSRC), and requires Plant Manager approval

prior to exceeding 72 hours.

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 allowed outage time (AOT) extension does not change

the operating practices of Diablo Canyon Power Plant (DCPP).

Although the proposed change increases the allowed time in which the

safety injection (SI) system may be out of service for maintenance

or testing, this extended AOT will only be used in emergent

circumstances.

Increasing the AOT for the SI pumps does not involve physical

alteration of any plant equipment and does not affect analysis

assumptions regarding functioning of required equipment designed to

mitigate the consequences of accidents. Further, the severity of

postulated accidents and resulting radiological effluent releases

will not be affected by the increased AOT.

Finally, the probabilistic risk assessment determined that the

increase in the core damage probability is not considered

significant.

Therefore, the proposed 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 increase to the SI pump AOTs does not change the

method by which DCPP operates. Further, the proposed change would

not result in any physical alteration to any plant system, and there

would not be a change in the method by which any safety related

system performs its function.

Therefore, the proposed change does not create the possibility

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

evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

There is no safety analysis impact since the extension of the SI

pump AOT interval will have no effect on any safety limit,

protection system setpoint, or limiting condition of operation.

There is no hardware change that would impact existing safety

analysis acceptance criteria.

Therefore, the proposed change does not involve a significant

reduction in a margin of safety.

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

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

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

amendment requests involve no significant hazards consideration.

Local Public Document Room location: California Polytechnic State

University, Robert E. Kennedy Library, Government Documents and Maps

Department, San Luis Obispo, California 93407

Attorney for licensee: Christopher J. Warner, Esq., Pacific Gas and

Electric Company, P.O. Box 7442, San Francisco, California 94120

NRC Project Director: William H. Bateman

Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,

Callaway County, Missouri

]Date of application request: April 17, 1996

Description of amendment request: The proposed amendment would

change Technical Specification (TS) 3/4.3 to support a future

modification to replace existing digital portions of the main steam and

feedwater isolation system (MSFIS) with digital processor equipment and

would authorize revision of the FSAR to include a description of the

MSFIS modification. The MSFIS modification is a change to the facility,

as described in the safety analysis report, that involves an unreviewed

safety question. The modification involves an unreviewed safety

questions because: (1) the MSFIS design will use software which could

result in a common mode failure, (2) the original NRC review of the

MSFIS did not evaluate 2 out of 3 coincidence circuitry, which could

introduce new system failure modes, and (3) the MSFIS modification

utilizes manual handswitches that could introduce new system failure

modes. The NRC will review the modification in accordance with 10 CFR

50.59(a)(2) in conjunction with the review of the proposed TS

amendment.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, 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 addition of the MSFIS actuation logic and relays to the TS

has no adverse impact on the probability of occurrences or the

consequences of an accident. The proposed amendment does not change

or alter the design assumptions for the systems or components used

to mitigate the consequences of an accident and the methodologies

used in the accident analysis remain unchanged. The operating limits

will not be changed.

No design basis accidents will be affected by this design change

since the logic which currently exists will continue to be

performed. Thus, the radiological consequences will not change.

The system response time is enveloped by the current 5 second

valve stroke time. The MSFIS response time will be less than 500

msec.

A common mode software failure could exist if both separation

groups have their PLCs [programmable logic controllers] (3 per train

- six total) malfunction at the same time. However, a diverse means

of isolating the feedwater lines exists given the ability of the

Main Feed Control Valves to close on a Feedwater Isolation Signal.

The MSIVs [main steam isolation valves] do not have a diverse means

of isolating their respective steam lines if a common mode software

failure occurs. As a result, this modification provides a means to

manually fast close the valves at the MSFIS cabinets. The operators

will be alerted of the failure conditions of any PLC logic channel

via MCB [main control board] annunciators and indicators. This

failure mode has a low probability of occurrence based upon the

inherent quality of the design provided by the V&V [verification &

validation] process. Therefore, the accident consequences are not

increased for this failure mode.

The test panel in the MSFIS cabinets has been laid out to

provide the same functions as the existing test panel, except that

PLC status indication and coincidence logic test functions are

provided. The Emergency Override Panel, located below the Test

Panel, provides the operator with the ability to bypass the FWIS

[feedwater isolation signal] signal and manually fast close each

MSIV as required by the Emergency Operating Procedures. The MSIV

manual FC [fast close] switch operation is necessary for a diverse

means of operation for software common mode failures. The FWIS

bypass switch will allow main feedwater flow to be re-established to

each Steam Generator.

[[Page 28620]]

The replacement system is functionally the same as the current

system since it performs the same logic, receives the same inputs,

and produces the same outputs. However, the system is more reliable

and possesses triple redundant logic. Therefore, the probability of

malfunction will not be increased.

The electrical load of the A-B PLC equipment and existing 48 VDC

[volts direct current] actuation relays is less than that of the

existing equipment so the system will not require any additional

cooling over the existing equipment. Proper grounding is provided

for the PLC 5 VDC and actuation relay 48 VDC power supplies, which

are electrically isolated from each other.

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

or different kind of accident from any accident previously

evaluated.

The addition of the MSFIS actuation logic and relays to the TS

will not create a new type of accident or malfunction than any

previously evaluated in the Safety Analysis Report. The safety

functions of the system are not changed in any manner, nor is the

reliability of any structure, system or component reduced. All

design and performance criteria continue to be met. Since the safety

functions and reliability are not adversely affected, the proposed

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

accident from any accident previously evaluated.

The operator's ability to adequately respond to an accident is

not hindered by the man-machine interface added as a result of this

modification since the operator interface is similar to the current

system and the MCB controls will not change. The operators will be

alerted to system malfunctions through annunciation. The current

system has a status output for each MSIV and FIV [feedwater

isolation valve] valve on the Engineered Safety Feature Status

Panel, which will be maintained. In addition, an isolated plant

annunciator interface will provide a MSFIS Channel Failure plant

annunciator window for both trains. Training will be provided to the

technicians, engineers, and operators on the new features of the

system prior to installation. Therefore, this modification does not

increase the consequential effects due to the man-machine interface.

The system is compatible with the normal and accident

environments and will be seismically qualified in accordance with

the SNUPPS [standardized nuclear unit power plant system] seismic

spectra profile. The equipment will be qualified for Electromagnetic

Interference concerns in accordance with EPRI [Electric Power

Research Institute] document TR-102323-EPRI Guideline and will meet

the EPRI EMI [electromagnetic interference] limiting practices.

The system has the same failure mode upon loss of power as the

current system and behaves similarly upon power restoration. A loss

of power will not result in a MSFIS actuation.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The addition of the MSFIS actuation logic and relays to the TS

will not affect or change a safety limit or affect plant operations.

This change will not reduce the margin of safety assumed in the

accident analysis nor reduce any margin of safety as defined in the

basis for any TS.

The system response time for any given valve will not exceed the

required valve stroke time. Since the MSFIS does not contain any

analog channels, no channel trip accuracies are impacted.

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: Callaway County Public

Library, 710 Court Street, Fulton, Missouri 65251

Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts

& Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037

NRC Project Director: William H. Bateman

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 25, 1995

Description of amendment request: The proposed changes would

provide an allowed outage time of 14 days for the pressurizer power-

operated relief valve (PORV) nitrogen accumulators, as well as provide

separate action statements for the PORV depending on the reason for the

PORV inoperability during plant operation in power Modes 1, 2, or 3.

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.

Specifically, operation of North Anna Power Station in

accordance with the proposed Technical Specifications changes will

not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The PORVs are assumed to mitigate the consequences of a steam

generator tube rupture as described in the North Anna UFSAR [Updated

Final Safety Analysis Report] as well as to limit the undesired

opening of the pressurizer safety valves for a primary overpressure

event. The proposed action statements ensure that the steam

generator tube rupture accident analysis requirements are met. The

proposed Technical Specification changes require the backup nitrogen

supply be available for the PORVs to be consideredoperable and add

action statements and surveillance requirements for the nitrogen

supply commensurate with its significance. The proposed action

statements enhance the availability of the automatic actuation of

the PORVs by not requiring the block valves to be closed when the

backup nitrogen supplies are inoperable. The proposed surveillance

requirements enhance the reliability of the backup nitrogen supply

to the PORVs by verifying that there is sufficient nitrogen pressure

in the accumulators for the PORVs to perform their design function.

The proposed Technical Specification changes do not change any

accident analyses, therefore, the probability of any accident and

its resulting consequences are not increased.

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

from any accident previously evaluated.

The proposed Technical Specification changes do not involve any

physical modification to the plant or result in a change in a method

of operation. The backup nitrogen supply continues to be required

for PORV operability. The proposed Technical Specification changes

provide operational flexibility and ensure the availability of the

PORVs using the normal supply of instrument air while the backup

nitrogen supply is being restored. This also prevents undesirable

challenges to the pressurizer safety valves. The new surveillance

requirements verify that there is sufficient nitrogen pressure in

the accumulators for the PORVs to perform their design functions.

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

The proposed Technical Specification changes do not affect any

safety limits or limiting safety system settings. The availability

of the PORVs will be maintained as required in Generic Letter 90-06.

The proposed Technical Specifications will continue to ensure that

the PORVs will be capable of performing their intended functions.

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: 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 23219

NRC Project Director: Eugene V. Imbro

Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301,

Point Beach Nuclear Power Plant, Unit Nos. 1 and 2, Town of Two

Creeks, Manitowoc County, Wisconsin

Date of amendment request: April 24, 1996

[[Page 28621]]

Description of amendment request: The proposed amendments would

revise Technical Specification (TS) Section 15.7, ``Radiological

Effluent Technical Specifications (RETS).'' Portions of the RETS would

be moved to licensee-controlled documents consistent with Nuclear

Regulatory Commission guidance on TS improvements. Changes to other

sections of the TSs are also proposed consistent with the removal of

portions of the RETS.

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?

The proposed amendment simplifies the RETS and implements the

recommendations of GL 89-01 and of GL 95-10. The proposed change

relocates the operational requirements of RETS but keeps the

programmatic controls for these requirements in the Technical

Specifications. Therefore, the proposed changes are administrative

in nature and do not affect plant operations. Hence, the proposed

amendment does not involve a significant increase in the probability

or consequences of an accident previously evaluated because no

safety-related equipment, safety function, or plant operation will

be altered as a result of this proposed change. Also, the changes

are unrelated to the initiation and mitigation of accidents and

equipment malfunctions addressed in the Final Safety Analysis

Report.

2. Does the proposed amendment create the possibility of a new

or different kind of accident from any accident previously

evaluated?

As stated above, the proposed action is the relocation of the

RETS procedural details to various manuals while retaining the

administrative controls in RETS. The relocation is consistent with

the intent of the guidance of GL 89-01 and of GL 95-10. It is

administrative and has no impact on plant operation or safety. No

safety-related equipment, safety function, or plant operation will

be altered as a result of this proposed change. No changes to plant

components or structures are introduced which could create new

accidents or malfunctions not previously evaluated.

Therefore, the proposed changes will not create the possibility

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

evaluated because no new accident scenario is created and no

previously evaluated accident scenario is changed by the relocation

of the procedural details of RETS from one controlled document to

another.

3. Does the proposed amendment involve a significant reduction

in the margin of safety?

The proposed change does not include a change to any plant

structure, system, component, or operation. The proposed changes do

not alter the basic regulatory requirements and do not affect any

safety analyses. The proposed change is administrative. The

procedural details of the current RETS are relocated while the

programmatic controls consistent with regulatory requirements,

including controls on revisions to the manuals receiving the RETS

procedural details, the Environmental Manual (EM), Radiological

Effluent Control Program Manual (RECM), Offsite Dose Calculation

Manual (ODCM), and Process Control Program (PCP), remain in RETS.

Therefore, the proposed changes do not involve a significant

reduction in a margin of safety.

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

this review, it appears that the three standards of 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: Joseph P. Mann Library, 1516

Sixteenth Street, Two Rivers, Wisconsin 54241

Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts,

and Trowbridge, 2300 N Street, NW., Washington, DC 20037

NRC Project Director: Gail H. Marcus

Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301,

Point Beach Nuclear Power Plant, Unit Nos. 1 and 2, Town of Two

Creeks, Manitowoc County, Wisconsin

Date of amendment request: April 29, 1996

Description of amendment request: The proposed amendments would

revise Technical Specification (TS) Section 15.3.14, ``Fire Protection

System,'' and Section 15.4.15, ``Fire Protection System.'' These

specifications would be relocated to other licensee-controlled

documents in accordance with Nuclear Regulatory Commission generic

guidance. Additional administrative changes consistent with the

relocation are also proposed.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration which is presented below:

1. Operation of this facility under the proposed Technical

Specifications change will not increase the probability or

consequences of an accident previously evaluated.

This change request proposes to remove certain fire protection

program requirements from the Point Beach Technical Specifications

and incorporate them into the Final Safety Analysis Report (FSAR)

and the Fire Protection Evaluation Report (FPER). No requirements

are eliminated, modified, or de-emphasized by this change. The

proposed amendment ensures that any future changes to the fire

protection program will be subject to an appropriate evaluation in

accordance with NRC regulations to ensure that there are no

unreviewed safety questions.

Therefore, these proposed changes are administrative in nature.

There are no proposed changes to the physical plant or the processes

which ensure the plant's capability to mitigate fires and achieve

safe shutdown. Therefore, there is no potential effect on the

probability or consequences of previously evaluated accidents.

2. Operation of this facility under the proposed Technical

Specifications change will not create the possibility of a new or

different kind of accident from any accident previously evaluated.

New or different accidents can only be created by new or

different accident initiators or sequences. Because there are no

proposed changes to the physical plant or the processes which ensure

the plant's fire protection capability, new or different kinds of

accident initiators will not be introduced by this change. The

proposed changes are administrative in nature.

3. Operation of this facility under the proposed Technical

Specifications change will not create a significant reduction in a

margin of safety.

The margins of safety for Point Beach are based on the design

and operation of the reactor and containment and the safety systems

that provide their protection. Because there are no proposed changes

to the physical plant or the processes which ensure the plant's fire

protection capability, there will be no effect on the reactor,

reactor containment, or the safety systems which provide their

protection. Therefore, the proposed changes will not create a

reduction in a margin of safety. The proposed changes are

administrative in nature.

Additionally, the proposed revision to Point Beach's operating

license will not allow Wisconsin Electric to make changes to the

approved fire protection program without prior approval of the

Nuclear Regulatory Commission should these proposed changes

adversely affect the ability to achieve and maintain safe shutdown

in the event of a fire. In accordance with NRC Generic Letter 86-10,

any proposed change to the approved fire protection program requires

the performance of a 10 CFR 50.59 evaluation and a fire hazards

analysis. Should these evaluations indicate that the ability to

reach and maintain safe shutdown has been adversely affected, prior

NRC review and approval will be obtained prior to effecting the

changes. Thus, a significant reduction in a margin of safety cannot

occur.

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: Joseph P. Mann Library, 1516

[[Page 28622]]

Sixteenth Street, Two Rivers, Wisconsin 54241

Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts,

and Trowbridge, 2300 N Street, NW., Washington, DC 20037

NRC Project Director: Gail H. Marcus

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

Creek Generating Station, Coffey County, Kansas

Date of amendment request: May 16, 1996. This supersedes the

October 24, 1995, request published in the Federal Register on November

27, 1995 (60 FR 58409).

Description of amendment request: This license amendment request

proposes to revise Surveillance Requirement 4.7.6.e.4 to reflect a

proposed design change to the output rating, from 15kW to 5kW, of the

charcoal filter adsorber unit heater in the pressurization system

portion of the control room emergency ventilation system (CREVS).

Surveillance Requirements 4.7.6.c.2, 4.7.6.d, and 4.9.13.b and c, are

also being revised to reflect a proposed change to the acceptance

criteria for the testing of carbon samples from the CREVS charcoal

adsorbers and the auxiliary/fuel building emergency exhaust system

charcoal adsorbers. Surveillance Requirement 4.7.7.a for the auxiliary

building portion of the auxiliary/fuel building emergency exhaust

system is also affected by this proposed change. However, since

Surveillance Requirement 4.7.7.a refers to Surveillance Requirements

4.9.13.b and c, no changes to 4.7.7.a are required.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The design function of the filter adsorber unit heater in the

pressurization system portion of CREVS is to reduce the relative

humidity of the air entering the charcoal filter beds to 70%

relative humidity. Although the original design specified a heater

with a rating of 15 kW, review of the design basis calculation for

this system indicates that only about 3.13 kW is actually required

(including applicable margins to allow for voltage variations). The

proposed change to the CREVS heaters output rating from 15 kW to 5

kW will not affect the method of operation of the system, and the

new heater capacity will still exceed filter operational

requirements and safety margin. Neither the heater change nor the

charcoal testing protocol changes will affect system operation or

performance, nor do they affect the probability of any event

initiators. These changes do not affect any Engineered Safety

Features actuation setpoints or accident mitigation capabilities.

Therefore, the proposed changes will not significantly increase the

consequences of an accident or malfunction of equipment important to

safety previously evaluated in the USAR [Updated Safety Analysis

Report].

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

or different kind of accident from any accident previously

evaluated.

The requested change to the CREVS heaters' output rating and the

changes to the charcoal sample testing protocol will not affect the

method of operation of the systems, and the new heater capacity will

still exceed filter operational requirements and safety margin by a

significant amount. The proposed changes only affect the heater size

in the system and the testing criteria for the charcoal samples. No

new or different accident scenarios, transient precursors, failure

mechanisms, or limiting single failures will be introduced as a

result of these changes. Therefore, the possibility of a new or

different kind of accident other than those already evaluated will

not be created by this change.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The requested change to the CREVS heaters' output rating will

reduce the heater output of the system, but the new heater output

will still exceed filter operational requirements and safety margin

by a significant amount. In addition, the reduction in heat load

output from the heater will increase the design margin between the

cooling capacity of the system air conditioning units and the

building heat load. The new charcoal adsorber sample laboratory

testing protocol is more stringent than the current testing practice

and more accurately demonstrates the required performance of the

adsorbers following a design basis LOCA [loss-of-coolant accident].

Therefore, these changes will not reduce the margin of safety of the

HVAC [heating, ventilation, and air conditioning] systems'

operation.

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 locations: Emporia State University,

William Allen White Library, 1200 Commercial Street, Emporia, Kansas

66801 and Washburn University School of Law Library, Topeka, Kansas

66621

Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and

Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037

NRC Project Director: William H. Bateman

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.

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

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

Date of application for amendments: January 12, 1996, as

supplemented March 4, April 3 and April 10, 1996.

Brief description of amendments: The amendments revise the

Technical Specification so that the containment integrated leak rate

Type A testing will now be performed consistent with the revised 10 CFR

Part 50, Appendix J, Option B, by referring to Regulatory Guide 1.163,

``Performance-Based Containment Leak-Test Program.'' No

[[Page 28623]]

changes to implement Option B for the Type B and Type C tests were

requested by the licensee at this time.

Date of issuance: May 13, 1996

Effective date: As of the date of issuance to be implemented within

30 days

Amendment Nos.: 144 and 138

Facility Operating License Nos. NPF-35 and NPF-52: Amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: January 21, 1996 (61 FR

3498); and April 10, 1996 (61 FR 15988) The Commission's related

evaluation of the amendments is contained in a Safety Evaluation dated

May 13, 1996.No significant hazards consideration comments received: No

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

Black Street, Rock Hill, South Carolina 29730

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: March 5, 1996

Brief description of amendments: These amendments delete the

requirement to perform a pressurizer heater surveillance test and

change the requirement for containment visual inspection to prevent

sump clogging. These changes are in accordance with selected line items

from NRC Generic Letter 93-05, ``Line-Item Technical Specification

Improvements to Reduce Surveillance Requirements for Testing During

Power Operation.''

Date of issuance: May 13, 1996

Effective date: May 13, 1996

Amendment Nos. 184 and 178Facility Operating Licenses Nos. DPR-31

and DPR-41: Amendments revised the Technical Specifications.

Date of initial notice in Federal Register: April 10, 1996 (61

FR15989) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated May 13, 1996.No significant

hazards consideration comments received: No

Local Public Document Room location: Florida International

University, University Park, Miami, Florida 33199.

Houston Lighting & Power Company, City Public Service Board of San

Antonio, Central Power and Light Company, City of Austin, Texas,

Docket Nos. 50-498, South Texas Project, Unit 1, Matagorda County,

Texas

Date of amendment request: January 22, 1996, as supplemented by

letter dated April 18, 1996.

Brief description of amendment: The amendment modified the steam

generator tube plugging criteria in Technical Specification 3/4.4.5,

Steam Generators, and the associated Bases, to allow the implementation

of alternate steam generator tube plugging criteria for the tube-to-

tubesheet joints (known in the industry as F*) for Unit 1.

Date of issuance: May 14, 1996Effective date: May 14, 1996

Amendment No.: 82

Facility Operating License No. NPF-76: The amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: February 28, 1996 (61

FR 7553) The additional information contained in the supplemental

letter dated April 18, 1996, was clarifying in nature and thus, within

the scope of the initial notice and did not affect the staff's proposed

no significant hazards consideration determination.The Commission's

related evaluation of the amendments is contained in a Safety

Evaluation dated May 14, 1996. No significant hazards consideration

comments received: No

Local Public Document Room location: Wharton County Junior

College, J. M. Hodges Learning Center, 911 Boling Highway, Wharton, TX

77488

Rochester Gas and Electric Corporation, Docket No. 50-244, R. E.

Ginna Nuclear Power Plant, Wayne County, New York

Date of application for amendment: February 9, 1996, as

supplementedMarch 15, 1996,

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