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 January 5, 1996, through January 19, 1996.

The last biweekly notice was published on January 22, 1996.

Notice Of Consideration Of Issuance Of Amendments To Facility

Operating Licenses, Proposed No Significant Hazards Consideration

Determination, And Opportunity For A Hearing

The Commission has made a proposed determination that the following

amendment requests involve no significant hazards consideration. Under

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

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

involve a significant increase in the probability or consequences of an

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

or different kind of accident from any accident previously evaluated;

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

for this proposed determination for each amendment request is shown

below.

The Commission is seeking public comments on this proposed

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

publication of this notice will be considered in making any final

determination.

Normally, the Commission will not issue the amendment until the

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

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

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

the Commission may issue the license amendment before the expiration of

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

the amendment involves no significant hazards consideration. The final

determination will consider all public and State comments received

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

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

opportunity for a hearing after issuance. The Commission expects that

the need to take this action will occur very infrequently.

Written comments may be submitted by mail to the Rules Review and

Directives Branch, Division of Freedom of Information and Publications

Services, Office of Administration, U.S. Nuclear Regulatory Commission,

Washington, DC 20555, and should cite the publication date and page

number of this Federal Register notice. Written comments may also be

delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike,

Rockville, Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays.

Copies of written comments received may be examined at the NRC Public

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

The filing of requests for a hearing and petitions for leave to

intervene is discussed below.

By March 1, 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.

[[Page 3498]]

Not later than 15 days prior to the first prehearing conference

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

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

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

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

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

brief explanation of the bases of the contention and a concise

statement of the alleged facts or expert opinion which support the

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

contention at the hearing. The petitioner must also provide references

to those specific sources and documents of which the petitioner is

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

facts or expert opinion. Petitioner must provide sufficient information

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

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

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

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

petitioner who fails to file such a supplement which satisfies these

requirements with respect to at least one contention will not be

permitted to participate as a party.

Those permitted to intervene become parties to the proceeding,

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

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

hearing, including the opportunity to present evidence and cross-

examine witnesses.

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

determination on the issue of no significant hazards consideration. The

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

If the final determination is that the amendment request involves

no significant hazards consideration, the Commission may issue the

amendment and make it immediately effective, notwithstanding the

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

of the amendment.

If the final determination is that the amendment request involves a

significant hazards consideration, any hearing held would take place

before the issuance of any amendment.

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

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

Commission, Washington, DC 20555, Attention: Docketing and Services

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

the Gelman Building, 2120 L Street, NW., Washington DC, by the above

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

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

Commission by a toll-free telephone call to Western Union at 1-(800)

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

should be given Datagram Identification Number N1023 and the following

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

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

date and page number of this Federal Register notice. A copy of the

petition should also be sent to the Office of the General Counsel, U.S.

Nuclear Regulatory Commission, Washington, DC 20555, and to the

attorney for the licensee.

Nontimely filings of petitions for leave to intervene, amended

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

not be entertained absent a determination by the Commission, the

presiding officer or the Atomic Safety and Licensing Board that the

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

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

For further details with respect to this action, see the

application for amendment which is available for public inspection at

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

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

the particular facility involved.

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

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

Date of amendment request: January 12, 1996

Description of amendment request: Compliance with 10 CFR Part 50,

Appendix J, provides assurance that the primary containment, including

those systems and components that penetrate the primary containment, do

not exceed the allowable leakage rate values specified in the Technical

Specifications and Bases. The allowable leakage rate is determined so

that the leakage assumed in the safety analyses is not exceeded.

On February 4, 1992, the NRC published a notice in the Federal

Register (57 FR 4166) discussing a planned initiative to begin

eliminating requirements marginal to safety that impose a significant

regulatory burden. Appendix J to 10 CFR Part 50, ``Primary Containment

Leakage Testing for Water-Cooled Power Reactors,'' was considered for

this initiative and the staff undertook a study of possible changes to

this regulation. The study examined the previous performance history of

domestic containments and examined the effect on risk of a revision to

the requirements of Appendix J. The results of this study are reported

in NUREG-1493, ``Performance-Based Leak-Test Program.''

Based on the results of this study, the staff developed a

performance based approach to containment leakage rate testing. On

September 12, 1995, the NRC approved issuance of this revision to 10

CFR Part 50, Appendix J, which was subsequently published in the

Federal Register on September 26, 1995, and became effective on October

26, 1995. The revision added Option B ``Performance-Based

Requirements'' to Appendix J to allow licensees to voluntarily replace

the prescriptive testing requirements of Appendix J with testing

requirements based on both overall and individual component leakage

rate performance.

Regulatory Guide 1.163, ``Performance-Based Containment Leak Test

Program,'' was developed as a method acceptable to the staff for

implementing Option B. Accordingly, the licensee has submitted, in its

application dated January 12, 1996, proposed changes to the TS 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:

1The proposed change will not involve a significant increase in

the probability or consequences of an accident previously evaluated.

Containment leak rate testing is not an initiator of any

accident; the proposed change does not affect reactor operations or

accident analysis, and has no significant radiological consequences.

Therefore, this proposed change will not involve an increase in the

probability or consequences of any previously-evaluated accident.

2. The proposed change will not create the possibility of any

new accident not previously evaluated.

The proposed change does not affect normal plant operations or

configuration, nor does it affect leak rate test methods. The test

history at Catawba (no ILRT [integrated leak rate test] failures)

provides continued assurance of the leak tightness of the

containment structure.

3. There is no significant reduction in a margin of safety.

[[Page 3499]]

The proposed changes are based on NRC-accepted provisions, and

maintain necessary levels of reliability of containment integrity.

The performanced-based approach to leakage rate testing recognizes

that historically good results of containment testing provide

appropriate assurance of future containment integrity; this supports

the conclusion that the impact on the health and safety of the

public as a result of extended test intervals is negligible.

Based on the above, no significant hazards consideration is

created by the proposed change.

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

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

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

amendment request involves no significant hazards consideration.

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

Black Street, Rock Hill, South Carolina 29730

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

South Church Street, Charlotte, North Carolina 28242

NRC Project Director: Herbert N. Berkow

Duquesne Light Company, et al., Docket Nos. 50-334 and 50-412,

Beaver Valley Power Station, Unit Nos. 1 and 2, Shippingport,

Pennsylvania

Date of amendment request: December 27, 1995

Description of amendment request: The proposed amendments would

modify Tables 3.3-11 and 4.3-7 of Beaver Valley Power Station Unit Nos.

1 and 2 (BVPS-1 and BVPS-2) Technical Specification (TS) 3.3.3.8 such

that only one valve position indication system for the power operated

relief valves and safety valves is required to be operable. The

licensee stated that the proposed amendments would then be consistent

with the NRC's Improved Standard Technical Specifications, NUREG-1431,

Revision 1, and with the guidance of Regulatory Guide 1.97, NUREG-0578,

and NUREG-0737.

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 change involves instrumentation which is redundant

in monitoring the position of valves and, as such, does not

influence the potential for an initiating event involving the power

operated relief valves (PORVs) or the safety valves (SVs).

Implementation of these changes will reduce the potential for

challenges to the plant due to a potential shutdown which should not

be necessary due to the restrictive nature of having unnecessary

redundant position indication in the technical specification. By

deleting the Unit No. 1 technical specification operability

requirements for the PORV acoustic detectors, and by deleting, on

both units, the technical specification operability requirements for

the SV temperature detector position indicators, the potential for

unnecessary shutdowns is reduced. When inoperable, the PORV acoustic

detectors and the SV temperature detectors presently invoke an

unnecessary action statement as another fully qualified safety-

related position indication system exists to provide indication. The

proposed change modifies Specification 3.3.3.8 actions and

surveillance requirements, but does not affect the BASES.

The remaining instrumentation on these tables [3.3-11 and 4.3-7]

will be unaffected. The remaining position indication systems for

the PORVs and SVs are fully qualified and satisfy regulatory

criteria for post accident monitoring of valve position. These

changes do not affect the ability to satisfy analysis assumptions

regarding operation of the PORVs and SVs. They do not affect the

ability to continue to meet the guidance of Regulatory Guide 1.97,

the post Three Mile Island criteria contained in NUREG 0578 and

NUREG 0737, and reflect the guidance provided in NUREG 1431,

``Improved Standard Technical Specifications'' (ISTS). Therefore, we

have concluded that these changes do not involve a significant

increase in the probability or consequences of an accident

previously evaluated in the Updated Final Safety Analysis Report

(UFSAR).

2. Does the change create the possibility of a new or different

kind of accident from any accident previously evaluated?

The proposed change will reduce the potential to challenge

safety systems due to eliminating the potential for unnecessary

plant shutdowns. The proposed changes are limited to PORV and SV

position indication and do not involve any physical changes to the

PORVs or SVs or their setpoints. These changes do not delete any

design basis accident functions previously provided by the PORVs or

SVs nor has the probability of inadvertent opening been increased.

Accordingly, no new single failure has been identified as a result

of these changes. Therefore, these changes will not create the

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

previously evaluated in the UFSAR.

3. Does the change involve a significant reduction in a margin

of safety?

The proposed changes have been incorporated to eliminate a

degree of equipment redundancy and is consistent with the Improved

Standard Technical Specifications (ISTS). The Unit No. 1

specification presently requires operability of both redundant PORV

position indication systems and the primary and backup SV position

indication systems. The Unit No. 2 specification also requires

operability of the primary and backup SV position indication

systems. These changes will potentially eliminate some challenges

and potential unnecessary shutdowns by eliminating equipment

determined to be no longer necessary. Only one safety-related

position indication system is necessary to satisfy regulatory

criteria; therefore, operation of the plant in accordance with the

proposed amendment would not involve a significant reduction in a

margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: B. F. Jones Memorial Library,

663 Franklin Avenue, Aliquippa, Pennsylvania 15001.

Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman,

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

NRC Project Director: John F. Stolz

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

Linn County, Iowa

Date of amendment request: December 22, 1995

Description of amendment request: The proposed amendment would

revise the Duane Arnold Energy Center (DAEC) Technical Specifications

(TS) Sections 3.7.A and 4.7.A, ``Primary Containment,'' by deleting

information also contained in 10 CFR Part 50, Appendix J, Option A and

incorporating references to the Primary Containment Leakage Rate

Testing Program. These changes will allow the use of the performance

based option of containment leak testing. The request also adds

Operability and Surveillance Requirements (SRs) for the drywell air

lock. Minor administrative changes are also made. These changes are

consistent with comparable specifications in the Improved Standard

Technical Specifications (ITS), NUREG-1433. In addition to the

licensee's proposed revision to the DAEC TS, the staff will be

executing administrative changes and corrections to the TS Bases, as

submitted in letters(2) dated February 13, 1995. Sections that will be

changed or corrected are Section 1.2, Bases; Section 2.2, Bases Reactor

Coolant System Integrity; Section 3.2, Bases; Section 3.7.H/4.7.H,

Bases Containment Atmosphere Dilution; and Section 3.7.I/4.7.I, Bases

Oxygen Concentration.

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

[[Page 3500]]

consideration, which is presented below:

. The proposed revision does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

Information contained in 10 CFR 50, Appendix J was deleted and

references to the Primary Containment Leakage Rate Testing Program

were added. These are administrative changes to allow the use of

performance-based containment leakage testing methods. The

containment testing program will conform with the requirements of

Option B of 10 CFR Part 50, Appendix J and approved exemptions. The

performance of the leakage tests themselves is not an input or

consideration in any accident previously evaluated, thus the

proposed change will not increase the probability of any such

accident occurring. The same operability requirements remain for the

primary containment, therefore the consequences of an accident are

not significantly increased.

Drywell air lock operability and surveillance requirements were

added. Actions for one air lock door inoperable have been added

consistent with the ITS. In addition, notes have been added to allow

entry and exit to perform repairs of the air lock components and to

explain that the previous overall leak test is not invalidated by an

inoperable door. This change represents an additional restriction on

plant operation, since the previous condition of one air lock door

inoperable did not require any actions to be taken. A requirement to

verify proper operation of interlock mechanism was also added. This

will ensure that one door is always closed which maintains primary

containment integrity.

The addition of these new drywell air lock requirements provides

more stringent provisions than previously existed in the [current

Technical Specifications]. The more stringent requirements will not

result in operation that will increase the probability of initiating

an analyzed event. If anything, the new requirements may decrease

the probability or consequences of an analyzed event by

incorporating the more restrictive changes discussed above. These

changes will not alter assumptions relative to mitigation of an

accident or transient event. The more restrictive requirements will

not alter the operation of process variables, structures, systems,

or components as described in the safety analyses.

The TS revision includes the relocation of certain requirements

from the current technical specification (CTS) to licensee

controlled documents. CTS 4.7.A.1.e contains a requirement to

replace the T-ring inflatable seals for the 18 inch purge valves

every four years. This provision is not in the ITS as it is a

maintenance issue and not a surveillance for operability. CTS

4.7.A.1.e also contains a requirement to verify (during Type C

testing) that the mechanical modification which limits the maximum

opening angle for the 18 inch purge valves is intact. The ITS only

requires this surveillance if the mechanical modification is not

permanent. At DAEC, the 18 inch purge valves are permanently blocked

to restrict opening to 30 deg.. These CTS provisions will be

relocated to plant procedures. Any changes to these relocated

requirements will require an evaluation in accordance with 10 CFR

50.59. CTS 4.7.A.1.a and 4.7.A.1.d contain some procedural details

that are not contained in Appendix J. These details will also be

relocated to plant procedures, consistent with the ITS. Since any

changes to these licensee controlled documents will be evaluated in

accordance with 10 CFR 50.59, no significant increase in the

probability or consequences of an accident previously evaluated will

be allowed.

The proposed revision does not involve any change to the

configuration or method of operation of any plant equipment that is

used to mitigate the consequences of an accident, nor does it affect

any assumptions or conditions in the accident analysis. The proposed

revision does not degrade any existing plant programs, nor modify

any functions of safety related systems or accident mitigation

functions previously credited at the DAEC. The proposed changes do

not impact initiators of analyzed events. They also do not impact

the assumed mitigation of accidents or transient events. These TS

changes will not alter assumptions made in the safety analysis and

licensing basis.

Therefore, the proposed revision does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

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

new or different kind of accident from any accident previously

evaluated.

Deleting information from the TS which is contained in 10 CFR

50, Appendix J and adding references to the Primary Containment

Leakage Rate Testing Program are purely administrative changes to

allow the use of performance-based containment leakage testing

methods. The containment testing program will conform with the

requirements of Option B of 10 CFR Part 50, Appendix J and approved

exemptions. The use of Option B will maintain the containment safety

functions as a barrier to the release of radioactivity to the

environment.

The proposed revision does not make any physical or operational

changes to existing plant systems or components, nor does it alter

any plant parameters, revise any safety limit setpoint, or provide

any new release pathways. The proposed revision does not change any

transient responses assumed in the Design Bases of the plant.

The proposed changes which relocate requirements to licensee

controlled documents will not alter the plant configuration (no new

or different type of equipment will be installed) or change the

methods governing normal plant operation. These changes will not

alter assumptions made in the safety analysis or licensing basis.

The proposed changes which add more restrictive requirements to

the CTS will not alter the plant configuration (no new or different

type of equipment will be installed) or change the methods governing

normal plant operation. These changes do impose different

requirements. However, they are consistent with assumptions made in

the safety analyses.

Therefore, the revision does not create the possibility of a new

or different kind of accident previously evaluated.

3. The proposed revision will not significantly reduce any margin

of safety.

Deleting information from the TS which is contained in 10 CFR

50, Appendix J and adding references to the Primary Containment

Leakage Rate Testing Program do not involve a significant reduction

in the margin of safety. These changes are administrative in nature

and either eliminate a redundant requirement or clarify the

applicability and acceptability of an alternative, NRC approved,

leak rate testing provision within the TS. The containment testing

program will conform to the requirements of Option B of 10 CFR Part

50, Appendix J and approved exemptions. The use of Option B will

maintain the containment safety functions as a barrier to the

release of radioactivity to the environment.

The proposed revision does not require any modifications to

existing plant systems or equipment, safety limit settings, or

parameters utilized in the licensing bases for the safety analysis.

The proposed revision does not change any safety analysis or any

accident mitigation action for which DAEC has previously taken

credit. The proposed changes do not involve any technical changes;

they have no impact on any safety analysis assumptions. The addition

of new requirements either increases or does not affect the margin

of safety.

The proposed changes that relocate requirements from the CTS to

licensee controlled documents will not reduce a margin of safety

since they have no impact on any safety analysis assumptions. In

addition, the requirements to be relocated from the CTS to the

licensee controlled document are unchanged. Since any future changes

to this licensee controlled document will be evaluated in accordance

with the requirements of 10 CFR 50.59, no significant reduction in a

margin of safety will be allowed.

The proposed changes are consistent with NUREG-1433, which was

approved by the NRC Staff. The changes are also consistent with NRC

guidance provided for the implementation of Option B. The change

controls for proposed relocated details and requirements are

acceptable. Therefore, revising the TS to reflect the NRC accepted

level of detail and requirements ensures that there is no reduction

in a margin of safety.

Therefore, the proposed revision will not significantly reduce

any margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Cedar Rapids Public Library,

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

[[Page 3501]]

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

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

NRC Project Director: Gail H. Marcus

Illinois Power Company and Soyland Power Cooperative, Inc., Docket

No. 50-461, Clinton Power Station, Unit No. 1, DeWitt County,

Illinois

Date of amendment request: December 14, 1995

Description of amendment request: The proposed amendment would

modify Technical Specifications 3.3.1.1, ``Reactor Protection System

(RPS) Instrumentation,'' and 3.3.6.1, ``Primary Containment and Drywell

Isolation Instrumentation,'' to eliminate periodic response time

testing of selected analog trip modules (ATMs). This request is

supported by analyses prepared by the Boiling Water Reactor Owners'

Group topical report NEDO-32291, ``System Analyses for Elimination of

Selected Response Time Testing Requirements,'' which demonstrate that

other periodic tests required by technical specifications, such as

channel calibrations, channel functional tests and logic system

functional tests, are adequate to ensure ATM response times remain

within acceptable 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:

(1) The purpose of the proposed Technical Specification (TS)

change is to eliminate response time testing requirements for

selected analog trip modules (ATMs) in the Reactor Protection System

(RPS) and the main steam isolation valve (MSIV) isolation actuation

instrumentation. The Boiling Water Reactor Owners' Group (BWROG) has

completed an evaluation which demonstrates that response time

testing is redundant to the other TS-required testing. These other

tests, in conjunction with actions taken in response to NRC Bulletin

90-01, ``Loss of Fill-Oil in Transmitters Manufactured by

Rosemount,'' and Supplement 1, are sufficient to identify failure

modes or degradations in instrument response time and ensure

operation of the associated systems within acceptable limits. There

are no known failure modes that can be detected by response time

testing that cannot also be detected by other TS-required testing.

This evaluation was documented in NEDO-32291, ``System Analyses for

Elimination of Selected Response Time Testing Requirements,''

January 1994. Illinois Power (IP) has confirmed the applicability of

this evaluation to Clinton Power Station (CPS). In addition, IP has

completed the actions identified in the NRC staff's safety

evaluation of NEDO-32291.

Because of the continued application of other existing TS-

required tests such as channel calibrations, channel checks, channel

functional tests, and logic system functional tests, the response

time of these systems will be maintained within the acceptance

limits assumed in plant safety analyses and required for successful

mitigation of an initiating event. The proposed changes do not

affect the capability of the associated systems to perform their

intended function within their required response time, nor do the

proposed changes themselves affect the operation of any equipment.

As a result, IP has concluded that the proposed changes do not

involve a significant increase in the probability or the

consequences of an accident previously evaluated.

(2) The proposed changes only apply to the testing requirements

for ATMs in the systems identified above and do not result in any

physical change to these or other components or their operation. As

a result, no new failure modes are introduced. Therefore, the

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

kind of accident from any accident previously evaluated.

(3) The current TS-required response times are based on the

maximum values assumed in the plant safety analyses. These analyses

conservatively establish the margin of safety. As described above,

the proposed changes do not affect the capability of the associated

systems to perform their intended function within the allowed

response time used as the basis for the plant safety analyses. The

potential failure modes for the components within the scope of this

request were evaluated for impact on instrument response time. This

evaluation confirmed that the remaining TS-required testing is

sufficient to identify failure modes or degradations in instrument

response times and to ensure that operation of the instrumentation

within the scope of this request is within acceptable limits. As a

result, it has been concluded that plant and system response to an

initiating event will remain in compliance with the assumptions of

the safety analysis.

Further, although not explicitly evaluated, the proposed changes

will provide an improvement to plant safety and operation by

reducing the time safety systems are unavailable, reducing the

potential for safety system actuations, reducing plant shutdown

risk, limiting radiation exposure to plant personnel, and

eliminating the diversion of key personnel resources to conduct

unnecessary testing. Therefore, IP has concluded that this request

will result in an overall increase 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: Vespasian Warner Public

Library, 120 West Johnson Street, Clinton, Illinois 61727

Attorney for licensee: Sheldon Zabel, Esq., Schiff, Hardin and

Waite, 7200 Sears Tower, 233 Wacker Drive, Chicago, Illinois 60606

NRC Project Director: Gail H. Marcus

Northern States Power Company, Docket No. 50-263, Monticello

Nuclear Generating Plant, Wright County, Minnesota

Date of amendment request: December 11, 1995

Description of amendment request: The proposed amendment would

modify Technical Specification (TS) Section 4.7, Surveillance

Requirements for Primary Containment Automatic Isolation Valves.

Specifically, the proposed amendment would delete TS Surveillance

Requirement 4.7.D.4, which requires replacement of the seat seals for

the drywell and suppression chamber purge and vent valves every 5

years.

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 amendment will not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

An evaluation of the operational performance of the 18-inch

purge and vent valves has concluded that deletion of the Monticello

Technical Specification surveillance requirement 4.7.D.4 will have

no adverse impact on the seat leakage performance of these primary

containment isolation valves, no adverse impact on the testing

performed in accordance with 10 CFR 50, Appendix J, and thus no

adverse impact on the containment isolation function of these

primary containment isolation valves. The material of which the T-

shaped elastomer seat is comprised of has been found to withstand

normal and accident thermal exposures for the design life of the

plant based on a thermal aging analysis. Radiation effects will not

have an adverse impact on the elastomer seat material. Therefore,

this amendment will not cause a significant increase in the

probability or consequences of an accident previously evaluated for

the Monticello plant.

The proposed amendment will not create the possibility of a new

or different kind of accident from any accident previously analyzed.

The proposed change to the Technical Specifications for the

Primary Containment Purge and Vent valves does not alter the

function of these components or their interrelationships with other

systems. Therefore, this amendment will not create the possibility

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

analyzed.

The proposed amendment will not involve a significant reduction

in the margin of safety.

The operating experience of these valves has demonstrated that

the testing performed

[[Page 3502]]

in accordance with 10 CFR 50, Appendix J, provides a high level of

confidence in the ability of these valves to perform their safety

function with respect to valve leak tightness. The proposed

amendment will 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: Minneapolis Public Library,

Technology and Science Department, 300 Nicollet Mall, Minneapolis,

Minnesota 55401

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

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

NRC Project Director: John N. Hannon

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: December 27, 1995

Description of amendment requests: The amendments would revise the

combined Technical Specifications (TS) 3/4.6.1.1, ``Containment

Integrity;'' 3/4.6.1.2, ``Containment Leakage;'' 3/4.6.1.3,

``Containment Air Locks;'' 3/4.6.1.6, ``Containment Structural

Integrity;'' 3/4.6.3, ``Containment Isolation Valves;'' and their

associated Bases; and would add TS 6.8.4.j, ``Containment Leakage Rate

Testing Program,'' to implement the performance-based leakage rate

testing program, as permitted by 10 CFR Part 50, Appendix J. These

changes will support the implementation of the performance-based

testing of Option B to Appendix J for Types A, B, and C containment

leakage rate testing and the appropriate rescheduling of testing. The

amendment changes the TS 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:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The proposed changes to Technical Specification (TS) 3/4.6.1.1,

3/4.6.1.2, 3/4.6.1.3, 3/4.6.1.6, 3/4.6.3, and the addition of 6.8.4

j., to implement the performance-based Containment Leakage Rate

Testing Program have no effect on plant operation. The proposed

changes only provide mechanisms within the TS for implementing a

performance-based methodology for determining the frequency of leak

rate testing that has been approved by the Commission. The test type

and test method used for testing would not be changed. The test

acceptance criteria would not be changed, and containment leakage

will continue to be maintained within the required limits.

Directly referencing the Containment Leakage Rate Testing

Program for containment ILRT [integrated leak rate testing] and LLRT

[local leak rate test] requirements does not involve any

modification to plant equipment or affect the operation or design

basis of the containment. Leakage rate testing is not a precursor to

or an initiating event for any accident.

Therefore, the proposed changes do not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

The proposed changes only allow for the implementation of 10 CFR

50, Appendix J, Option B, testing frequencies and do not involve any

modifications to any plant equipment or affect the operation or

design basis of the containment. The proposed changes do not affect

the response of the containment during a design basis accident.

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 proposed changes do not affect or change a Safety Limit or

affect plant operations. The changes only implement the allowed 10

CFR 50, Appendix J, Option B testing frequencies that have been

determined by the Commission not to involve a safety concern. The

testing method, acceptance criteria, and basis for testing are not

changed and still provide assurance that the containment will

provide its intended function.

Therefore, the proposed changes do not involve a significant

reduction in a margin of safety.

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

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

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

amendment requests involve no significant hazards consideration.

Local Public Document Room location: California Polytechnic State

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

Department, San Luis Obispo, California 93407

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

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

NRC Project Director: William H. Bateman

Saxton Nuclear Experimental Corporation (SNEC), Docket No. 50-146,

Saxton Nuclear Experimental Facility (SNEF), Bedford County,

Pennsylvania

Date of amendment request: November 21, 1995.

Description of amendment request: The proposed amendment would

change the license and technical specifications to add GPU Nuclear

Corporation (GPUN) as a licensee for the SNEF along with SNEC and would

transfer from SNEC to GPUN all management-related responsibilities for

the SNEF. Responsibility for safely maintaining the containment vessel

and performing characterization activities would change from SNEC to

GPUN. Technical specification organizational positions would be changed

from SNEC titles to GPUN titles. GPUN would take responsibility from

SNEC for administration of all SNEF functions, for radiation safety

activities, and for providing on-site management and continuing

oversight of production activities. The appointment of members to the

Saxton Radiation Safety Committee and the reporting of the Committee

would change from the SNEC President to the GPUN Vice President of the

Nuclear Services Division. The GPUN President would have the authority

to request audits and would receive audit reports instead of the SNEC

President. Procedure control methodology and the administrative

procedure for procedures would be changed from SNEC procedures to GPUN

procedures. The responsibility for records retention and reporting

would change from SNEC to GPUN. The organization chart for the facility

would be changed to reflect the addition of GPUN as a licensee.

Basis for proposed no significant Hazards Consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

The proposed changes do not involve a significant hazards

considerations because the changes would not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

Because the proposed changes are administrative in nature they

would have no effect on the likelihood or impact on the potential

accidents of fire, flood or radiological hazard.

[[Page 3503]]

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

from any previously analyzed.

Because the proposed changes are administrative in nature they

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

accident from any accident previously analyzed.

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

Because the proposed changes are administrative in nature they

would not involve any reduction in a margin of safety.

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: Saxton Community Library, 911

Church Street, Saxton, Pennsylvania 16678 Attorney for the Licensee:

Ernest L. Blake, Jr., Esquire, Shaw, Pittman, Potts, and Trowbridge,

2300 N Street, NW., Washington, DC 20037

NRC Project Director: Seymour H. Weiss

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah

Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

Date of amendment request: January 4, 1996 (TS 95-22)

Description of amendment request: The proposed change would extend

the functional testing interval for the following isolation radiation

monitor instruments from monthly to quarterly: (1) Engineered Safety

Feature Actuation System Instrumentation Surveillance Requirements

Table 4.3-2, Item 3.c.3, Containment Purge Air Exhaust Monitor

Radioactivity-High; (2) Radiation Monitoring Instrumentation

Surveillance Requirements Table 4.3-3, Item 1.a, Fuel Storage Pool Area

Radiation Monitor; (3) Table 4.3-3, Item 2.a, Containment Purge Air

Exhaust; (4) Table 4.3-3, Item 2.b.i, Containment Gaseous Activity RCS

Leakage Detection; (5) Table 4.3-3, Item 2.b.ii, Containment

Particulate Activity RCS Leakage Detection; and (6) Table 4.3-3, Item

2.c, Control Room Isolation.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

TVA has evaluated the proposed technical specification (TS)

change and has determined that it does not represent a significant

hazards consideration based on criteria established in 10 CFR

50.92(c). Operation of Sequoyah Nuclear Plant (SQN) in accordance

with the proposed amendment will not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

Review of the past history for the affected and similar

radiation monitors revealed that extending the functional testing

interval for these monitors will not adversely affect system

operability and will effectively increase system availability. These

radiation monitors are not accident initiating equipment, thus

increasing the surveillance interval on these monitors will not

affect the probability of any accident previously evaluated. Based

on the above statements, it is concluded that the probability or

consequences of an accident previously evaluated is not increased.

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

from any previously analyzed.

No new type of accident or malfunction will be created since the

radiation monitors are not accident initiating equipment. The

proposed change merely increases the functional testing interval for

the affected radiation monitors, and does not change the method and

manner of plant operation. The safety design bases in the Updated

Final Safety Analysis Report have not been altered.

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

or different kind of accident from any previously analyzed.

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

The proposed changes do not change the plant configuration in a

way that introduces a new potential hazard to the plant and do not

involve a significant reduction in the margin of safety. The

proposed changes do not affect applicable safety analysis acceptance

criteria and will not affect system operating conditions.

Additionally, plant operating experience with similar monitors has

shown that there has not been additional failures due to the

quarterly testing frequency. Thus, it is concluded that the margin

of safety is not reduced.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Chattanooga-Hamilton County

Library, 1101 Broad Street, Chattanooga, Tennessee 37402

Attorney for licensee: General Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902

NRC Project Director: Frederick J. Hebdon

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

Creek Generating Station, Coffey County, Kansas

Date of amendment request: November 22, 1995

Description of amendment request: The proposed amendment replaces

the requirements associated with the boron dilution mitigation system

(BDMS) in the Wolf Creek Generating Station Technical Specifications

with alarms, indicators, procedures, and controls to assure proper

resolution of potential inadvertent boron dilution events.

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 change does not involve a significant increase in

the probability or consequences of an accident previously evaluated.

The only event potentially impacted by the proposed change is

the inadvertent boron dilution event. The discussion of the

probability and consequences of an inadvertent boron dilution event

at WCGS is provided in USAR [Updated Safety Analysis Report] Section

15.4.6. Primarily, the proposed changes revise the method of

detecting and mitigating the event. The only aspect of the changes

that impact[s] the potential causes of an inadvertent boron dilution

event is the increased requirement to isolate potential dilution

sources in Modes 3, 4 and 5. As a result, the overall probability of

the event is slightly decreased.

The alternate methods to detect and mitigate this event achieve

the same basic goal as the current BDMS; to prevent a return to

critical during an inadvertent dilution event. The proposed changes

to the BDMS will result in an improved system that will provide an

improved response to the inadvertent boron dilution event, and that

will prevent a return to critical. Thus, it can be concluded that

the proposed change will not significantly increase the consequences

of a postulated inadvertent boron dilution event.

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

or different kind of accident from any accident previously

evaluated.

The revisions to plant procedural requirements to either operate

a reactor coolant pump or to isolate/control potential dilution

sources does [sic] not create the potential for a new or different

kind of accident because these new requirements are configurations

which have always been allowed. Similarly, the new normal position

for the letdown divert valve does not create a new or different

accident because the new normal position has always been an allowed

position. The other procedural changes only increase the plant

operators' awareness of potential boron dilution problems or provide

the steps needed to respond to available indications and alarms to

mitigate the potential event. As a result, these procedural changes

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

accident.

The proposed changes also include addition of new redundant VCT

high level alarms and a new alarm indicating that the

[[Page 3504]]

letdown divert valve is not in the ``VCT'' position. Because the alarms

are passive, they do not create 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 design criterion and margin of safety for the current BDMS

is that the dilution event is terminated prior to the loss of all

shutdown margin. The same criterion will be met following the

implementation of the proposed changes. Therefore, there is no

reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room 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

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

Creek Generating Station, Coffey County, Kansas

Date of amendment request: December 20, 1995

Description of amendment request: This amendment request proposes

to revise Technical Specification 3/4.6.1.1, ``Containment Integrity,''

and 3/4.6.1.3, ``Containment Air Locks,'' and to add Technical

Specification 6.8.4i, ``Containment Leakage Rate Testing Program,'' to

implement the new performance-based leakage rate testing program as

permitted by 10 CFR 50, Appendix J. Also, Technical Specification 1.7e,

``Containment Integrity,'' would be revised to reference Technical

Specification 4.6.1.1.c. These proposed changes will implement the

performance-based testing of Option B to Appendix J, for Type A, B, and

C containment leak testing 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:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The proposed changes to Technical Specifications 3/4.6.1.1 and

3/4.6.1.3, and the addition of Technical Specification 6.8.4i to

implement the new performance based Containment Leakage Rate Testing

Program, have no effect on plant operation. The proposed changes

only provide mechanisms within the technical specifications for

implementing a performance-based methodology, for determining the

frequency of leak rate testing, which has been approved by the NRC.

The test type and test method used for testing would not be changed.

The test acceptance criteria would not be changed, and containment

leakage will continue to be maintained within the required limits.

Directly referencing the Containment Leakage Rate Testing

Program for containment integrated leak rate test and local leak

rate test requirements does not involve any modification to plant

equipment or affect the operation or design basis of the

containment. Leakage rate testing is not a precursor to or an

initiating event for any accident.

Therefore, the proposed changes do not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

The proposed changes only allow for implementation of 10 CFR 50,

Appendix J, Option B, testing frequencies and do not involve any

modifications to any plant equipment or affect the operation or

design basis of the containment. The proposed changes do not affect

the response of the containment during a design basis accident.

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 proposed changes do not affect or change a Safety Limit, any

limiting condition for operation or affect plant operations. The

changes only implement the allowed Option B testing frequencies that

have been determined by the NRC not to involve a safety concern. The

testing method, acceptance criteria, and bases are not changed and

still provide assurance that the containment will provide its

intended function.

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 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.

[[Page 3505]]

Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden

Nuclear Power Station, Units 2 and 3, Grundy County, Illinois

Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station,

Units 1 and 2, Rock Island County, Illinois

Date of application for amendments: November 14, 1995, as

supplemented January 4, 1996.

Brief description of amendments: The amendments revise the

Technical Specifications to incorporate 10 CFR Part 50, Appendix J,

``Primary Reactor Containment Leakage Testing for Water-Cooled Power

Reactors,'' Option B. Technical Specification changes for the LaSalle

facility will be addressed under separate correspondence.

Date of issuance: January 11, 1996

Effective date: January 11, 1996

Amendment Nos.: 148, 142, 169, and 165

Facility Operating License Nos. DPR-19, DPR-25, DPR-29 and DPR-30.

The amendments revised the Technical Specifications.

Date of initial notice in Federal Register: December 7, 1995 (60 FR

62896). The January 4, 1996, supplement provided a specific

implementation date for the requested amenement. This information was

within the scope of the original application and did not change the

staff's initial proposed no significant hazards consideration

determination. The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated January 11, 1996. No significant

hazards consideration comments received: No

Local Public Document Room location: for Dresden, Morris Area

Public Library District, 604 Liberty Street, Morris, Illinois 60450;

for Quad Cities, Dixon Public Library, 221 Hennepin Avenue, Dixon,

Illinois 61021.

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

Neck Plant, Middlesex County, Connecticut

Date of application for amendment: November 14, 1995

Brief description of amendment: The amendment revises the Haddam

Neck Technical Specifications (TS) to provide an one-time exception to

TS 3.9.12, '' Fuel Building Storage Air Cleanup System,'' to allow the

fuel storage building air cleanup system to be inoperable for a limited

duration during intervals in which new fuel rack modules will be moved

into and old fuel rack modules will be moved out of the fuel storage

building.

Date of Issuance: January 17, 1996

Effective date: As of the date of issuance, to be implemented

within 30 days.

Amendment No.: 187

Facility Operating License No. DPR-61. Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: November 28, 1995 (60

FR 58688) The Commission's related evaluation of this amendment is

contained in a Safety Evaluation dated January 17, 1996 No significant

hazards consideration comments received: No.

Local Public Document Room location: Russell Library, 123 Broad

Street, Middletown, CT 06457.

Consumers Power Company, Docket No. 50-155, Big Rock Point Plant,

Charlevoix County, Michigan Date of application for amendment:

November 8, 1995, as supplemented November 17, 1995

Brief description of amendment: The amendment removes the

prescriptive Type A containment leakage test rate frequency of 40 plus

or minus 10 months and adds a reference to perform containment leakage

rate tests in accordance with the criteria specified in Appendix J of

10 CFR Part 50 as modified by approved exemptions. In addition, the

amendment revises the test pressure for Type B and C testing to correct

a typographical error.

Date of issuance: January 16, 1996

Effective date: January 16, 1996

Amendment No.: 117

Facility Operating License No. DPR-6. Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: December 6, 1995 (60 FR

62489) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated January 16, 1996. No significant

hazards consideration comments received: No.

Local Public Document Room location: North Central Michigan

College, 1515 Howard Street, Petoskey, Michigan 49770.

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

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

Date of application for amendments: March 29, 1995, as supplemented

by letters dated September 18 and November 16, 1995

Brief description of amendments: The amendments revise Technical

Specification requirements for the Low Temperature Overpressure

Protection system and update the heatup and cooldown curves for both

units.

Date of issuance: January 11, 1996

Effective date: As of the date of issuance to be implemented within

60 days

Amendment Nos.: Unit 1 - 162; Unit 2 - 144

Facility Operating License Nos. NPF-9 and NPF-17: Amendments

revised the Technical Specifications. Date of initial notice in Federal

Register: September 27, 1995 (60 FR 49933) The September 18 and

November 16, 1995, letters provided clarifying information that did not

change the scope of the March 29, 1995, application and the initial

proposed no significant hazards consideration determination. The

Commission's related evaluation of the amendments is contained in a

Safety Evaluation dated January 11, 1996. No significant hazards

consideration comments received: No.

Local Public Document Room location: Atkins Library, University of

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

Georgia Power Company, Oglethorpe Power Corporation, Municipal

Electric Authority of Georgia, City of Dalton, Georgia, Docket Nos.

50-321 and 50-366, Edwin I. Hatch Nuclear Plant, Units 1 and 2,

Appling County, Georgia

Date of application for amendments: December 2, 1994

Brief description of amendments: The amendments replace Appendix B,

``Environmental Technical Specifications,'' with an Environmental

Protection Plan (Nonradiological) and revise the Operating Licenses to

reflect these changes.

Date of issuance: December 19, 1995

Effective date: As of the date of issuance to be implemented within

30 days

Amendment Nos.: Unit 1 - 199 - Unit 2 - 140

Facility Operating License Nos. DPR-57 and NPF-5. Amendments

revised the Technical Specifications and Operating Licenses.

Date of initial notice in Federal Register: January 4, 1995 (60 FR

502) The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated December 19, 1995. No significant hazards

consideration comments received: No

Local Public Document Room location: Appling County Public Library,

301 City Hall Drive, Baxley, Georgia 31513

[[Page 3506]]

Georgia Power Company, Oglethorpe Power Corporation, Municipal

Electric Authority of Georgia, City of Dalton, Georgia, Docket Nos.

50-424 and 50-425, Vogtle Electric Generating Plant, Units 1 and 2,

Burke County, Georgia

Date of application for amendments: March 17, 1995, as supplemented

by letter dated July 6, 1995

Brief description of amendments: The amendments revise Technical

Specification 3/4.9.4, Containment Building Penetrations, to allow the

personnel airlock to be open during core alterations or movement of

irradiated fuel within the containment.

Date of issuance: November 30, 1995

Effective date: As of the date of issuance to be implemented within

30 days

Amendment Nos.: 92 and 70

Facility Operating License Nos. NPF-68 and NPF-81: Amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: July 5, 1995 (60 FR

35077) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated November 30, 1995. No

significant hazards consideration comments received: No

Local Public Document Room location: Burke County Library, 412

Fourth Street, Waynesboro, Georgia 30830

Gulf States Utilities Company, Cajun Electric Power Cooperative,

and Entergy Operations, Inc., Docket No. 50-458, River Bend

Station, Unit 1, West Feliciana Parish, Louisiana

Date of amendment request: August 17, 1995, as supplemented by

letters dated November 22, and December 18, 20, and 27, 1995

Brief description of amendment: The amendment revised the primary

containment air lock technical specifications to allow the air locks to

be open in Mode 5 (refueling) during core alterations except for

movement of recently irradiated fuel. All other provisions of the

August 17, 1995, requests are defered.

Date of issuance: January 11, 1996

Effective date: January 11, 1996

Amendment No.: 85

Facility Operating License No. NPF-47. The amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: September 13, 1995 (60

FR 47619) The additional information contained in the supplemental

letters dated November 22, and December 18, 20, and 27, 1995, was

clarifying in nature and thus, within the scope of the initial notice

and did not affect the staff's proposed no significant hazards

consideration determination. The Commission's related evaluation of the

amendment is contained in a Safety Evaluation dated Janaury 11, 1996.

No significant hazards consideration comments received. No.

Local Public Document Room location: Government Documents

Department, Louisiana State University, Baton Rouge, LA 70803

Gulf States Utilities Company, Cajun Electric Power Cooperative,

and Entergy Operations, Inc., Docket No. 50-458, River Bend

Station, Unit 1, West Feliciana Parish, Louisiana

Date of amendment request: November 20, 1995

Brief description of amendment: The proposed amendment revised the

technical specifications to eliminate the response time testing

requirements for selected Reactor Protection System Instrumentation.

Date of issuance: January 11, 1996

Effective date: January 11, 1996

Amendment No.: 86

Facility Operating License No. NPF-47. The amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: December 6, 1995 (60 FR

62492) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated Janaury 11, 1996. No significant

hazards consideration comments received. No.

Local Public Document Room location: Government Documents

Department, Louisiana State University, Baton Rouge, LA 70803

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316,

Donald C. Cook Nuclear Plant, Unit Nos. 1 and 2, Berrien County,

Michigan

Date of application for amendments: May 25, 1995 (AEP:NRC:1071T)

Brief description of amendments: The amendments incorporate a

cycle- and burnup-dependent peaking factor penalty in the Core

Operating Limits Report and add an appropriate reference to the COLR

and update the topical report reference in the Technical

Specifications.

Date of issuance: January 4, 1996

Effective date: January 4, 1996, with full implementation within 45

days

Amendment Nos.: Unit 1, 206, Unit 2, 190

Facility Operating License Nos. DPR-58 and DPR-74. Amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: The Commission's

related evaluation of the amendments is contained in a Safety

Evaluation dated January 4, 1996. No significant hazards consideration

comments received: No.

Local Public Document Room location: Maud Preston Palenske Memorial

Library, 500 Market Street, St. Joseph, Michigan 49085.

Northeast Nuclear Energy Company, Docket No. 50-245, Millstone

Nuclear Power Station, Unit 1, New London County, Connecticut

Date of application for amendment: October 3, 1995

Brief description of amendment: The amendment removes the Limiting

Condition for Operation (LCO) and Surveillance Requirements for the

loss-of-normal power (LNP) trip function from Tables 3.2.2 and 4.2.1

and inserts new LCO 3.2.F and Surveillance Requirement 4.2.F. In

addition, the amendment adds a new table to specify the required LNP

instrumentation for each bus, updates the Table of Contents, makes some

editorial changes, and revises the associated Bases section.

Date of issuance: January 17, 1996

Effective date: As of the date of issuance, to be implemented

within 30 days.

Amendment No.: 92

Facility Operating License No. DPR-21. Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: December 4, 1995 (60 FR

62111) The Commission's related evaluation of the amendment is

contained in a Safety evaluation dated January 17, 1996. No significant

hazards consideration comments received: No.

Local Public Document Room location: Learning Resources Center,

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

Norwich, CT 06360.

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

Millstone Nuclear Power Station, Unit No. 2, New London County,

Connecticut

Date of application for amendment: September 11, 1995, as

supplemented November 15, 1995.

Brief description of amendment: The amendment changes Technical

Specification Sections 3.4.8 and 3.9.9, Tables 2.2-1, 3.3-3, 3.3-5 and

3.3-8, and Bases Sections 3/4.2.1, 3/4.4.8 and 3/4.11.2.1. These

changes combine several different administrative changes which will

correct typographical errors, provide clarifications, or make editorial

changes.

Date of issuance: January 17, 1996

Effective date: As of the date of issuance, to be implemented

within 60 days.

Amendment No.: 194

[[Page 3507]]

Facility Operating License No. DPR-65. Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: October 11, 1995 (60 FR

52933) The November 15, 1995, letter provided clarifying information

that did not change the initial proposed no significant hazards

consideration determination. The Commission's related evaluation of the

amendment is contained in a Safety Evaluation dated January 17, 1996.

No significant hazards consideration comments received: No.

Local Public Document Room location: Learning Resources Center,

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

Norwich, CT 06360.

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

Millstone Nuclear Power Station, Unit No. 2, New London County,

Connecticut

Date of application for amendment: September 29, 1995, as

supplemented November 9, 1995.

Brief description of amendment: The amendment provides three

changes to the Technical Specifications (TS) relating to the

pressurizer safety valves (PSV) and the main steam safety valves

(MSSV).

The first change is to TS 3.4.2.1 and 3.4.2.2 and involves relaxing

the as-found setpoint tolerance for the pressurizer safety valves

(PSVs) and the main steam safety valves (MSSVs) from the current value

of plus or minus 1% to plus or minus 3%. Table 4.7-1 is also modified

to correct the as-found tolerance for the MSSV from plus or minus 1% to

plus or minus 3%. Notes are added to TS 3.4.2.2 and Table 4.7-1 which

specify that the lift setting should be determined at nominal operating

conditions and should be set at plus or minus 1% of the lift setting.

For the second change, Surveillance Requirement 4.7.1.1 and Table

4.7-1 are modified to eliminate the need to verify the orifice size of

each MSSV.

The third change modifies the statement for TS 3.7.1.1 so that if a

MSSV is inoperable and compensating action cannot be taken, the plant

must be brought to hot shutdown (Mode 4) within 12 hours instead of

cold shutdown (Mode 5) in 30 hours.

Date of issuance: January 18, 1996

Effective date: As of the date of issuance, to be implemented

within 60 days.

Amendment No.: 195

Facility Operating License No. DPR-65. Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: October 25, 1995 (60 FR

54723) The November 9, 1995, letter provided clarifying information

that did not change the initial proposed no significant hazards

consideration determination. The Commission's related evaluation of the

amendment is contained in a Safety Evaluation dated January 18, 1996.

No significant hazards consideration comments received: No.

Local Public Document Room location: Learning Resources Center,

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

Norwich, CT 06360.

PECO Energy Company, Public Service Electric and Gas Company

Delmarva Power and Light Company, and Atlantic City Electric

Company, Docket Nos. 50-277 and 50-278, Peach Bottom Atomic Power

Station, Unit Nos. 2 and 3, York County, Pennsylvania

Date of application for amendments: November 30, 1995

Brief description of amendments: The amendments change the

technical specification requirements for control rod drive scram

accumulator and charging water header minimum pressure.

Date of issuance: January 11, 1996

Effective date: Unit 2, as of date of issuance, to be implemented

concurrently with Amendment 210, issued August 30, 1995; Unit 3, as of

date of issuance, to be implemented concurrently with Amendment 214,

issued August 30, 1995.

Amendments Nos.: 211 and 216

Facility Operating License Nos. DPR-44 and DPR-56: The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: December 8, 1995 (60 FR

63073) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated January 11, 1996 No significant

hazards consideration comments received: No

Local Public Document Room location: Government Publications

Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education

Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg,

Pennsylvania 17105.

PECO Energy Company, Public Service Electric and Gas Company

Delmarva Power and Light Company, and Atlantic City Electric

Company, Docket Nos. 50-277 and 50-278, Peach Bottom Atomic Power

Station, Unit Nos. 2 and 3, York County, Pennsylvania

Date of application for amendments: November 21, 1995

Brief description of amendments: The amendments change the test

pressure requirements for the high pressure coolant injection system

and the reactor core isolation cooling system surveillance tests. The

amendments also change Section 5.5.7 of the technical specifications to

eliminate reference to a section which was previously eliminated.

Date of issuance: January 11, 1996

Effective date: As of date of issuance.

Amendments Nos.: 212 and 217

Facility Operating License Nos. DPR-44 and DPR-56: The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: December 5, 1995 (60 FR

62271) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated January 11, 1996 No significant

hazards consideration comments received: No

Local Public Document Room location: Government Publications

Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education

Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg,

Pennsylvania 17105.

PECO Energy Company, Public Service Electric and Gas Company

Delmarva Power and Light Company, and Atlantic City Electric

Company, Docket Nos. 50-277 and 50-278, Peach Bottom Atomic Power

Station, Unit Nos. 2 and 3, York County, Pennsylvania

Date of application for amendments: December 19, 1995

Brief description of amendments: These amendments change the

ventilation filter test program bypass and penetration leakage test

acceptance criteria from less than 0.05 percent to less than 1.0

percent. The change corrects an administrative error that occurred

during the development of the Peach Bottom Improved Technical

Specifications which were issued as Amendments 210 and 214 to the Peach

Bottom licenses on August 30, 1995.

Date of issuance: January 16, 1996

Effective date: Unit 2, effective as of date of issuance, to be

implemented concurrently with Amendment 210, issued August 30, 1995;

Unit 3, effective as of date of issuance, to be implemented

concurrently with Amendment 214, issued August 30, 1995.

Amendments Nos.: 213 and 218

Facility Operating License Nos. DPR-44 and DPR-56: The amendments

revised the Technical Specifications. Public comments requested as to

proposed no significant hazards consideration: Yes (60 FR 66997,

December 27, 1995). That notice provided an opportunity to submit

comments on the Commission's proposed no significant hazards

[[Page 3508]]

consideration determination. No comments have been received. The notice

also provided for an opportunity to request a hearing by January 26,

1996, but indicated that if the Commission makes a final no significant

hazards consideration determination any such hearing would take place

after issuance of the amendment. The Commission's related evaluation of

the amendments, finding of exigent circumstances, and final

determination of no significant hazards consideration are contained in

a Safety Evaluation dated January 16, 1996

Local Public Document Room location: Government Publications

Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education

Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg,

Pennsylvania 17105.

Dated at Rockville, Maryland, this 23rd day of January 1996.

For the Nuclear Regulatory Commission

Steven A. Varga,

Director, Division of Reactor Projects - I/II, Office of Nuclear

Reactor Regulation.

[Doc. 96-1683 Filed 1-30-96; 8:45 am]

BILLING CODE 7590-01-F