Biweekly Notice

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 March 4, 1996, through March 15, 1996. The

last biweekly notice was published on March 13, 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 April 26, 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

[[Page 13522]]

Board will issue a notice of a hearing or an appropriate order.

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

shall set forth with particularity the interest of the petitioner in

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

the proceeding. The petition should specifically explain the reasons

why intervention should be permitted with particular reference to the

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

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

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

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

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

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

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

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

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

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

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

the specificity requirements described above.

Not later than 15 days prior to the first prehearing conference

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

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

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

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

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

brief explanation of the bases of the contention and a concise

statement of the alleged facts or expert opinion which support the

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

contention at the hearing. The petitioner must also provide references

to those specific sources and documents of which the petitioner is

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

facts or expert opinion. Petitioner must provide sufficient information

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

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

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

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

petitioner who fails to file such a supplement which satisfies these

requirements with respect to at least one contention will not be

permitted to participate as a party.

Those permitted to intervene become parties to the proceeding,

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

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

hearing, including the opportunity to present evidence and cross-

examine witnesses.

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

determination on the issue of no significant hazards consideration. The

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

If the final determination is that the amendment request involves

no significant hazards consideration, the Commission may issue the

amendment and make it immediately effective, notwithstanding the

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

of the amendment.

If the final determination is that the amendment request involves a

significant hazards consideration, any hearing held would take place

before the issuance of any amendment.

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

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

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

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

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

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

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

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

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

should be given Datagram Identification Number N1023 and the following

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

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

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

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

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

attorney for the licensee.

Nontimely filings of petitions for leave to intervene, amended

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

not be entertained absent a determination by the Commission, the

presiding officer or the Atomic Safety and Licensing Board that the

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

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

For further details with respect to this action, see the

application for amendment which is available for public inspection at

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

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

the particular facility involved.

Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN

50-529, and STN 50-530, Palo Verde Nuclear Generating Station,

Units Nos. 1, 2, and 3, Maricopa County, Arizona

Date of amendments request: February 1, 1996

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

revise Technical Specifications (TS) Sections 3/4.1.1.1, 6.9.1.9, and

6.9.1.10 to relocate the shutdown margin (reactor trip breakers open)

to the Core Operating Limits Report (COLR); (2) revise TS 3/4.3.2

(Tables 3.3-3 and 3.3-4), to specify an additional restriction for the

allowed low pressurizer pressure trip setpoint when reducing reactor

coolant system (RCS) pressure in Mode 3; (3) revise TS Section 2.2.1

(Table 2.2-1) to make it consistent with the footnote in TS Tables 3.3-

3 and 3.3-4; and (4) revise TS Sections 3/4.5.2 and 3/4.5.3 to specify

an additional restriction to require that two emergency core cooling

system (ECCS) subsystems be operable in Mode 3 whenever the RCS cold

leg temperature is equal to or above 485 degrees F. In addition, the

Table of Contents and the Bases would be revised to be consistent with

these changes.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The proposed changes do not significantly increase the

probability or consequences of an accident previously evaluated in

the Updated Final Safety Analysis Report (UFSAR). The proposed

changes to TS Tables 2.2-1, 3.3-3, and 3.3-4 to add additional

restrictions to the pressurizer pressure - low trip setpoint

requirements are more conservative than the current Technical

Specifications and will reflect the updated Mode 3 steam line break

safety analyses assumptions. The proposed changes to TS sections 3/

4.5.2 and 3/4.5.3 to add additional restrictions to the requirement

to have two ECCS Subsystems operable are also more conservative than

the current Technical Specifications and will reflect the updated

Mode 3 steam line break safety analyses assumptions. Since these

changes are more restrictive, they would not contribute to the

initiation of any accident, nor would they increase the consequences

of an accident, but

[[Page 13523]]

they would enhance the plant response to a steam line break in Mode

3 to reduce consequences. The proposed changes to relocate the

shutdown margin - reactor trip breakers open to the COLR will have

no effect on the initiation or consequences of an accident. The

shutdown margin-reactor trip breakers open, which would be

determined using NRC approved analytical methods, as required by the

proposed changes, would ensure that the probability and consequences

of an accident would not increase. The changes to the titles of TS

3/4.5.2 and 3/4.5.3, and to the Table of Contents, are editorial and

have no effect on the operation of the plant or on any structures,

systems or components.

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

or different kind of accident from any accident previously

evaluated.

The proposed TS changes do not create the possibility of an

accident of a new or different kind. The proposed changes to TS

Tables 2.2-1, 3.3-3, and 3.3-4, and TS section 3/4.5.2 and 3/4.5.3,

to add additional restrictions to the pressurizer pressure - low

trip setpoint requirement and add additional restrictions to the

requirement to have two ECCS Subsystems operable are more

conservative than the current Technical Specifications and will

reflect the updated Mode 3 steam line break safety analyses

assumptions. Since these changes are more restrictive, and therefore

bounded by the current TS, they would not contribute to the

initiation of any kind of new or different accident. The proposed

changes to relocate the shutdown margin -reactor trip breakers open

to the COLR will have no effect on the possibility of a new or

different kind of accident. The shutdown margin-reactor trip

breakers open, which would be determined using NRC approved

analytical methods as required by the proposed changes, would ensure

that there would be no possibility of a new or different kind of

accident from any accident previously evaluated. The changes to the

titles of TS 3/4.5.2 and 3/4.5.3, and to the Table of Contents, are

editorial and have no effect on the operation of the plant or on any

structures, systems or components.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The proposed TS changes do not involve a reduction in any margin

of safety. The proposed changes to TS Tables 2.2-1, 3.3-3, and 3.3-

4, and TS section 3/4.5.2 and 3/4.5.3, to add additional

restrictions to the pressurizer pressure - low trip setpoint

requirement and add additional restrictions to the requirement to

have two ECCS Subsystems operable are more conservative than the

current Technical Specifications and will reflect the updated Mode 3

steam line break safety analyses assumptions. Since these changes

are more restrictive, they do not involve a reduction in any margin

of safety as currently established by the existing TS. The proposed

changes to relocate the shutdown margin - reactor trip breakers open

to the COLR will have no effect on any margin of safety. The

shutdown margin - reactor trip breakers open would be determined

using NRC approved analytical methods as required by the proposed

changes, thus ensuring that there would be no reduction in any

margin of safety. The changes to the titles of TS 3/4.5.2 and 3/

4.5.3, and to the Table of Contents, are editorial and have no

effect on the operation of the plant or on any structures, systems

or components.

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

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

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

amendments request involve no significant hazards consideration.

Local Public Document Room location: Phoenix Public Library, 1221

N. Central Avenue, Phoenix, Arizona 85004

Attorney for licensee: Nancy C. Loftin, Esq., Corporate Secretary

and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail

Station 9068, Phoenix, Arizona 85072-3999

NRC Project Director: William H. Bateman

Duke Power Company, Docket Nos. 50-269, 50-270 and 50-287, Oconee

Nuclear Station, Units 1, 2 and 3, Oconee County, South Carolina

Date of amendment request: February 15, 1996

Description of amendment request: The proposed amendments would

revise Technical Specification (TS) 3.7 to add operability requirements

for the Keowee Hydro units during periods of commercial power

generation. These requirements are based on lake level and power level

of the Keowee Hydro units. Also, two surveillance requirements would be

added to TS 4.6 to (1) address periodic testing of the circuitry that

was added by the modification approved in NRC's SER dated August 15,

1995, and (2) add a load rejection surveillance to ensure that the

response of the Keowee Hydro units is bounded by the design criteria

used to develop the Keowee operating restrictions.

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 not] involve a significant increase in the probability

or consequences of an accident previously evaluated:

Each accident analysis addressed within the Oconee Final Safety

Analysis Report (FSAR) has been examined with respect to the change

proposed within this amendment request. The probability of any

Design Basis Accident (DBA) is not significantly increased by this

change. In addition, the consequences of the accidents are within

the bounds of the FSAR analyses.

The design basis of the auxiliary electrical systems is to

supply the required engineered safeguards (ES) loads of one unit and

the safe shutdown loads of the other two units. The systems are

arranged so that no single failure will jeopardize plant safety. The

addition of the operability requirement and surveillances for the

Keowee Hydro units will ensure that the electrical systems can meet

their design basis.

(2) [Does not] create the possibility of a new or different kind

of accident from any kind of accident previously evaluated:

Addition of the operability requirement and surveillances will

not create a new or different kind of accident. The addition of the

circuitry which is covered by the operability requirement and

surveillances has been reviewed and approved by the NRC. Therefore,

operation of ONS [Oconee Nuclear Station] in accordance with this

Technical Specification amendment will not create any failure modes

not bounded by previously evaluated accidents. Consequently, this

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

accident from any kind of accident previously evaluated.

(3) [Does not] involve a significant reduction in a margin of

safety:

The design basis of auxiliary electrical systems is to supply

the required ES loads of one Unit and safe shutdown loads of the

other two units. The ability of the Keowee Hydro units to provide

emergency power following an accident during a period of Keowee

Hydro commercial power generation was reviewed and approved by the

NRC in [an] SER dated August 15, 1995. Therefore, there will be no

significant reduction in 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: Oconee County Library, 501

West South Broad Street, Walhalla, South Carolina 29691

Attorney for licensee: J. Michael McGarry, III, Winston and Strawn,

1200 17th Street, NW., Washington, DC 20036

NRC Project Director: Herbert N. Berkow

Duke Power Company, Docket Nos. 50-269, 50-270 and 50-287, Oconee

Nuclear Station, Units 1, 2 and 3, Oconee County, South Carolina

Date of amendment request: February 20, 1996

Description of amendment request: The proposed amendments would

revise Technical Specifications (TS) 3.1.5, 3.1.10, and 4.1. The TS

changes would: (1) reduce the frequency for the concentrated boric acid

storage tank boron concentration surveillance, (2) delete the chemical

and radiochemical surveillance requirements for the reactor

[[Page 13524]]

coolant for Sr189 and Sr190, gross beta

activity, gross alpha activity, dissolved gas concentration in the

reactor coolant, and gross beta activity in the steam generator

feedwater, and (3) relocate the surveillance requirements for tritium,

chloride, fluoride and oxygen to the Selected Licensee Commitments

(SLC) Manual. The proposed changes would also delete some temperature

and pressure requirements on control rod operation.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration. The licensee has determined that operation of the

facility in accordance with the proposed amendments would not:

(1) Involve a significant increase in the probability or

consequences of an accident previously evaluated:

Each accident analysis addressed within the Oconee Final Safety

Analysis Report (FSAR) has been examined with respect to the

proposed amendment request. The probability of any Design Basis

Accident (DBA) is not significantly increased by the proposed

amendment due to the fact that the identified cause in the FSAR

accidents is not impacted. In addition, the consequences of the

accidents are within the bounds of the FSAR analyses since the

proposed amendment does not change the accident analysis methods or

assumptions described in the FSAR.

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

accident from any kind of accident previously evaluated:

The proposed amendment revises and eliminates several of the RCS

[Reactor Coolant System] chemistry Technical Specification

surveillance requirements. The changes in the surveillance

requirements do not alter the plant safety features or the method of

operation at ONS [Oconee Nuclear Station]. Therefore, operation of

ONS in accordance with the proposed Technical Specification will not

create any failure modes not bounded by previously evaluated

accidents.

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

The proposed amendment does not impact the mitigation of any of

the accidents analyzed in the FSAR. Therefore, there is not a

significant reduction in the margin of safety associated with the

proposed amendment.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Oconee County Library, 501

West South Broad Street, Walhalla, South Carolina 29691

Attorney for licensee: J. Michael McGarry, III, Winston and Strawn,

1200 17th Street, NW., Washington, DC 20036

NRC Project Director: Herbert N. Berkow

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

Nuclear Station, Unit 1, Claiborne County, Mississippi

Date of amendment request: February 22, 1996

Description of amendment request: The licensee has proposed to

increase the safety function lift setpoint tolerances for the safety

and relief valves that are listed in Surveillance Requirement 3.4.4.1

(Page 3.4-10) of the Technical Specifications TSs) for the Grand Gulf

Nuclear Station, Unit 1. The tolerances would be increased from the

current plus/minus 1 percent of the safety function (i.e., safety

relief valve) lift setpoint to plus/minus 3 percent.

The frequency of verifying these setpoints would not be changed by

this amendment request. Also, the other surveillance requirements in

the TSs on these valves and the number of these valves required to be

operable are not being changed by this amendment request.

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 (NSHC) in Attachment 2 to its application of February 22,

1996.

In its application, the licensee stated that it has used the NRC

staff's safety evaluation report (SER), NEDC 31753-P-A, issued in the

NRC letter of March 8, 1993, which evaluated General Electric (GE)

topical report NEDC-31753P, ``BWROG [BWR Owners' Group] In-Service

Pressure Relief Technical Specification Revision Licensing Topical

Report,'' dated February 1990.

The licensee's NSHC analysis is presented below:

Entergy Operations, Inc. is proposing that the Operating License

for Grand Gulf Nuclear Station (GGNS) be amended to increase the

tolerance of the safety function lift setpoints [from plus/minus 1%]

to plus/minus 3%. The GGNS Inservice Testing (IST) program controls

the frequency of safety relief valve (S/RV) testing as required by

the GGNS Operating License; therefore, this proposal will also

incorporate changes [concerning the setpoint tolerances] to

applicable IST procedures. GGNS will incorporate the recommendations

of the NEDC-31753-P-A [NRC staff's] SER, by resetting the safety

function [S/RV] lift setpoints for all tested valves to within plus/

minus 1% of the design lift setpoint and increasing the test sample

size by two valves for each valve found outside the plus/minus 3%

safety function lift setpoint. S/RV test sample population will be

determined based upon the currently licensed ASME [American Society

of Mechanical Engineers] Boiler and Pressure Vessel Code.

The commission has provided standards for determining whether a

no significant hazards consideration exists as stated in

10CFR50.92(c). A proposed amendment to an operating license involves

no significant hazards if the 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 considerations in its request for a license amendment. In

accordance with 10CFR50.91(a), Entergy Operations, Inc. is providing

the following analysis of the proposed amendment against the three

standards in 10CFR50.92(c):

a. No significant increase in the probability or consequences of

an accident previously evaluated results from this change.

The GGNS safety design bases for the S/RVs are:

) Prevent overpressurization of the nuclear system that could

lead to failure of the reactor coolant pressure boundary,

) Provide automatic depressurization for small breaks in the

nuclear system,

) Permit verification of operability,

) Withstand adverse combinations of loadings and forces during

abnormal, accident, or special event conditions.

The most limiting vessel overpressurization event is a closure

of all main steam isolation valves with a high flux scram. This

event was analyzed for GGNS using the minimum number of S/RVs

required by the GGNS Operating License. The safety function lift

setpoint tolerance used in the analysis bounds the proposed plus/

minus 3% setpoint tolerance. The analysis indicates that the S/RVs

are capable of maintaining adequate margin below the Operating

License Reactor Coolant System Pressure of 1325 psig.

Anticipated operational transients can also challenge the

operation of the S/RVs, for instance, Generator Load Reject without

Bypass. Analyses have been performed on the limiting events that

bound other pressure transient events using safety function limit

setpoint tolerances that bound the proposed plus/minus 3% tolerance

request. Fuel operating limits are based on the results of these

analyses; therefore, adequate fuel thermal margin is maintained.

Plant transients and events that require the use of automatic

depressurization and the low-low set feature utilize the relief mode

of S/RV operation. This proposed change does not affect the relief

mode of S/RV operation.

The verification of valve operability will still be performed in

accordance with the GGNS Inservice Testing Program, and S/RV safety

mode operability will be verified prior to reinstallation. Analysis

of the loads placed on each S/RV sub-system (discharge piping,

spargers and associated components) verifies that adequate margin

exists to ensure that the

[[Page 13525]]

overpressurization system can perform its designed function.

The negative tolerance of the safety function lift setpoint

remains above the highest setpoint of the S/RV relief mode, and

therefore normal vessel pressure. This margin provides reasonable

assurance that inadvertent opening of an S/RV will not occur during

power operations.

GGNS will replace each S/RV removed for IST program testing with

an S/RV that has been reset to within plus/minus 1% of the designed

safety function lift setpoint. During each refueling outage, at

least six of the installed S/RVs will be tested for safety lift

setpoint in accordance with the current IST program plant

procedures. This sample population is in agreement with the current

ASME Boiler and Pressure Vessel Code requirements for the GGNS IST

program, and is more restrictive than the ANSI/ASME OM-1-1981

requirement upon which the setpoint tolerance was based. For S/RV

setpoint testing ([the] as-found [setpoint]), additional valves will

be tested if the as-found setpoint is outside plus/minus 3% of its

designed safety function lift setpoint. Sample expansion will be

consistent with the NEDC 31753-P-A SER requirement of two additional

valves per valve failure.

The GGNS UFSAR currently requires at least fifty percent of the

installed valves to be removed and tested during each refueling

outage. GGNS FSAR Questions & Responses 211.49 discusses

the bases for this requirement. The concern regarded the performance

of S/RVs installed in operating plants at the time of GGNS

construction and licensing, and that new plants should have

significantly better performing S/RVs. The fifty percent requirement

provides a very conservative margin of testing to demonstrate that

no common cause of S/RV failure occurs within any one operating

cycle. The minimum testing of six valves proposed for each outage,

with additional testing for each failure from the initial test

population, provides reasonable assurance that no common cause

failure is occurring without early detection. [The minimum testing

of six valves is in agreement with the current ASME Code

requirements and is consistent with the current industry practices

that was accepted in the NRC staff's safety evaluation report, NEDC

31753-P-A.]

One of the major factors in the requirement of additional

testing population beyond ASME Boiler and Pressure Vessel Code is

many of the older plants were experiencing failures with multiple

stage pilot operated S/RVs. The safety function of this type of S/RV

requires operation of a pilot valve that is susceptible to excessive

leakage and corrosive bonding to cylinder walls; thereby preventing

proper safety function operation. The GGNS Dikkers S/RVs are direct

acting, and do not require the operation of a pilot valve for the

safety function. The Dikkers S/RV Instruction Manual recommends ``to

replace part of the installed valves each maintenance stop

(refueling outage)'', and does not prescribe any particular [number

of valves to be tested].

Therefore, no significant increase in the probability or

consequences of an accident previously evaluated results from this

proposed change.

b. This change would not create the possibility of a new or

different kind of accident from any previously analyzed.

The plant specific analyses verify that each S/RV will still

perform the intended function of preventing overpressurization of

the nuclear system. The vessel will have adequate margin below the

Operating License Reactor Coolant System Pressure of 1325 psig, and

plant system response will not deviate from the expected sequence of

events. Each system, structure, and component that communicates with

the reactor vessel has been verified to be within its design and

operational margin, and no unanticipated plant transients will occur

as a result of the safety lift function setpoint tolerance change.

The negative tolerance of the safety function lift setpoint

remains above the highest setpoint of the S/RV relief mode, and

therefore normal vessel pressure. This margin provides reasonable

assurance that inadvertent opening of an S/RV will not occur during

power operations.

This proposed change does not add any new systems, structures or

supports, nor does it introduce new S/RV operating modes.

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

or different kind of accident from any previously analyzed.

c. This change would not involve a significant reduction in the

margin of safety.

The increase in the S/RV safety function lift tolerance has been

analyzed for bounding limiting events and accident conditions. [The

safety function lift setpoint tolerance used in the analysis bounds

the proposed plus/minus 3% setpoint tolerance.] No condition exists

that reduces the margin of safety on the reactor coolant pressure

boundary or any system, structure or component that is required to

operate during vessel overpressurization events. Fuel operating

limits are based on the results of these analyses; therefore,

adequate fuel thermal margin is maintained.

[The negative tolerance of the safety function lift setpoint

remains above the highest setpoint of the S/RV relief mode, and

therefore normal vessel pressure. This margin provides reasonable

assurance that inadvertent opening of an S/RV will not occur during

power operations.]

Therefore, this change would not involve a significant reduction

in the margin of safety.

Based on the above evaluation, Entergy Operations, 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

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

Nuclear Generating Station, Ocean County, New Jersey

Date of amendment request: February 22, 1996

Description of amendment request: The amendment proposes to delete

a specification which requires a thorough inspection of the Emergency

Diesel Generator (EDG) every 24 months during shutdown. In addition

this Technical Specification proposes to delete the phrase ``in any

thirty day period'' from a specification concerning Allowed Outage time

(AOT).

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:

GPU Nuclear has determined that this [technical specification

change request] TSCR poses no significant hazard as defined by the

NRC in 10 CFR 50.92.

1. State the basis for the determination that the proposed

activity will or will not increase the probability of occurrence of

the consequences of an accident.

The proposed activity deletes the requirement to inspect EDGs

during shut down from the Technical Specifications. It further

modifies the operability of a single EDG for a limited and defined

period of time. These changes do not affect the design or

performance of the EDGs or their ability to perform their design

function. Analysis using PRA techniques indicates the changes do not

significantly increase the probability or consequences of an

accident.

2. State the basis for the determination that the activity does

or does not create a possibility of an accident or malfunction of a

different type than any previously identified in the SAR.

The EDGs are not the source of any accident described in the

SAR. These changes do not modify the design or performance of the

EDGs and do not affect plant functions or actions. Therefore, the

proposed change does not create the possibility of an accident or

malfunction of a different type than those previously identified.

3. State the basis for the determination that the margin of

safety is not reduced. The proposed changes are designed to improve

EDG reliability and availability during shutdown periods by

providing flexibility in the scheduling and performance of

maintenance. The surveillance intervals are unchanged and

operability requirements are only modified to an acceptable degree.

The proposed activity does not alter the basis of

[[Page 13526]]

any technical specification that is related to the establishment or

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

safety is not significantly reduced by this action.

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, et al., Docket No. 50-219, Oyster Creek

Nuclear Generating Station, Ocean County, New Jersey

Date of amendment request: February 23, 1996

Description of amendment request: The proposed change to the

Technical Specifications would allow the implementation of 10 CFR 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:

GPU Nuclear has determined that this TSCR [technical

specification change request] involves no significant hazards

considerations as defined by NRC in 10 CFR 50.92.

The major changes from the existing Oyster Creek Technical

Specifications requested in accordance with the Option B

requirements:

1. Operation of the facility in accordance with the proposed

amendment would not involve a significant increase in the

probability of occurrence or the consequences of an accident or

malfunction of equipment important to safety as previously evaluated

in the Safety Analysis Report.

The proposed change implements Option B of 10 CFR 50, Appendix J

on performance based containment leakage testing. The proposed

change does not involve a change to the plant design or operation.

Therefore, the proposed change does not affect any of the parameters

or conditions that contribute to initiation of any of the analyzed

accidents or malfunctions. The proposed change does request an

allowable extension of containment testing. Therefore, a

hypothetical leak could remain undetected for a greater period of

time. This slight increase in risk has been determined to be

insignificant as:

Type A Testing

NUREG 1493 determined that the effect of containment leakage on

overall accident risk is small as risk is dominated by accident

sequences that result in the failure or bypass of the containment.

Industry wide PCILRTs have demonstrated that only a small fraction

of the leaks discovered during testing exceeded acceptance criteria,

and that the leak rate has been only marginally above the acceptable

limit. Only 3% of all leaks can be detected only by PCILRT,

therefore, only 3% of the theoretical leaks are affected by the

extension to the Type A test interval. Experience at Oyster Creek

agrees with the industry wide data in that the majority of the

detected leakage from the primary containment is found through Type

B and C testing.

NUREG 1493 found that these observations, together with the

insensitivity of reactor accident risk to the containment leakage

rate, demonstrates that increasing the Type A leakage test intervals

would have a minimal impact on public risk.

Type B and C Testing

Penetrations are designed to ensure reliability of the

containment isolation function. Type B penetrations use a double

passive seal (e.g. o-ring, gasket) and Type C penetrations use a

double isolation valve design to ensure reliability of the isolation

function. Because valves perform the isolation function actively,

they are more likely to fail on demand (e.g. failure to completely

close on demand). To address this failure mode, Type C valves are

subjected to increased design constraints and testing to ensure both

acceptable leak rates and stroke times. The proposed change does not

alter the installation, operation, operating environment, or testing

method of these valves. Therefore, the proposed change does not

introduce any new component failure modes, nor does it affect the

probability of occurrence of any existing evaluated failure mode.

The failure of any single penetration barrier (isolation valve

or passive seal) does not cause penetration failure. Therefore, a

double failure would have to occur to cause a failure of the

penetration and affect containment. Additionally, the proposed

change does not change the acceptance criteria for acceptable

leakage testing.

The proposed change does not alter plant design or operation,

nor does it alter the allowable maximum leakage rate limit. Thus,

the proposed change does not affect the probability of occurrence

nor the consequences of any evaluated accident or malfunction of

equipment important to safety.

2. Operation of the facility in accordance with the proposed

amendment would not create the possibility of an accident or

malfunction different from any accident or malfunction previously

evaluated.

The proposed change does not involve a change to the plant

design or operation. As a result, the proposed change does not

affect any of the parameters or conditions that could contribute to

initiation of any accidents. This change only involves the reduction

in Type A, B and C test frequencies, and the Type A test pressure.

Type A Testing

The only changes proposed to the Type A testing are to frequency

and test pressure. As the proposed test pressure is grater than the

existing test pressure, no new type of accident or malfunction is

created, and the increase in pressure provides an additional margin

of safety. The increase in pressure provides an additional margin of

safety. The increase in surveillance interval cannot introduce any

new type of accident or malfunction.

The PCILRT is presently performed at 20 psig. Performance of the

PCILRT at Pa (35 PSIG) will provide a more direct leak rate for

analysis.Pa is the design pressure of the torus (the drywell

design pressure is 44 psig, but the torus is non isolable form the

drywell. Therefore, Pa will not create the possibility of the

failure of the torus due to overpressurization. No new accident

modes can be created by extending the test intervals. No safety

related functions or components are altered as a result of this

change. Therefore, no new accident or malfunction different form

those evaluated in the Safety Analysis Report can result due to the

increase in test pressure or increase in surveillance interval.

Type B and C Testing

The proposed change only deals with the frequency of performing

Type B and C testing. It does not change what components are tested

or the method of testing. There is no proposed change to the design

or operation of the plant. Therefore, no new accident or malfunction

different form those evaluated in the Safety Analysis Report can

result due to the increase in test pressure or increase in

surveillance interval.

3. Operation of the facility in accordance with the proposed

amendment would not decrease the margin of safety as defined in the

bases of the Technical Specifications.

Type A Testing

Except for the method of defining the test frequency and

pressure at which the PCILRT is performed, the methods for

performing the actual test are not changed. However, the proposed

change can increase the probability that an increase in leakage

could go undetected for an extended period of time. NUREG 1493 has

determined that under several different accident scenarios, the

increased risk of radioactivity release from containment is

negligible with the implementation of these proposed changes.

Type B and C Testing

The proposed change only affects the frequency of Type B and C

testing. The methods for performing the actual test are not changed.

The design or operation of Type B and C components are not changed.

The proposed change will result in a longer interval between tests

of good performing Type B and C components.

The margin of safety that has the potential of being impacted by

the proposed change involves the offsite dose consequences of

postulated accidents which are directly related to containment

leakage rate. The containment isolation system is designed to limit

leakage to La, which is defined by the Oyster Creek Technical

Specifications to be 1.0 percent by weight of the containment air at

35 psig per 24 hours. The limitation on

[[Page 13527]]

containment leakage rate is designed to ensure the total leakage

volume will not exceed the value assumed in the accident analyses at

the peak accident pressure (Pa). The margin of safety for the

offsite dose consequences of postulated accidents directly related

to the containment leakage rate is maintained by meeting the 1.0

La acceptance criteria. The La value is not being modified

by this proposed Technical Specification change request.

Therefore, the margin of safety as defined in the bases for the

Technical Specification will not be reduced.

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

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

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

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

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 amendment requests: February 22, 1996 (AEP:NRC:0659AA)

Description of amendment requests: The proposed amendments would

revise the technical specifications to remove the requirement that the

Operations Superintendent must hold or have held a Senior Operator

License at Cook Nuclear Plant, or a similar reactor. In addition, a

mid-level operations manager will only be required to hold a Senior

Operator License if the Operations Superintendent does not hold one.

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:

Per 10 CFR 50.92, this proposed change does not involve a

significant hazards consideration because the change does 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.

Criterion 1

The amendment request does not involve a significant increase in

the probability or consequences of [an] accident previously

evaluated because the proposed change to the Technical Specification

does not affect the assumptions, parameters, or results of any UFSAR

[updated final safety analysis report] accident analysis. The

proposed amendment does not modify any existing equipment. It is

concluded that the changes do not involve a significant increase in

the probability or consequences of an accident previously evaluated.

Criterion 2

The proposed change does not involve physical changes to the

plant or changes in plant operating configuration. The proposed

change updates the requirements for the Operations Superintendent.

Thus, it is concluded that the proposed changes do not create the

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

previously evaluated.

Criterion 3

The proposed change updates the requirements for Operations

Superintendent. 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 requests involve no significant hazards consideration.

Local Public Document Room location: Maud Preston Palenske Memorial

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

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

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

NRC Project Director: John N. Hannon

Indiana Michigan Power Company, Docket No. 50-316, Donald C. Cook

Nuclear Plant, Unit No. 2, Berrien County, Michigan

Date of amendment request: March 12, 1996 (AEP:NRC:1248)

Description of amendment request: The proposed amendment would

remove the technical specifications related to shutdown and control rod

position indication while in modes 3, 4, and 5. The change would make

the Unit 2 technical specifications consistent with the Unit 1

technical specifications and the 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:

Per 10 CFR 50.92, this proposed change does not involve a

significant hazards consideration because the change does 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.

Criterion 1

The boron concentration in the reactor coolant system will be

high enough to assure adequate SDM in modes 3, 4, and 5. The

calculation to obtain the required boron concentration takes into

account the position of the rods. Shutdown margin is assumed as an

initial condition in the safety analysis. The safety analysis

establishes a SDM that ensures specified acceptable fuel design

limits are not exceeded. As long as the SDM is satisfied, no change

in the probability or consequences of an accident previously

evaluated will result from the proposed deletion of the ``position

indicator - shutdown'' specification. It is noted that this change

is consistent with the new ISTS approved by the NRC as NUREG-1431,

Rev. 1.

Criterion 2

The ability to insert the control and shutdown rods provided by

the rod control system is not affected by the OPERABILITY status of

the ARPI system. As mentioned previously, the reactor coolant system

boron concentration will be high enough to assure adequate SDM is

maintained. Therefore, it is concluded that the proposed changes do

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

from any accident previously evaluated.

Criterion 3

The margin of safety requirements are not affected by the

removal of this T/S. The required SDM which is an initial condition

in the safety analysis, is unaffected since the reactor coolant

system boron concentration is increased to address the potential

``all rods out'' configuration. Based on these considerations, it is

concluded that 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: Maud Preston Palenske Memorial

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

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

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

NRC Project Director: John N. Hannon

Maine Yankee Atomic Power Company, Docket No. 50-309, Maine Yankee

Atomic Power Station, Lincoln County, Maine

Date of amendment request: November 29, 1995

Description of amendment request: The proposed amendment would

[[Page 13528]]

modify the Technical Specifications to remove the requirement for

additional pressure relief by a residual heat removal (RHR) spring

relief valve during low temperature overpressure protection (LTOP)

conditions.

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 to delete Technical Specification 3.4.D.3b

has been evaluated against the standards of 10 CFR 50.92 and has

been determined not to involve a significant hazards consideration.

This proposed change does not:

1. Involve a significant increase in the probability or

consequence of an accident previously analyzed. The Power Operative

Relief Valves (PORVs) remain operable to mitigate any LTOP event.

Thus, this change does not result in an increase in the probability

or consequences of an accident previously analyzed.

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

from any previously evaluated. Removing the RHR spring relief valve as

an additional relief requirement does not create the possibility of a

new or different kind of accident since the proposal involves neither a

hardware modification nor the creation of a unique operating condition.

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

Removing the RHR spring relief valve as an additional requirement

does not change the results of any of the FSAR Chapter 14 events.

The PORVs remain operable to maintain 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: Wiscasset Public Library, High

Street, P.O. Box 367, Wiscasset, ME 04578

Attorney for licensee: Mary Ann Lynch, Esquire, Maine Yankee Atomic

Power Company, 329 Bath Road, Brunswick, ME 04011NRC Deputy Director:

John Zwolinski

Maine Yankee Atomic Power Company, Docket No. 50-309, Maine Yankee

Atomic Power Station, Lincoln County, Maine

Date of amendment request: November 29, 1995

Description of amendment request: The proposed amendment would

modify Technical Specification (TS) 3.14 to decrease the maximum steam

generator (SG) primary-to-secondary leakage rate from 0.15 gpm to 0.10

gpm and would modify TS 4.10 by revising the requirements for

unscheduled SG tube inspections that are performed on each SG following

a primary-to-secondary tube leak.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed changes do not involve a significant increase in

the probability or consequences of an accident previously evaluated.

A steam generator leakage assumption greater than the proposed 0.10

gpm/SG limit has been used in the FSAR [Final Safety Analysis

Report] Chapter 14 safety analyses. Thus, the FSAR Chapter 14 safety

analyses remain bounding. Assuring that an adequate leakage limit

exists that initiates corrective actions in a timely manner is

important to ensuring a steam generator tube rupture event does not

take place. This change modifies the steam generator post-leakage

testing requirements to focus inspections on leaking tubes and areas

likely to produce similar leakage, in lieu of an expanded test

campaign of all three steam generators. Without this change,

Technical Specifications require inspection of 3% of the tubes in

each steam generator. By inspecting the critical areas of the

affected steam generator and possibly expanding inspections to the

critical areas of the remaining steam generators, the probability

and/or consequences of previously evaluated accidents (e.g., steam

generator tube rupture) are not increased.

2. The proposed changes do not create the possibility of a new

or different kind of accident from any accident previously

evaluated. The proposed changes will not involve a modification to

existing hardware at the plant. The decrease in the maximum

allowable steam generator primary leakage rate tends to provide

additional time for operator action to take place which, if timely

enough, would avoid the consequences of a tube rupture event. The

proposed inspection campaign requires inspection of the critical

area and may be expanded to the other steam generators to ensure

that additional tubes will not fail due to similar causes. This

modified inspection campaign does not introduce the possibility of a

new or different kind of accident from any accident previously

evaluated.

3. The proposed changes do not involve a significant reduction

in a margin of safety. The FSAR Chapter 14 safety analyses assume a

higher steam generator leakage rate and therefore remain

conservative. The proposed reduction in the allowable leakage

provides a greater margin of safety since it is more conservative

than the present value. This change modifies inspection requirements

of Technical Specifications and does not impact the plant design or

equipment. The modified inspection requirements following a plant

shutdown due to tube leakage concentrate steam generator tube

inspections in those areas believed to be most susceptible to flaws.

For these reasons, we believe the proposed changes increase the

margin of safety by inspecting the critical areas of the steam

generator(s) in lieu of additional random inspections.

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: Wiscasset Public Library, High

Street, P.O. Box 367, Wiscasset, ME 04578

Attorney for licensee: Mary Ann Lynch, Esquire, Maine Yankee Atomic

Power Company, 329 Bath Road, Brunswick, ME 04011NRC Deputy Director:

John Zwolinski

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

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

Date of amendment request: November 8, 1995

Description of amendment request: The amendment request would

revise the Technical Specifications (TS) for the jet pumps to be

consistent with the limiting conditions for operation and surveillance

requirements in the Standard Technical Specifications for General

Electric Plants (NUREG-1433).

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 an [significant hazards

consideration] SHC because the change would not:

1. Involve a significant increase in the probability or

consequences of an accident previously analyzed.

The new LCO [Limiting Condition for Operation] does not diminish

the existing requirement that all jet pumps must be operable, nor

does it affect the time available to achieve cold shutdown should a

pump become inoperable. The new LCO does eliminate the ability to

continue to operate with the indication (but not the function) of a

single jet pump inoperable. This does not increase the possibility

of an unnecessary plant shutdown due to inoperable instrumentation

since sufficient flexibility exists in the surveillance requirement

so that operability of the jet pumps can be verified. This change

eliminates the LCO that allowed continued operation with conditions

that could potentially mask an inoperable pump. The new LCO is more

limiting in ensuring that the plant is operated in a condition for

which accidents were analyzed.

The new surveillance requirement provides a more accurate method

of ensuring

[[Page 13529]]

the jet pumps remain operable. The new surveillance criteria are

more sensitive to jet pump failures and the degradation of the jet

pumps prior to failure.

Based on the above, the proposed change does not involve a

significant increase in 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 new LCO and surveillance does not change the manner in which

the plant is operated, nor does it reduce the operability

requirements of any jet pump, Therefore, no new or different kind of

accident can be created by the new specification. The surveillances

that will be performed do not require any new hardware or plant

evolutions. Therefore, the proposed change to the LCO and

surveillance cannot create the possibility of a new or different

kind of accident.

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

The margin of safety that currently exists is not diminished by

this change. The requirement to place the reactor in cold shutdown

within 24 hours should a jet pump become inoperable is maintained.

The LCO which allowed continued operation with indication for one

pump inoperable has been eliminated.

The new surveillance requirement continues to demonstrate the

operability of the jet pumps and during operation, continues to be

performed at the same interval as in the current technical

specifications. The note (which allows the surveillance to be

deferred until four hours after the associated recirculation loop is

in operation and 24 hours after exceeding 25% of rated thermal

power) does not significantly affect the margin of safety. The time

that the unit would be operating in these conditions would be small,

and the stress placed on the pump at less than 25% power is lower.

Based on the above, 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: Learning Resources Center,

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

Norwich, CT 06360.

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

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

CT 06141-0270.

NRC Project Director: Phillip F. McKee

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

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

Connecticut

Date of amendment request: November 3, 1995

Description of amendment request: The proposed amendment will

extend the allowed outage time from 48 hours to 7 days for an emergency

core cooling system train that is declared inoperable as a result of an

inoperable low pressure safety injection subsystem.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

Pursuant to 10CFR50.92, Northeast Nuclear Energy Company (NNECO)

has reviewed the proposed change to extend the allowed outage time

(AOT) for an inoperable low pressure safety injection (LPSI)

subsystem from the existing limit of 48 hours to 7 days. In

addition, the change to modify the completion time for the Action

Statement and the criteria for the Surveillance Requirements were

also reviewed. NNECO concludes that these changes do not involve a

significant hazards consideration (SHC) since the proposed change

satisfies the criteria in 10CFR50.92(c). That is, the proposed

change does not:

1. Involve a significant increase in the probability or

consequences of an accident previously analyzed.

The proposed amendments for Millstone Unit No. 2 will extend the

action completion AOT for a single inoperable LPSI train from 48

hours to 7 days. A LPSI subsystem is designed as a part of each

emergency core cooling system (ECCS) train to supplement safety

injection tank inventory during the early stages of mitigating a

design basis accident (DBA). As such, components of the LPSI

subsystem are not accident initiators, and an extended AOT to

restore operability of an inoperable LPSI subsystem would not

increase the probability of occurrence of accidents previously

analyzed.

The safety analyses for Millstone Unit No. 2 demonstrates that

ECCS performance acceptance criteria are satisfied with only one of

the two redundant ECCS trains operating during the postulated DBA.

The proposed technical specification revisions involve the AOT for a

single inoperable LPSI subsystem, and do not change the conditions

assumed for the minimum amount of operating equipment needed for

accident mitigation. Therefore, the consequences of an accident

previously evaluated will not be significantly increased.

In addition, CE NPSD-995 recognizes that when an ECCS train is

inoperable due to a LPSI subsystem being unavailable, due either to

being declared inoperable (by failing a surveillance requirement) or

is intentionally taken out-of-service (for corrective or preventive

maintenance), the core damage frequency (CDF) during power operation

increases. The results of the PRA presented in CE NPSD-995 show that

the proposed increase in the ECCS AOT (due to LPSI unavailability)

from 48 hours to 7 days does not cause a significant increase in the

overall CDF of Millstone Unit No. 2.

The analyses indicate that continued plant operation with a

single LPSI subsystem out-of-service may result in a small increase

in ``at power risk;'' however, that risk increase will be negligibly

small and controlled effectively via the Maintenance Rule and the

risk monitor program that minimizes the outage time and prevents

entering into an unacceptable risk configuration. In addition, the

proposed AOT extension for the LPSI subsystem is evaluated as having

negligible impact on the large early radiological release

probability for Combustion Engineering pressurized water reactors in

the event of a design basis accident.

Therefore, operation in accordance with the proposed amendment

would not involve a significant increase in the probability or

consequences of an accident previously evaluated.

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

from any previously analyzed.

The proposed amendment will not change the physical plant or the

modes of plant operation defined in the technical specifications.

The changes do not involve the addition or modification of equipment

nor do they alter the design of plant systems. Therefore, operation

of Millstone Unit No. 2 in accordance with its proposed amendment

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

accident from any accident previously evaluated.

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

The margin of safety associated with the ECCS train is

established by acceptance criteria for system performance defined in

10CFR50.46. The proposed amendment will not change this acceptance

criteria nor the operability requirements for equipment that is used

to achieve such performance as demonstrated in the Millstone Unit

No. 2 safety analyses. Moreover, an integrated assessment of the

risk impact of extending the AOT for a single inoperable LPSI train

has concluded that the risk contribution is small. Therefore,

operation of Millstone Unit No. 2 in accordance with its proposed

amendment would not involve a significant reduction in a margin of

safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Learning Resources Center,

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

Norwich, CT 06360.

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

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

CT 06141-0270.

[[Page 13530]]

NRC Project Director: Phillip F. McKee

Northeast Nuclear Energy Company, et al., Docket No. 50-423,

Millstone Nuclear Power Station, Unit No. 3, New London County,

Connecticut

Date of amendment request: September 12, 1995

Description of amendment request: The amendment would revise and

reformat Technical Specification (TS) 6.3.1 to add the requirement that

the Assistant Operations Manager shall hold a senior reactor operator

(SRO) license if the Operations Manager does not hold an SRO license

for Millstone Unit 3. Also the footnote would be deleted from TS 6.3.1

that previously granted a one-time three year exception to the

qualification requirements for the Operations Manager and an exception

for the Assistant Operations Manager to hold a license instead of the

Operations Manager.

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 an [significant hazards

consideration] SHC because the change would not:

1. Involve a significant increase in the probability or

consequences of an accident previously analyzed.

The proposed change affects an administrative control, which was

based on the guidance of ANSI N18.1-1971. ANSI N18.1-1971

recommended that the Operations Manager hold an SRO license. The

current guidance in Section 4.2.2 of ANSI/ANS 3.1-1987 recommends,

as one option, that the Operations Manager have held a license for a

similar unit and the Operations Middle Manager hold an SRO license.

While the Operations Middle Manager position does not exist at

Millstone Unit No. 3, [Northeast Nuclear Energy Company] NNECO has

created the position of Assistant Operations Manager. The individual

in this position would meet the requirements for, and would have

responsibilities as recommended in, ANSI/ANS 3.1-1987 for the

Operations Middle Manager position.

Therefore, the proposed change requests an exception to ANSI

N18.1-1971 to allow use of ANSI/ANS 3.1-1987 in a limited

circumstance. Specifically, the proposed revision to Technical

Specification 6.3.1 would require the Operations Manager to either

hold an SRO license at Millstone Unit No. 3 or have held an SRO at a

[pressurized water reactor] PWR.

If the Operations Manager does not hold an SRO license at

Millstone Unit No. 3, the specification will require the Assistant

Operations Manager to hold, and continue to hold, an SRO license.

The proposed change includes the requirement for the Operations

Manager to have held a license for a similar unit (a PWR) in

accordance with Section 4.2.2 of ANSI/ANS 3.1-1987. For those areas

of knowledge that require an SRO license, the Assistant Operations

Manager will provide the technical guidance normally provided by the

Operations Manager.

The proposed change does not alter the design of any system,

structure, or component, nor does it change the way plant systems

are operated. It does not reduce the knowledge, qualifications, or

skills of licensed operators, and does not affect the way the

Operations Department is managed by the Operations Manager. The

Operations Manager will continue to maintain the effective

performance of his personnel and ensure the plant is operated safely

and in accordance with the requirements of the operating license.

Additionally, the Control Room Operators will continue to be

supervised by the licensed Shift Supervisors.

The proposed change does not detract from the Operations

Manager's ability to perform his primary responsibilities. In this

case, by having previously held an SRO license, the Operations

Manager has achieved the necessary training, skills, and experience

to fully understand the operation of plant equipment and the watch

requirements for operators. In summary, the proposed change does not

affect the ability of the Operations Manager to provide the plant

oversight required of his position. Thus, it does not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

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

from any previously analyzed.

The proposed change to Technical Specification 6.3.1 does not

affect the design or function of any plant system, structure, or

component, nor does it change the way plant systems are operated. It

does not affect the performance of NRC licensed operators. Operation

of the plant in conformance with technical specifications and other

license requirements will continue to be supervised by personnel who

hold an NRC SRO license. The proposed change to Technical

Specification 6.3.1 ensures that the Operations Manager will be a

knowledgeable and qualified individual to have held an SRO license

at a PWR. Based on the above, the proposed change does not create

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

previously evaluated.

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

The proposed change involves an administrative control that is

not related to the margin of safety. The proposed change does not

reduce the level of knowledge or experience required of an

individual who fills the Operations Manager position, nor does it

affect the conservative manner in which the plant is operated. The

Control Room Operators will continue to be supervised by personnel

who hold an SRO license. Thus, the proposed change does not involve

a significant reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Learning Resources Center,

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

Norwich, CT 06360.

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

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

CT 06141-0270.

NRC Project Director: Phillip F. McKee

Northeast Nuclear Energy Company, et al., Docket No. 50-423,

Millstone Nuclear Power Station, Unit No. 3, New London County,

Connecticut

Date of amendment request: November 21, 1995

Description of amendment request: The licensee proposes to change

Technical Specification Section 1.33 and Bases Sections 3/4.3.3.9 and

3/4.3.3.10, and 3/4.11.2.1. The changes clarify the definition of

source check to include a source check from a light emitting diode

(LED), as well as from ionizing radiation.

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:

... NNECO concludes that these changes do not involve a

significant hazards consideration since the proposed changes satisfy

the criteria in 10CFR50.92(c). That is, the proposed changes do not:

1. Involve a significant increase in the probability or

consequences of an accident previously analyzed.

The proposed changes to the definition of source check clarifies

the source check for the liquid and gaseous effluent radiation

monitors. These monitors do not provide a safety function and only

serve to provide radiological information to plant operators,

therefore, the changes will not increase the probability or

consequences of an accident previously evaluated.

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

from any previously analyzed.

The proposed changes to the definition of source check have no

effect on the ability of the monitors to perform their designed

function. The clarification to the surveillance do not involve any

physical modifications to any equipment, structures, or components.

The monitors already have the internal LEDs which were originally

used to perform the source check. The proposed changes have no

impact on design basis accidents, and the changes will not modify

plant response or create a new or unanalyzed event.

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

[[Page 13531]]

The proposed changes to the definition of source check do not

have any impact on the protective boundaries and, therefore, have no

impact on the safety limits for these boundaries. The

instrumentation associated with these changes do not provide a

safety function and only serve to provide radiological information

to plant operators. The instrumentation has no affect on the

operation of any safety-related equipment. As such, these changes

have no impact on the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Learning Resources Center,

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

Norwich, CT 06360.

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

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

CT 06141-0270.

NRC Project Director: Phillip F. McKee

PECO Energy Company, Public Service Electric and Gas Company,

Delmarva Power and Light Company, and Atlantic City Electric

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

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

Date of application for amendments: February 15, 1996

Description of amendment request: The amendment changes the

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

Option B, by creating Technical Specification Section 5.5.12, ``Primary

Containment Leakage Rate Testing Program,'' which refers 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 changes do not involve a significant increase in

the probability or consequences of any accident previously

evaluated.

The adoption of 10 CFR 50, Appendix J Option B will not involve

a significant increase in the probability or consequences of any

accident previously evaluated. The proposed changes to the TS

[Technical Specifications] reflect the use of the performance-based

containment leakage-testing program. The USNRC has approved the use

of a performance-based option for containment leakage testing

programs when it amended 10 CFR 50, Appendix J (60 FR 49495). For

adoption of the revised regulation, licensees are required to

incorporate into their TS, by general reference, the USNRC

regulatory guide or other plant-specific implementing document used

to develop their performance-based leakage testing program. A new

Administrative Control subsection (5.5.12, ``Primary Containment

Leakage Rate Testing Program'') has been added that requires the

establishment and maintenance of a Primary Containment Leakage Rate

Testing Program. The TS will still require the performance of a

periodic general visual inspection of the containment to ensure

early detection of any structural deterioration of the containment

that may occur.

As concluded in NUREG-1493, given the insensitivity of risk to

containment leakage rate and the small fraction of leakage paths

detected solely by ILRT [Integrated Leak Rate Test] testing,

increasing the interval between ILRTs is possible with minimal

impact on public risk. Additionally, performance-based alternatives

to current LLRT [Local Leak Rate Test] requirements are feasible

without significant risk impacts. Additionally, these changes will

not alter any safety limits which ensure the integrity of fuel

barriers, and will not result in a significant increase to onsite or

offsite dose.

No physical changes are being made to the plant, nor are there

any changes being made in the operation of the plant as a result of

these changes which could involve a significant increase in the

probability or consequences of any accident previously evaluated.

Additionally, these changes will not alter the operation of

equipment assumed to be available for the mitigation of accidents or

transients.

2) The proposed changes do not create the possibility of a new

or different kind of accident from any previously evaluated.

The adoption of 10 CFR 50, Appendix J Option B will not create

the possibility of a new or different type of accident from any

previously evaluated. These changes to the PBAPS, Units 2 and 3 TS

will not involve any changes to plant systems, structures or

components (SCCs) which could act as new accident initiators. These

changes will not impact the manner in which SSCs are tested such

that a new or different type of accident from any previously

evaluated could be created.

3) The proposed changes do not result in a significant reduction

in the margin of safety.

No margins of safety are reduced as a result of the proposed

adoption of 10 CFR 50, Appendix J Option B. As stated previously,

the USNRC has approved the use of this performance-based option for

containment leakage testing programs when it amended 10 CFR 50,

Appendix J (60 FR 49495). These changes will not impact core limits

or any other parameters that are used in the mitigation of a UFSAR

[Updated Final Safety Analysis Report] design-basis accident or

transient. Additionally, these changes do not introduce any hardware

changes, and will not alter the intended operation of plant

structures, systems or components utilized in the mitigation of

UFSAR design-basis accidents or transients. These changes will not

introduce any new failure modes of plant equipment not previously

evaluated.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Government Publications

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

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

Pennsylvania 17105.

Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and

General Counsel, PECO Energy Company, 2301 Market Street, Philadelphia,

Pennsylvania 19101

NRC Project Director: John F. Stolz

Pennsylvania Power and Light Company, Docket No. 50-387,

Susquehanna Steam Electric Station, Unit 1, Luzerne County,

Pennsylvania

Date of amendment request: January 26, 1996

Description of amendment request: The proposed amendment removes

three pressure relief valves from Technical Specification Table 3.6.3-

1, ``Primary Containment Isolation Valves,'' since these valves are no

longer needed to support the steam condensing mode of the residual heat

removal (RHR) system and are being removed from 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:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

With the prior deletion of the steam condensing mode of RHR and

the isolation of the high and low pressure interfaces, the three

pressure relief valves that are being removed from the plant have no

active function. Their passive function of maintaining system or

containment integrity will be fulfilled by blind flanges. Also, the

RHR and RCIC [reactor core isolation cooling] piping are provided

with overpressure protection from other pressure relief valves.

Therefore, the removal of these pressure relief valves does not

involve a significant increase in the probability or consequences of

an accident previously evaluated.

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

from any accident previously evaluated.

The pressure relief valves that are being removed had two

primary functions. First,

[[Page 13532]]

they provided overpressure protection for the RHR and RCIC piping

during the steam condensing mode of RHR. Since the steam condensing

mode has been deleted from the plant, these valves no longer have

that function. Also, overpressure protection of the RHR and RCIC

piping is provided by other existing pressure relief valves. Second,

these valves maintained system or containment integrity. When the

pressure relief valves are removed from the plant, they will be

replaced with blind flanges or equivalent that will maintain system

or containment integrity. Therefore, the removal of the three

pressure relief valves does not create the possibility of a new or

different kind of accident from any accident previously evaluated.

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

Since the steam condensing mode of RHR has been eliminated, the

three pressure relief valves have no active function. Their passive

function of maintaining system or containment integrity will be

fulfilled by blind flanges or equivalent. Also, overpressure

protection of RHR and RCIC piping is provided by other existing

pressure relief valves. Therefore, the removal of the three pressure

relief valves 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: Osterhout Free Library,

Reference Department, 71 South Franklin Street, Wilkes-Barre,

Pennsylvania 18701

Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts

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

NRC Project Director: John F. Stolz

Power Authority of the State of New York, Docket No. 50-333, James

A. FitzPatrick Nuclear Power Plant, Oswego County, New York

Date of amendment request: January 25, 1996

Description of amendment request: The amendment proposes to revise

the allowed out-of-service times for single inoperable Emergency Diesel

Generators (EDGs) to accommodate on-line maintenance of the EDGs. In

addition, two line item changes are proposed: (1) to improve safety by

reducing EDG testing at power; and (2) to revise the ac power

requirements during cold shutdown or refueling modes to make the James

A. FitzPatrick (JAF) Technical Specifications consistent with the

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:

Operation of the FitzPatrick plant in accordance with the

proposed Amendment would not involve a significant hazards

consideration as defined in 10 CFR 50.92, since it would not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

a. EMERGENCY DIESEL GENERATOR LCO [Limiting Conditions for

Operation] AT POWER

The proposed changes to the Technical Specifications will allow

longer Allowed Out of Service Times [AOTs] to perform necessary

repair and maintenance on individual Emergency Diesel Generators

while at power. This extended AOT will enhance scheduling of

preventive maintenance of individual EDGs without significantly

increasing the probability or consequences of an accident previously

evaluated. The risk evaluations contained in the JAF quantitative

analyses of the EDGs determined that the probability of an accident

by increasing the AOT for an individual EDG from 7 days to 14 days

is non-risk-significant. The primary reason for this low relative

risk is due to the designed redundancy and capability to respond to

an accident when a single diesel generator is out of service. LOCA

[loss-of-coolant accident] Analyses that assume the worst case line

break while an EDG is out of service indicate the plant can be

safely shut down with the remaining EDGs. Even if another EDG should

fail during the AOT, at least one Core Spray and one Residual Heat

Removal (RHR) Low Pressure Coolant Injection pump can provide the

required flow to bring the plant to safe shut down. Furthermore,

long term suppression pool and reactor shutdown cooling is provided

by any one of the three remaining RHR pumps for a single EDG out of

service or by two remaining RHR pumps assuming an additional EDG

failure during the AOT.

Increasing the EDG AOT 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 for a single EDG.

Therefore, the proposed change does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

b. EMERGENCY DIESEL GENERATOR LCO DURING PLANT SHUTDOWN

Changing the number of EDGs required during plant shutdown 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 change in

the LCO during shutdown.

c. EMERGENCY DIESEL GENERATOR SURVEILLANCE AT POWER OPERATION

The proposed change to the Technical Specification will reduce

the required number of tests to be performed when an EDG or EDG

System is inoperable. This proposed change to TS requirements

addresses the concern of excessive testing that could result in EDG

wear which is counter-productive to safety in terms of equipment

degradation and availability. This change is consistent with Generic

Letter 93-05 guidance for implementing such recommendations. The

proposed Technical Specifications will not result in a change to the

design or operation of the facility, therefore, this change will not

result in 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.

a. EMERGENCY DIESEL GENERATOR LCO AT POWER

Extending the AOT for an individual EDG does not necessitate

physical alteration of the plant or changes in parameters governing

normal plant operation. Thus, this change does not create the

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

previously evaluated for JAF plant.

b. EMERGENCY DIESEL GENERATOR LCO DURING PLANT SHUTDOWN

Changing the number of EDGs required during shutdown does not

necessitate physical alteration of the plant or changes in

parameters governing normal plant operation. Thus, this change does

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

from any accident previously evaluated for JAF plant.

c. EMERGENCY DIESEL GENERATOR SURVEILLANCE AT POWER OPERATION

The proposed change does not change design, operation or the

testing process. The nature of this change precludes the possibility

of a new or different kind of accident. The proposed change to

complete the required action does not involve any hardware changes,

nor changes to the operation of the equipment nor does it change the

ability of the equipment to perform its intended function.

Performing the testing on an extended time cannot initiate any type

of accident.

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

a. EMERGENCY DIESEL GENERATOR LCO AT POWER

As discussed above, the JAF quantitative evaluation determined

that the change in risk associated with extending the AOT for a

single EDG is non-risk-significant. In addition, the design provides

adequate redundancy for safe shut down during the AOT for a single

EDG out of service. This is supported by the LOCA analyses including

analyses for long term suppression pool and reactor shutdown

cooling.

b. EMERGENCY DIESEL GENERATOR LCO DURING PLANT SHUTDOWN

The margin of safety is not affected by changing the number of

EDGs required during shutdown. One offsite power source or one EDG

ensure the availability of the

[[Page 13533]]

required power to recover from postulated accident events during

shutdown. When the required number of operable systems is not met,

all work that could potentially initiate a postulated accident event

during shutdown is suspended.

c. EMERGENCY DIESEL GENERATOR SURVEILLANCE AT POWER OPERATION

The proposed change to Technical Specifications reduces testing

at reactor power. The overall effect is a net gain in plant safety

by avoiding the potential for unnecessary wear that could degrade

the EDGs at power. Implementation of these changes is consistent

with the guidance provided by the NRC in Generic Letter 93-05. The

proposed change to the EDG testing requirements does not reduce the

ability of the equipment to perform its intended safety function.

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: Reference and Documents

Department, Penfield Library, State University of New York, Oswego, New

York 13126.

Attorney for licensee: Mr. Charles M. Pratt, 1633 Broadway, New

York, New York 10019.

NRC Project Director: Ledyard B. Marsh

Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281,

Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia

Date of amendment request: January 30, 1996

Description of amendment request: The proposed Technical

Specifications change will delete the requirement that oxygen

concentrations for both normal and transient conditions not exceed

saturation when the reactor coolant is below 250 degrees F. The

Technical Specifications change will also eliminate the surveillance

requirement for reactor coolant chemistry sampling of chloride,

fluoride, and oxygen concentration during maintenance activities when

fuel is removed from the reactor vessel and the Reactor Coolant System

(RCS) is drained below the reactor vessel flange regardless of whether

the upper internal and/or vessel heat are in place or not.

Administrative result of the changes being made, capitalize Technical

Specifications defined terms to maintain consistency within the

Technical Specifications, and the word ``degrees'' is spelled-out when

referring to the Fahrenheit temperature, rather than using the symbol.

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 Surry Power Station in accordance

with the proposed changes will not:

(1) Involve a significant increase in the probability or

consequences of an accident previously evaluated.

Since the RCS and the RHR [Residual Heat Removal] System are

drained when the RCS inventory is reduced below the reactor vessel

flange for maintenance or refueling activities, the concentrations

of chlorides and fluorides will not change. During these maintenance

or refueling activities, only controlled makeup to the RCS is

planned, and any planned or unplanned makeup to the RCS would be

detected by available level indication. Sampling for chloride and

fluoride concentrations in the RCS will be performed prior to

draining the system. Sampling of the reactor coolant for chloride

and fluoride concentrations will resume when the RCS is filled. The

chloride and fluoride concentrations will be known and will be

maintained consistent with the Technical Specification Limiting

Condition for Operation and Action Statements. Also, when the RCS

inventory is drained below the reactor vessel flange, the RCS is

vented and open to the containment building atmosphere with the

reactor coolant liquid considered oxygen saturated. Technical

Specification 3.1.F.4 allows normal and off-normal ``saturated''

oxygen concentrations when reactor coolant temperature is below 250

degrees F. Consequently, sampling the reactor coolant for oxygen

concentration under these conditions is not required and the

Technical Specification Table 4.1-2B specified sampling frequency of

five (5) times per week is not necessary since the oxygen

concentration continues to remain in compliance with the Technical

Specification limit, measures are available and action can be taken

to correct the condition prior to any deleterious effect.

Surry Technical Specifications 3.1.F.1 prohibits reactor coolant

temperature from exceeding 250 degrees F unless chloride, fluoride,

and oxygen concentrations are within specified limits. Therefore a

significant increase in the probability or consequences of an

accident previously evaluated does not exist.

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

accident from any accident previously evaluated.

The materials that are exposed to reactor coolant are corrosion

resistant. They were chosen for specific applications within the

system and for their compatibility with the reactor coolant. The

chemical composition of the reactor coolant will be maintained

within the specifications given within Technical Specification

3.1.F, Updated Final Safety Analysis Report Table 4.2-2, and

Technical Specification Table 4.1-2B. Because of the time dependent

nature of any adverse affects from chloride, fluoride, and oxygen

concentrations in excess of the Technical Specifications limits,

measures are available and can be taken to correct the condition

while the reactor is in a safe shutdown condition, prior to any

deleterious effect. No hardware modifications are involved. System

configuration and plant operations are not being changed. Therefore,

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

accident previously evaluated has not been created.

(3) Involve a significant reduction in the margin of safety.

This change does not involve a significant reduction in the

margin of safety since the chloride and fluoride concentrations are

maintained within their specified values prior to RCS drain down and

following refill. The time period during which the RCS inventory is

reduced below the reactor vessel flange and fuel is removed from the

vessel, is short and insignificant in terms of the parameters

necessary to initiate a corrosion concern. Existing Technical

Specifications Action Statements and Allowed Technical Specification

values for normal and off-normal concentrations of chlorides and

fluorides are not being changed. No hardware modifications are

involved. System configuration and plant operations are not being

changed. Surry Technical Specification 3.1.F.1 remains unaffected by

this change and continues to prohibit reactor coolant temperature

from exceeding 250 degrees F unless chloride, fluoride, and oxygen

concentrations are within specified limits.

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: Swem Library, College of

William and Mary, Williamsburg, Virginia 23185.

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

Previously Published Notices Of Consideration Of Issuance Of

Amendments To Facility Operating Licenses, Proposed No Significant

Hazards Consideration Determination, And Opportunity For A Hearing

The following notices were previously published as separate

individual notices. The notice content was the same as above. They were

published as individual notices either because time did not allow the

Commission to wait for this biweekly notice or because the action

involved exigent circumstances. They are repeated here because the

biweekly notice lists all amendments issued or proposed to be issued

[[Page 13534]]

involving no significant hazards consideration.

For details, see the individual notice in the Federal Register on

the day and page cited. This notice does not extend the notice period

of the original notice.

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

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

Docket No. 50-498, South Texas Project, Unit 1, Matagorda County,

Texas

Date of amendment request: February 29, 1996

Description of amendment request: The proposed amendment would

include the addition of Technical Specification 3.10.8 which would

allow a one-time only extension of the standby diesel generator (SDG)

allowed outage time for a cumulative 21 days on ``A'' train SDG. In

addition, it would also allow a one-time only extension of the allowed

outage time on ``A'' train essential cooling water loop for a

cumulative 7 days. This one-time only change would become effective on

April 10, 1996, and expire on May 15, 1996.Date of individual notice in

the Federal Register: March 8, 1996 (61 FR 9502)

Expiration date of individual notice: April 8, 1996

Local Public Document Room location: Wharton County Junior College,

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

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, 1995, as supplemented by letters

dated June 22, August 28, November 22, and December 19, 1995, and

January 4, January 8 (two letters), and January 23, 1996

Description of amendment request: The proposed amendment would

provide a special test exception that would allow an extension of the

standby diesel generator (SDG) allowed outage time for a cumulative 21

days on each SDG once per fuel cycle, and it would also allow an

extension of the essential cooling water (ECW) loop allowed outage time

for a cumulative 7 days on each ECW loop once per fuel cycle. These

extended allowed outage times will be used to perform required

inspections and maintenance on the SDGs and the ECW system during power

operation.

Date of individual notice in the Federal Register: February 8, 1996

(61 FR 4805)

Expiration date of individual notice: March 11, 1996

Local Public Document Room location: Wharton County Junior College,

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

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

Nuclear Generating Plant, Wright County, Minnesota

Date of amendment request: March 1, 1996 (supersedes December 11,

1995, application)

Description of amendment request: The proposed amendment would

revise Technical Specification Section 4.7, ``Surveillance Requirements

for Primary Containment Automatic Isolation Valves.'' Specifically, the

proposed amendment would revise the replacement frequency of the seat

seals for the drywell and suppression chamber purge and vent valves

from every 5 years to every six operating cycles.

Date of individual notice in the Federal Register: March 8, 1996

(61 FR 9504)

Expiration date of individual notice: April 8, 1996

Local Public Document Room location: Minneapolis Public Library,

Technology and Science Department, 300 Nicollet Mall, Minneapolis,

Minnesota 55401

Notice Of Issuance Of Amendments To Facility Operating Licenses

During the period since publication of the last biweekly notice,

the Commission has issued the following amendments. The Commission has

determined for each of these amendments that the application complies

with the standards and requirements of the Atomic Energy Act of 1954,

as amended (the Act), and the Commission's rules and regulations. The

Commission has made appropriate findings as required by the Act and the

Commission's rules and regulations in 10 CFR Chapter I, which are set

forth in the license amendment.

Notice of Consideration of Issuance of Amendment to Facility

Operating License, Proposed No Significant Hazards Consideration

Determination, and Opportunity for A Hearing in connection with these

actions was published in the Federal Register as indicated.

Unless otherwise indicated, the Commission has determined that

these amendments satisfy the criteria for categorical exclusion in

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

no environmental impact statement or environmental assessment need be

prepared for these amendments. If the Commission has prepared an

environmental assessment under the special circumstances provision in

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

it is so indicated.

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

applications for amendment, (2) the amendment, and (3) the Commission's

related letter, Safety Evaluation and/or Environmental Assessment as

indicated. All of these items are available for public inspection at

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

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

the particular facilities involved.

Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN

50-529, and STN 50-530, Palo Verde Nuclear Generating Station,

Units 1, 2, and 3, Maricopa County, Arizona

Date of application for amendments: November 7, 1995, as

supplemented by letter dated January 17, 1996.

Brief description of amendments: These amendments adopt the

improved Standard Technical Specifications (NUREG-1432) format and

content of Section 5.0, ``Design Features,'' as modified by approved

changes to the improved Standard Technical Specifications.

Date of issuance: March 6, 1996

Effective date: March 6, 1996, to be implemented within 45 days of

the date of issuance.

Amendment Nos.: Unit 1 - Amendment No. 104; Unit 2 - Amendment No.

93; Unit 3 - Amendment No. 76

Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The

amendments revised the Technical Specifications.

Date of initial notice in Federal Register: December 20, 1995 (60

FR 65673) The January 17, 1996, supplemental letter provided clarifying

information and did not change the initial no significant hazards

consideration determination. The Commission's related evaluation of the

amendments is contained in a Safety Evaluation dated March 6, 1996.No

significant hazards consideration comments received: No.

Local Public Document Room location: Phoenix Public Library, 1221

N. Central Avenue, Phoenix, Arizona 85004

[[Page 13535]]

Carolina Power & Light Company, et al., Docket No. 50-400, Shearon

Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties,

North Carolina

Date of application for amendment: February 16, 1996

Brief description of amendment: The amendment allows a one-time

extension for the performance of the trip actuating device operational

test for one of the safety injection manual initiation switches listed

in Technical Specification Table 4.3-2, Item 1a.Date of issuance: March

11, 1996

Effective date: March 11, 1996

Amendment No. 63

Facility Operating License No. NPF-63. Amendment revises the

Technical Specifications.Public comments requested as to proposed no

significant hazards consideration: Yes (61 FR 7125). That notice

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

no significant hazards consideration determination. No comments have

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

hearing by March 27, 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 amendment, finding of exigent

circumstances, and final determination of no significant hazards

consideration is contained in a Safety Evaluation dated March 11, 1996

Local Public Document Room location: Cameron Village Regional

Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.

Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455,

Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois, Docket

Nos. STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and

2, Will County, Illinois

Date of application for amendments: January 11, 1996

Brief description of amendments: The amendments revise the action

statements and allowed outage time for inoperability of one channel and

both channels of source range neutron flux instrumentation in Shutdown

Modes 3, 4, and 5.

Date of issuance: March 15, 1996

Effective date: March 15, 1996

Amendment Nos.: 80, 80, 72, and 72

Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77:

The amendments revised the Technical Specifications.

Date of initial notice in Federal Register: January 31, 1996 (61 FR

3509) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated March 15, 1996. No significant

hazards consideration comments received: No

Local Public Document Room location: For Byron, the Byron Public

Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010;

for Braidwood, the Wilmington Public Library, 201 S. Kankakee Street,

Wilmington, Illinois 60481.

Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle

County Station, Units 1 and 2, LaSalle County, Illinois

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

supplemented January 4, 1996 and February 29, 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.

Date of issuance: March 11, 1996 Effective date: Immediately, to be

implemented no later than June 30, 1996.

Amendment Nos.: 110 and 95

Facility Operating License Nos. NPF-11 and NPF-18: The amendments

revised the Technical Specifications.

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

62896) The January 4, 1996, submittal provided additional clarifying

information that did not change the initial proposed no significant

hazards consideration determination. The Commission's related

evaluation of the amendments is contained in a Safety Evaluation dated

March 11, 1996. No significant hazards consideration comments received:

No

Local Public Document Room location: Jacobs Memorial Library,

Illinois Valley Community College, Oglesby, Illinois 61348.

Detroit Edison Company, Docket No. 50-341, Fermi-2, Monroe County,

Michigan

Date of application for amendment: September 20, 1995, as

supplemented December 18 and December 22, 1995.

Brief description of amendment: The amendment allows a one-time

surveillance interval extension for certain 18-month surveillances

listed in new Technical Specification Tables 4.0.2-1 and 4.0.2-2. Date

of issuance: March 1, 1996

Effective date:

March 1, 1996, with full implementation within 90 days.

Amendment No.: 106

Facility Operating License No. NPF-43. Amendment revises the

Technical Specifications.

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

FR 58400). The December 18, 1995, letter corrected a typographical

error on one of the proposed TS pages and provided a corrected Table of

Contents page to reflect the addition of the new Tables. The December

22, 1995, letter provided additional information on the licensee's

review of historical plant drift data. This information was within the

scope of the original application and did not change the staff's

initial no significant hazards consideration determination. The

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

Safety Evaluation dated March 1, 1996.No significant hazards

consideration comments received: No.

Local Public Document Room location: Monroe County Library System,

3700 South Custer Road, Monroe, Michigan 48161

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: November 10, 1995

Brief description of amendments: The amendments revise the

Technical Specifications for containment systems to reflect the

adoption of the requirements of 10 CFR Part 50, Appendix J, Option B,

and the implementation of a performance-based containment leak-rate

testing program at the Edwin I. Hatch Nuclear Plant, Units 1 and 2.

Date of issuance: March 6, 1996

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

90 days

Amendment Nos.: Unit 1 - 200 - Unit 2 - 141

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

revised the Technical Specifications.

Date of initial notice in Federal Register: December 20, 1995 (60

FR 65679) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated March 6, 1996.No significant

hazards consideration comments received: No

Local Public Document Room location: Appling County Public Library,

301 City Hall Drive, Baxley, Georgia 31513

[[Page 13536]]

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

Nuclear Generating Station, Ocean County, New Jersey

Date of application for amendment: December 5, 1995

Brief description of amendment: The amendment revises the submittal

date for the Annual Exposure Data Report bringing Oyster Creek into

conference with 10 CFR 20.2206 and relaxes an overly restrictive

administrative requirement.

Date of Issuance: March 4, 1996

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

within 30 days.

Amendment No.: 183

Facility Operating License No. DPR-16. Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: January 22, 1996 (61 FR

1629). The Commission's related evaluation of this amendment is

contained in a Safety Evaluation dated March 4, 1996. No significant

hazards consideration comments received: No.

Local Public Document Room location: Ocean County Library,

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

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

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

Illinois

Date of application for amendment: December 14, 1995

Brief description of amendment: The amendment modifies Technical

Specification 3.4.2, ``Flow Control Valves (FCVs),'' by deleting

Surveillance Requirement (SR) 3.4.2.2, which required periodic

verification that the average rate of movement of each reactor

recirculation system FCV was limited to less than or equal to 11% per

second in the opening and closing directions. Due to a plant

modification, the requirement is not applicable.

Date of issuance: March 11, 1996

Effective date: March 11, 1996

Amendment No.: 103

Facility Operating License No. NPF-62: The amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: January 22, 1996 (61 FR

1630) The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated March 11, 1996. No significant hazards

consideration comments received: No

Local Public Document Room location: The Vespasian Warner Public

Library, 120 West Johnson Street, Clinton, Illinois 61727

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: November 10, 1995

(AEP:NRC:0896X). This application superseded a request dated June 15,

1995 (AEP:NRC:0896V).

Brief description of amendments: The amendments change the 18-month

emergency diesel generator surveillance test from a 24-hour run to an

8-hour run and add voltage and frequency measurement and power factor

monitoring.

Date of issuance: March 11, 1996

Effective date: March 11, 1996, with full implementation within 45

days

Amendment Nos.: Unit 1 - 207, Unit 2 - 191

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

revised the Technical Specifications.

Date of initial notice in Federal Register: December 20, 1995 (60

FR 65682) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated March 11, 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.

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: June 20, 1995, as supplemented

December 19, 1995.

Brief description of amendments: The amendments relocate the fire

protection program elements from the Technical Specifications and

incorporate, by reference, the NRC-approved Fire Protection Program and

major commitments, including the fire hazards analysis, into the

Updated Final Safety Analysis Report. In addition, the amendments

revise the operating licenses to include the NRC's standard fire

protection license condition.

Date of issuance: March 11, 1996

Effective date: March 11, 1996, with full implementation within 180

days

Amendment Nos.: Unit 1 - 208, Unit 2 - 192

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

revised the Technical Specifications and the operating licenses.

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

FR 47620). The December 19, 1995, supplement clarified the license

conditions by providing specific approval dates for previous fire

protection safety evaluations. 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 March 11, 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, et al., Docket No. 50-423,

MillstoneNuclear Power Station, Unit No. 3, New London County,

Connecticut

Date of application for amendment: June 29, 1995

Brief description of amendment: The amendment revises the Technical

Specifications to extend the surveillance schedule from 18 months to

each refueling interval (nominally 24 months) for specifications

4.6.4.2, 4.7.1.2.1.c, 4.7.3.b, 4.7.4.b,and 4.7.10.e. It also deletes

specification 4.6.4.2.a and the phrase ``during shutdown'' from these

specifications.Date of issuance: March 4, 1996

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

within 90 days.

Amendment No.: 127

Facility Operating License No. NPF-49. Amendment revised the

Technical Specifications.

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

FR 58402) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated March 4, 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.

Pacific Gas and Electric Company, Docket No. 50-275, Diablo Canyon

Nuclear Power Plant, Unit No. 1, San Luis Obispo County, California

Date of application for amendment: January 18, 1996

Brief description of amendment: The amendment revises the combined

Technical Specifications (TS) for the Diablo Canyon Nuclear Power

Plant, Unit No. 1. TS 3.8.1.1, ``Electrical Power Systems - A.C.

Sources - Operating,'' is revised to allow operation of Unit 1 in Mode

3 (Hot Standby) during installation of a replacement non-vital

auxiliary transformer 11, for a one time

[[Page 13537]]

extension of up to 48 hours beyond the 72 hours allowed by TS 3.8.1.1,

Action Statement (a).

Date of issuance: March 8, 1996

Effective date: March 8, 1996

Amendment No.: Unit 1 - Amendment No. 111

Facility Operating License No. DPR-80: The amendment revises the

Technical Specifications.

Date of initial notice in Federal Register: February 1, 1996 (61 FR

3737) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated March 8, 1996.No significant

hazards consideration comments received: No.

Local Public Document Room location: California Polytechnic State

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

Department, San Luis Obispo, California 93407

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 application for amendments: December 27, 1995

Brief description of amendments: The amendments revised 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; their associated Bases; and adds

Specification 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, rather than paraphrasing the

requirements of the regulation. These changes will support the

implementation of the performance based testing of Option B to Appendix

J, for Type A, B, and C containment leakage rate testing and the

appropriate rescheduling of testing.

Date of issuance: March 1, 1996 Effective date: March 1, 1996

Amendment Nos.: Unit 1 - Amendment No. 110; Unit 2 - Amendment No.

109

Facility Operating License Nos. DPR-80 and DPR-82: The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: January 31, 1996 (61 FR

3502) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated March 1, 1996.No significant

hazards consideration comments received: No.

Local Public Document Room location: California Polytechnic State

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

Department, San Luis Obispo, California 93407

Public Service Electric & Gas Company, Docket No. 50-354, Hope

Creek Generating Station, Salem County, New Jersey

Date of application for amendment: April 13, 1994, as supplemented

December 6, 1995

Brief description of amendment: The proposed changes revise the

Quality Assurance audit frequencies in the Hope Creek Technical

Specifications. These revisions will permit an audit frequency based on

performance and transfer subsequent control over the audit program to

the Updated Final Safety Analysis Report.

Date of issuance: March 11, 1996

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

60 days from the date of issuance.

Amendment No.: 95

Facility Operating License No. NPF-57: This amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: June 8, 1994 (59 FR

29633) The December 6, 1995, letter provided clarifying information

that did not change the initial proposed no significant hazards

consideration determination nor the original Federal Register

notice.The Commission's related evaluation of the amendment is

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

hazards consideration comments received: No

Local Public Document Room location: Pennsville Public Library, 190

S. Broadway, Pennsville, New Jersey 08070

Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-

311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem

County, New Jersey

Date of application for amendments: April 13, 1994, as supplemented

December 6, 1995.

Brief description of amendments: The proposed changes revise the

Quality Assurance audit frequencies in the Salem Unit Nos. 1 and 2

Technical Specifications. These revisions will permit an audit

frequency based on performance and transfer subsequent control over the

audit program to the Updated Final Safety Analysis Report.

Date of issuance: March 11, 1996

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

60 days from the date of issuance.

Amendment Nos. 181 and 162

Facility Operating License Nos. DPR-70 and DPR-75. The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: June 8, 1994 (59 FR

29633) The December 6, 1995, letter provided clarifying information

that did not change the initial proposed no significant hazards

consideration determination nor the original Federal Register

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

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

hazards consideration comments received: No

Local Public Document Room location: Salem Free Public Library, 112

West Broadway, Salem, New Jersey 08079

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

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

Date of application for amendments: December 8, 1995 (TS 93-09)

Brief description of amendments: The amendments revise the

setpoints and time delays for the auxiliary feedwater loss-of-power and

the 6.9-kilovolt shutdown board loss-of-voltage and degraded voltage

instruments.

Date of issuance: March 1, 1996

Effective date: March 1, 1996

Amendment Nos.: 219 and 209

Facility Operating License Nos. DPR-77 and DPR-79: Amendments

revise the technical specifications.

Date of initial notice in Federal Register: January 3, 1996 (61 FR

181) The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated March 1, 1996.No significant hazards

consideration comments received: None

Local Public Document Room location: Chattanooga-Hamilton County

Library, 1101 Broad Street, Chattanooga, Tennessee 37402

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

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

Date of application for amendments: January 4, 1996 (TS 95-22)

Brief description of amendments: The amendments change the

surveillance test frequency specified for the functional tests of the

containment, fuel storage pool, and control room radiation monitors

from monthly to quarterly.

Date of issuance: March 4, 1996

Effective date: March 4, 1996

Amendment Nos.: 220 and 210

Facility Operating License Nos. DPR-77 and DPR-79: Amendments

revise the technical specifications.

Date of initial notice in Federal Register: January 31, 1996 (61 FR

3503)

[[Page 13538]]

The Commission's related evaluation of the amendment is contained in a

Safety Evaluation dated Macrh 4, 1996.No significant hazards

consideration comments received: None

Local Public Document Room location: Chattanooga-Hamilton County

Library, 1101 Broad Street, Chattanooga, Tennessee 37402

The Cleveland Electric Illuminating Company, Centerior Service

Company, Duquesne Light Company, Ohio Edison Company, Pennsylvania

Power Company, Toledo Edison Company, Docket No. 50-440, Perry

Nuclear Power Plant, Unit No. 1, Lake County, Ohio

Date of application for amendment: January 16, 1996, and supplement

dated March 1, 1996

Brief description of amendment: This amendment approves that part

of the request that defers the drywell bypass leakage test during the

current refueling outage. The remainder of the licensee's request is

still under NRC staff review.

Date of issuance: March 8, 1996

Effective date: March 8, 1996

Amendment No. 82

Facility Operating License No. NPF-58: This amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: February 2, 1996 (61 FR

3951) The March 1, 1996, supplemental letter was clarifying in nature

and did not affect the initital no significant hazards consideration

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

contained in a Safety Evaluation dated March 8, 1996. No significant

hazards consideration comments received: No

Local Public Document Room location: Perry Public Library, 3753

Main Street, Perry, Ohio 44081

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

Callaway County, Missouri

Date of application for amendment: December 9, 1994, as

supplemented by letters dated September 13, 1995, and February 9, 1996.

Brief description of amendment: The amendment revises Technical

Specifications (TS) 4.3.2.2, TS 4.7.1.2.1, and the Bases for TS 3/4

7.1.2 to decrease the frequency of auxiliary feedwater pump testing,

remove inconsistencies in testing requirements for the turbine-driven

auxiliary feedwater pump, and clarify performance parameters in the TS

Bases.

Date of issuance: March 11, 1996

Effective date: March 11, 1996, to be implemented within 30 days

from the date of issuance.

Amendment No.: 108

Facility Operating License No. NPF-30: The amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: February 1, 1995 (60 FR

6314). The September 13, 1995, and February 9, 1996, supplemental

letters provided additional clarifying information and did not change

the original no significant hazards consideration determination. The

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

Safety Evaluation dated March 11, 1996.No significant hazards

consideration comments received: No.

Local Public Document Room location: Callaway County Public

Library, 710 Court Street, Fulton, Missouri 65251.

Virginia Electric and Power Company, et al., Docket Nos. 50-338 and

50-339, North Anna Power Station, Units No. 1 and No. 2, Louisa

County, Virginia

Date of application for amendments: September 19, 1995

Brief description of amendments: The amendments revised the maximum

allowable power range neutron flux high setpoints for operation with

inoperable main steam safety valves.

Date of issuance: March 6, 1996

Effective date: March 6, 1996

Amendment Nos.: 199 and 180

Facility Operating License Nos. NPF-4 and NPF-7. Amendments revised

the Technical Specifications.

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

54724) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated March 6, 1996No significant

hazards consideration comments received: No.

Local Public Document Room location: The Alderman Library, Special

Collections Department, University of Virginia, Charlottesville,

Virginia 22903-2498.

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

Creek Generating Station, Coffey County, Kansas

Date of amendment request: November 22, 1995

Brief description of amendment: The amendment replaces the

Technical Specification (TS) requirements associated with the boron

dilution mitigation system (BDMS) with alarms, indicators, procedures

and controls to allow proper resolution of potential boron dilution

events.

Date of issuance: March 1, 1996

Effective date: March 1, 1996, to be implemented prior to the

startup from the eighth refueling outage.

Amendment No.: 96

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

Technical Specifications.

Date of initial notice in Federal Register: January 31, 1996 (61 FR

3503) The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated March 1, 1996.No significant hazards

consideration comments received: No. Local Public Document Room

locations: Emporia State University, William Allen White Library, 1200

Commercial Street, Emporia, Kansas 66801 and Washburn University School

of Law Library, Topeka, Kansas 66621

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

Creek Generating Station, Coffey County, Kansas

Date of amendment request: December 20, 1995, as supplemented by

letter dated February 8, 1996.

Brief description of amendment: The amendment revises the Technical

Specifications to reflect the approval of the use of 10 CFR Part 50,

Appendix J, Option B for the Wolf Creek Generating Station containment

leakage rate test program.

Date of issuance: March 1, 1996

Effective date: March 1, 1996, to be implemented prior to startup

from the eighth refueling outage.

Amendment No.: 97

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

Technical Specifications.

Date of initial notice in Federal Register: January 31, 1996 (61 FR

3504) The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated March 1, 1996.No significant hazards

consideration comments received: No.

Local Public Document Room locations: Emporia State University,

William Allen White Library, 1200 Commercial Street, Emporia, Kansas

66801 and Washburn University School of Law Library, Topeka, Kansas

66621

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

Creek Generating Station, Coffey County, Kansas

Date of amendment request: December 13, 1995

Brief description of amendment: The amendment revises the minimum

and maximum flow requirements for the centrifugal charging pumps (CCPs)

and safety injection pumps (SIPs) specified in Technical Specification

(TS) Surveillance Requirement 4.5.2.h. Specifically, the amendment (1)

decreases the minimum limits on the sum of the injection line flow

rates,

[[Page 13539]]

excluding the highest flow rate, from 346 gallons per minute (gpm) to

330 gpm for the CCPs and from 459 gpm to 450 gpm for the SIPs, and (2)

revises the maximum pump flow rate for the SIPs from 665 to 670 gpm,

but retains the CCPs maximum pump flow rate at its current value of 556

gpm.Date of issuance: March 5, 1996

Effective date: March 5, 1996, to be implemented prior to startup

from the eighth refueling outage.

Amendment No.: 98

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

Technical Specifications.

Date of initial notice in Federal Register: January 22, 1996 (61 FR

1639) The February 5, 1996, supplemental letter provided additional

clarifying information and did not change the original no significant

hazards consideration determination. The Commission's related

evaluation of the amendment is contained in a Safety Evaluation dated

March 5, 1996.No significant hazards consideration comments received:

No. Local Public Document Room locations: Emporia State University,

William Allen White Library, 1200 Commercial Street, Emporia, Kansas

66801 and Washburn University School of Law Library, Topeka, Kansas

66621

Notice Of Issuance Of Amendments To Facility Operating Licenses And

Final Determination Of No Significant Hazards Consideration And

Opportunity For A Hearing (Exigent Public Announcement Or Emergency

Circumstances)

During the period since publication of the last biweekly notice,

the Commission has issued the following amendments. The Commission has

determined for each of these amendments that the application for the

amendment complies with the standards and requirements of the Atomic

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

and regulations. The Commission has made appropriate findings as

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

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

Because of exigent or emergency circumstances associated with the

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

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

Consideration of Issuance of Amendment, Proposed No Significant Hazards

Consideration Determination, and Opportunity for a Hearing.

For exigent circumstances, the Commission has either issued a

Federal Register notice providing opportunity for public comment or has

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

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

the Commission's proposed determination of no significant hazards

consideration. The Commission has provided a reasonable opportunity for

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

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

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

transcribed as appropriate and the licensee has been informed of the

public comments.

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

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

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

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

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

significant hazards consideration determination. In such case, the

license amendment has been issued without opportunity for comment. If

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

Commission may provide an opportunity for public comment. If comments

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

been consulted by telephone whenever possible.

Under its regulations, the Commission may issue and make an

amendment immediately effective, notwithstanding the pendency before it

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

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

significant hazards consideration is involved.

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

made a final determination that the amendment involves no significant

hazards consideration. The basis for this determination is contained in

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

been issued and made effective as indicated.

Unless otherwise indicated, the Commission has determined that

these amendments satisfy the criteria for categorical exclusion in

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

no environmental impact statement or environmental assessment need be

prepared for these amendments. If the Commission has prepared an

environmental assessment under the special circumstances provision in

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

it is so indicated.

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

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

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

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

available for public inspection at the Commission's Public Document

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

the local public document room for the particular facility involved.

The Commission is also offering an opportunity for a hearing with

respect to the issuance of the amendment. By April 26, 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

[[Page 13540]]

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

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

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

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

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

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

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

the specificity requirements described above.

Not later than 15 days prior to the first prehearing conference

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

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

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

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

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

brief explanation of the bases of the contention and a concise

statement of the alleged facts or expert opinion which support the

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

contention at the hearing. The petitioner must also provide references

to those specific sources and documents of which the petitioner is

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

facts or expert opinion. Petitioner must provide sufficient information

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

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

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

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

petitioner who fails to file such a supplement which satisfies these

requirements with respect to at least one contention will not be

permitted to participate as a party.

Those permitted to intervene become parties to the proceeding,

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

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

hearing, including the opportunity to present evidence and cross-

examine witnesses. Since the Commission has made a final determination

that the amendment involves no significant hazards consideration, if a

hearing is requested, it will not stay the effectiveness of the

amendment. Any hearing held would take place while the amendment is in

effect.

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 the

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

Toledo Edison Company, Centerior Service Company, and The Cleveland

Electric Illuminating Company, Docket No. 50-346, Davis-Besse

Nuclear Power Station, Unit No. 1, Ottawa County, Ohio

Date of application for amendment: March 6, 1996

Brief description of amendment: This amendment revises TS 3/4 5.2,

ECCS SUBSYSTEMS - T avg greater than or equal to 280 deg.F by

modifying Surveillance Requirement 4.5.2.b to defer venting of the

Emergency Core Cooling System flow path which does not have manual

venting capability until the tenth refueling outage.

Date of issuance: March 7, 1996

Effective date: March 7, 1996

Amendment No: 208

Facility Operating License No. NPF-3: Amendment revised the

Technical Specifications. Public comments requested as to proposed no

significant hazards consideration: No. The Commission's related

evaluation of the amendments, finding of emergency circumstances, and

final determination of no significant hazards consideration are

contained in a Safety Evaluation dated March 7, 1996.

Local Public Document Room location: University of Toledo, William

Carlson Library, Government Documents Collection, 2801 West Bancroft

Avenue, Toledo, Ohio 43606

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

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

NRC Project Director: Gail H. Marcus

Dated at Rockville, Maryland, this 20th day of March 1996.

For the Nuclear Regulatory Commission

Steven A. Varga, Director,

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

Regulation

[Doc. 96-7259 Filed 3-26-96; 8:45 am]

BILLING CODE 7590-01-F