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

NUCLEAR REGULATORY COMMISSION

Biweekly Notice, Applications and Amendments to Facility

Operating Licenses Involving No Significant Hazards Considerations

I. Background

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

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

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

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

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

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

the Commission the authority to issue and make immediately effective

any amendment to an operating license upon a determination by the

Commission that such amendment involves no significant hazards

consideration, notwithstanding the pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all notices of amendments issued, or

proposed to be issued from February 5, 1996, through February 15, 1996.

The last biweekly notice was published on February 14, 1996 (61 FR

5809).

[[Page 7543]]

Notice of Consideration of Issuance of Amendments to Facility Operating

Licenses, Proposed No Significant Hazards Consideration Determination,

and Opportunity for a Hearing

The Commission has made a proposed determination that the following

amendment requests involve no significant hazards consideration. Under

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

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

involve a significant increase in the probability or consequences of an

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

or different kind of accident from any accident previously evaluated;

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

for this proposed determination for each amendment request is shown

below.

The Commission is seeking public comments on this proposed

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

publication of this notice will be considered in making any final

determination.

Normally, the Commission will not issue the amendment until the

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

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

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

the Commission may issue the license amendment before the expiration of

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

the amendment involves no significant hazards consideration. The final

determination will consider all public and State comments received

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

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

opportunity for a hearing after issuance. The Commission expects that

the need to take this action will occur very infrequently.

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

Directives Branch, Division of Freedom of Information and Publications

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

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

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

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

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

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

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

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

intervene is discussed below.

By March 29, 1996, the licensee may file a request for a hearing

with respect to issuance of the amendment to the subject facility

operating license and any person whose interest may be affected by this

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

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

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

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

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

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

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

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

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

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

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

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

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

Secretary or the designated Atomic Safety and Licensing Board will

issue a notice of a hearing or an appropriate order.

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

shall set forth with particularity the interest of the petitioner in

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

the proceeding. The petition should specifically explain the reasons

why intervention should be permitted with particular reference to the

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

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

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

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

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

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

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

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

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

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

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

the specificity requirements described above.

Not later than 15 days prior to the first prehearing conference

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

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

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

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

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

brief explanation of the bases of the contention and a concise

statement of the alleged facts or expert opinion which support the

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

contention at the hearing. The petitioner must also 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

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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: December 20, 1995.

Description of amendments request: The proposed amendment would

change the instrumentation setpoint for the reactor trip and main steam

isolation signal (MSIS) actuation on low steam generator pressure from

greater than or equal to 919 psia with an allowable value of greater

than or equal to 911 psia to greater than or equal to 895 psia with an

allowable value of greater than or equal to 890 psia.

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 amendment does not involve any change to the method

of operation of any plant equipment that is used to mitigate the

consequences of an accident. The proposed change only affects the

instrument setpoint for steam generator low pressure reactor trip

and MSIS actuation. The proposed setpoint meets the requirement of

ensuring a reactor trip and MSIS actuation prior to steam generator

pressure reaching the analytical limits even under worst-case

accident conditions. Thus, the proposed change does not involve a

significant increase in the probability of an accident previously

evaluated.

The proposed amendment does not alter any of the assumptions or

bounding conditions currently in the UFSAR [updated final safety

analysis report] and meets the requirement of ensuring a reactor

trip and MSIS actuation prior to steam generator pressure reaching

the analytical setpoint under worst-case accident conditions. As a

result, the proposed amendment does not involve a significant

increase in the consequences of an accident previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

The proposed change does not involve any change to the method of

operation of any plant equipment that is used to mitigate the

consequences of an accident. Accordingly, no new failure modes have

been defined for any plant system or component important to safety

nor has any new limiting failure been identified as a result of the

proposed change. The intent of the proposed change is to increase

the margin between normal operating parameters and trip setpoints.

This minimizes the possibility of unnecessary challenges to safety

systems improving the safety of operation. The method of protecting

the facility for an excess steam demand event remains unchanged and

therefore, the amendment does not create the possibility of a new or

different kind of accident from any accident previously evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The proposed change is the implementation of a setpoint value

which was derived using methodologies endorsed by Revision 2 of NRC

Regulatory Guide 1.105, ``Instrument Setpoints.'' The new setpoint

ensures that sufficient margin exists below the full load operating

value for steam pressure so as not to interfere with normal plant

operation, but still high enough to provide the required protection

(reactor trip and main steam line isolation) in the event of an

excessive steam demand event. The new setpoint ensures that safety

margins are maintained within the results of existing calculations.

The margin of safety between the analyzed trip value and the point

at which safety analysis results become unacceptable remain

unchanged since the analytical setpoints are not affected by the

amendment. The new setpoint resulted from the reduced instrument

uncertainty and will ensure that the reactor trip and MSIS actuation

on low steam generator pressure will occur before the analyzed value

and hence, 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

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.

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: January 5, 1996.

Description of amendments request: The proposed amendment would

revise paragraph 2.C.(1) of the operating licenses and Section 1.26 of

the TS for each of the three PVNGS Units to increase the authorized 100

percent reactor core power (rated thermal power) from 3800 megawatts

thermal (Mwt) to 3876 Mwt, an increase of 2 percent. The proposed

amendment would also revise TS 4.1.1.4, TS 3.1.3.4, and TS 3.2.6

(Figure 3.2-1) to lower the allowable reactor coolant system cold leg

temperature limits for each of the three PVNGS Units, and revise TS

3.4.2.1 and TS 3.4.2.2 to lower the pressurizer safety valve setpoints

for Units 1 and 3 to support the increased power 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, 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 amendment does not change the method of operation

or modify the plant configuration other than minor changes in

equipment setpoints. Thus no increase in the probability of an

accident is created by this amendment. System and programmatic

reviews have been performed on the nuclear

[[Page 7545]]

steam supply system controls, reactor coolant system mechanical, steam

generator mechanical, balance of plant systems, and fire protection,

equipment qualification, and probabilistic risk assessment programs.

The conclusion of these reviews was that operation in accordance

with the changes proposed in this amendment was acceptable and posed

no significant risk to the health and safety of the public. The

analyses supporting this amendment demonstrate that the consequences

of events using the changes specified in the amendment are within

the criteria which are the current licensing basis for the PVNGS

Units. Therefore the amendment, as proposed, does not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

The proposed amendment does not modify the configuration of the

units except for minor equipment setpoints. No equipment changes and

no new methods of plant operation are being proposed, therefore, no

new failure modes are introduced by the proposed amendment. The

setpoint changes proposed have been evaluated and shown to be

acceptable in providing their design function. The increased rated

thermal power and associated changes have been incorporated into the

safety analysis performed in support of this amendment request and

the results have been shown to be similar to those previously

obtained. No possibility of a new or different kind of accident from

any accident previously evaluated will be created as a result of the

proposed amendment.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The changes proposed were evaluated in the safety analysis

performed to justify the amendment request. Although the

consequences of some events increased slightly, the results continue

to meet the criteria which form the PVNGS licensing basis. The

programmatic and system reviews provide further assurance of the

capability of the units to continue to operate safely with the

changes proposed in this amendment. Therefore the amendment does not

involve a significant reduction in a margin of safety.

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.

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

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

Date of amendment request: January 29, 1996.

Description of amendment request: The proposed change would revise

the technical specifications (TS) table 4.1-3, item 4 to change the

frequency of main steam safety valve (MSSV) testing to that specified

in NUREG-1431, the improved ``Standard Technical Specifications,

Westinghouse Plants'' (one third of the MSSVs each refueling outage).

In addition, the licensee proposed adding the MSSV test acceptance

requirements.

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.

Neither the valves' nor the system's configuration or functions

are being altered. The valves' setpoints and their ``as-left''

range, +/-1%, will not be changed. The changes are to the testing

frequency and the ``as found'' tolerance of the MSSV setpoint.

The proposed changes in testing frequency and the higher

tolerance are in the less conservative direction, but are not

significant for several reasons. First, the new standards are based

on the American Society of Mechanical Engineers (ASME) Boiler and

Pressure Vessel Code. The new standards have been accepted by the

nuclear industry and the NRC, and are referenced in the improved

Standard Technical Specifications. Based on a discussion with the H.

B. Robinson Steam Electric Plant (HBRSEP), Unit No. 2 MSSV

manufacturer (i.e., Crosby), HBRSEP, Unit No. 2 has not experienced

more problems with the Crosby MSSVs than the nuclear industry in

general, thus, the new level of safety will be equivalent to that of

the nuclear industry. Second, if a MSSV does fail the surveillance

test, the proposed TS will require additional MSSVs to be tested.

This requirement provides assurance that testing will reveal

possible generic problems. The impact of the tolerance on the

Chapter 15 accidents was analyzed and found to be within acceptable

limits.

Since no Updated Final Safety Analysis Report (UFSAR) Chapter 15

accident analysis is significantly impacted by the proposed changes,

there would be no increase in the consequences of an accident

previously evaluated. The testing in accordance with the ASME Boiler

and Pressure Vessel Code will provide an adequate level of assurance

that the MSSVs will be able to perform their intended function;

therefore the probability of a previously evaluated accident is not

increased.

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

or different kind of accident from any accident previously

evaluated.

No new systems or equipment are involved with the proposed

changes; and the plant's configuration and operational procedures

are unaffected. Since the proposed changes do not impact the plant's

operation, it can not create a new or different kind of accident.

3. The proposed changes do not involve a significant reduction

in the margin of safety.

The change in testing frequency is in a less conservative

direction, but it is based on the ASME Code and the improved

Standard Technical Specifications. Since HBRSEP, Unit No. 2 has not

experienced a greater number of failures associated with these MSSVs

than the nuclear industry in general, the decrease in the MSSV

testing frequency will not significantly impact the margin of

safety. Also, analyses have been performed that demonstrate that the

impact of the setpoint tolerance change on the UFSAR Chapter 15

accident analysis results is not significant. Therefore, the

proposed changes do not involve a significant reduction in a margin

of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Hartsville Memorial Library,

147 West College Avenue, Hartsville, South Carolina 29550.

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

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

Raleigh, North Carolina 27602.

NRC Project Director: David B. Matthews.

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

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

Date of amendment request: January 31, 1996.

Description of amendment request: The proposed change would revise

the Technical Specifications section 4.4 to allow the use of 10 CFR

Part 50, Appendix J, Option B, Performance-Based Containment Leakage

Rate Testing. A new TS section 6.12 is proposed to describe the

containment leakage rate testing program, committing to meet 10 CFR

50.54(o) and 10 CFR Part 50, Appendix J, Option B for type A tests; and

to meet 10 CFR part 50, Appendix J, Option A, for types B and C tests.

The bases would be changed to reflect the proposed changes.

[[Page 7546]]

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

The proposed change does not involve a significant hazards

consideration for the following reasons.

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The HBRSEP [H. B. Robinson Steam Electric Plant], Unit No. 2

Type A testing history provides substantial justification for the

proposed test schedule change to one test in a 10 year period. Three

Structural Integrity Tests (SITs) and seven Integrated Leak Rate

Tests (ILRTs) have been performed with acceptable results. Previous

testing has affirmed the acceptable reliability of the containment

structure to minimize leakage as designed, and provides assurance

that its performance to continuously function as designed is not

challenged due to this test schedule extension to once in 10 years.

Therefore, this proposed change to the TS that revises the Type

A testing frequency does not involve an increase in the probability

of an accident previously evaluated.

This proposed change to revise the test schedule frequency does

not impact nor alter the design of any system, structure or

component. The limit on allowable leakage is not increased. Type A

testing provides periodic verification of the leak tight integrity

of the containment and the systems and components that penetrate the

containment structure.

NUREG-1493, ``Performance-Based Containment Leak-Test Program,''

provides the technical basis for the NRC's rulemaking to revise

containment leakage testing requirements for nuclear power reactors

in 10 CFR 50, Appendix J. Section 10.1.2 of NUREG-1493, ``Summary of

Technical Findings, Leakage-Testing Intervals,'' states the

following.

1. Reducing the frequency of Type A tests (ILRTs) from the

current three per 10 years to one per 20 years was found to lead to

an imperceptible increase in risk. The estimated increase in risk is

very small because ILRTs identify only a few potential containment

leakage paths that cannot be identified by Type B and C testing, and

the leaks found by Type A tests have been only marginally above

existing requirements.

2. Given the insensitivity of risk to containment leakage rate

and the small fraction of leakage paths detected solely by Type A

testing, increasing the interval between ILRTs is possible with

minimal impact on public risk.

Therefore, based on the previous Type A test results, the

proposed change does not involve a significant increase in the

consequences of an accident previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

The proposed change only incorporates the performance based

testing approach authorized in 10 CFR 50, Appendix J, Option B, and

is justified based on previous plant-specific Type A test results.

Plant structures, systems, and components will not be operated in a

different manner as a result of this proposed change and no physical

modifications to equipment are involved. The interval extensions

allowed by Option B of 10 CFR 50, Appendix J, do not have the

potential for creating the possibility of new or different type of

accidents from those previously evaluated.

3. The proposed change does not involve a significant reduction

in the margin of safety.

The proposed change does not change the allowable leak rate from

the containment, it only allows an extension of the interval between

the performance of Type A leak rate testing. NUREG-1493, which

provides the technical basis for the NRC's rulemaking to revise

containment leakage testing requirements for nuclear power reactors

in 10 CFR 50, Appendix J. Section 10.1.2 of NUREG-1493, ``Summary of

Technical Findings, Leakage-Testing Intervals,'' states the

following.

``1. Reducing the frequency of Type A tests (ILRTs) from the

current three per 10 years to one per 20 years was found to lead to

an imperceptible increase in risk. The estimated increase in risk is

very small because ILRTs identify only a few potential containment

leakage paths that cannot be identified by Type B and C testing, and

the leaks found by Type A tests have been only marginally above

existing requirements.

2. Given the insensitivity of risk to containment leakage rate

and the small fraction of leakage paths detected solely by Type A

testing, increasing the interval between ILRTs is possible with

minimal impact on public risk.''

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room Location: Hartsville Memorial Library,

147 West College Avenue, Hartsville, South Carolina 29550.

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

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

Raleigh, North Carolina 27602.

NRC Project Director: David B. Matthews.

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

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

Date of amendment request: January 29, 1996.

Description of amendment request: The proposed change would revise

the technical specifications (TS) to: (1) add TS 4.6.1.5 to provide

criteria for 24-hour full-load testing of the emergency diesel

generators (EDGs) to be performed during each refueling outage; (2)

revise TS 4.6.1.2 to allow testing of the EDG protective bypasses

listed in TS 3.7.1.d to be done independent of the safety injection or

loss of offsite power testing; and (3) revise TS 4.6.1.3 to include the

EDG protective bypass inspection and a requirement to inspect the EDGs

at least once every refueling outage.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

The proposed changes do not involve a significant hazards

consideration for the following reasons.

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

the probability or consequences of an accident previously evaluated.

The proposed changes do not involve a significant increase in

the probability of an accident previously evaluated. The proposed

changes require additional testing of the EDGs and will change the

requirement for when the protective bypasses are tested. The

function of the EDGs remains unchanged. Since the additional testing

involves the EDGs, which are required to mitigate an accident and

are not involved in the initiation of an accident, the proposed

changes will not increase the probability of an accident.

The proposed changes do not involve a significant increase in

the consequences of an accident previously evaluated. The proposed

changes require additional testing to verify the reliability of the

EDGs and to show the EDGs can withstand maximum accident loading

conditions. The proposed changes will also require the testing of

the EDG protective bypasses to be accomplished during EDG outages

and not during the SI/LOOP testing during a refueling outage. The

ability of the EDGs to perform their accident mitigation function

remains unchanged. Therefore, the proposed changes will not increase

the consequences of an accident.

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 do not create the possibility of a new kind

of accident from any previously evaluated. The proposed changes are

an enhancement to the EDG testing requirements. The most significant

change will require additional testing of the EDGs to demonstrate

adequate reliability and to determine if the EDGs can withstand

maximum accident loading conditions. The remaining changes will

augment the TS to allow on-line EDG inspections and testing. Since

the function of the EDGs remains unchanged and they are not the

initiator of an accident, the proposed changes will not

[[Page 7547]]

create the possibility of a new kind of accident from any previously

evaluated.

The proposed changes do not create the possibility of a

different kind of accident from any accident previously evaluated.

The proposed changes require additional testing of the EDGs (i.e.,

the 24 hour full-load test) and revise the requirement for testing

the EDG protective bypasses during the SI/LOOP testing. The

additional testing of the EDGs will demonstrate sufficient

reliability and determine if the EDGs can withstand maximum accident

loading conditions. The EDG protective bypasses will be statically

tested during an EDG outage thus preventing possible damage to

equipment from a transient if the protective bypass fails. The

function of the EDGs remains unchanged by these proposed changes.

Since the EDGs are required to mitigate an accident and are not the

initiators of an accident, the proposed changes will not create a

different kind of accident from any kind of accident previously

evaluated.

3. The proposed changes do not involve a significant reduction

in the margin of safety.

The proposed changes do not reduce the margin of safety as

defined in the TS. The proposed changes are being submitted as an

enhancement to the testing requirements outlined in the TS. The

changes include additional testing, revising the requirement to test

the engine protective bypasses during the SI/LOOP testing and

clarification of the periodicity of inspecting the EDGs. The

additional testing demonstrates increased reliability and determines

that the EDGs can cope with maximum accident loading. The remaining

proposed changes provide clarification as to when the EDG

inspections and testing are required. The ability of the EDGs to

perform their function will not be reduced. Therefore, the margin of

safety will not be reduced by the proposed changes.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Hartsville Memorial Library,

147 West College Avenue, Hartsville, South Carolina 29550.

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

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

Raleigh, North Carolina 27602.

NRC Project Director: David B. Matthews.

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 amendment request: December 6, 1995.

Description of amendment request: The proposed amendment would

change the technical specifications of these plants to incorporate 10

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

for Water-Cooled Power Reactors'', 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:

ComEd proposes to revise Byron Nuclear Power Station, Units 1

and 2 (Byron), and Braidwood Nuclear Power Station, Units 1 and 2

(Braidwood) Technical Specification (TS) Section 3/4.6.1, ``Primary

Containment,'' and the associated Bases to reflect recent changes to

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

Testing for Water-Cooled Power Reactors.'' The proposed revisions

include:

1. Adding TS Definitions 1.15.a for the maximum allowable

primary containment leakage rate (La) and 1.20.a for the

maximum calculated primary containment pressure (Pa). The

redundant definitions throughout TS Section 3/4.6.1 are deleted,

2. Adding numerous statements throughout TS Section 3/4.6.1 that

leak rate testing is performed in accordance with Regulatory Guide

(RG) 1.163, Revision 0, ``Performance-Based Containment Leak-Test

Program,'' and its referenced documents,

3. Deleting TS requirements that are taken verbatim from 10 CFR

50, Appendix J. The specific requirements will be placed in the

containment leakage rate test program in accordance with RG 1.163,

and its referenced documents, and

4. Clarifying Technical Specification Surveillance Requirement

(TSSR) 4.6.1.1.a for consistency with NUREG-1431, Revision 1,

``Standard Technical Specifications for Westinghouse Plants.''

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

the probability or consequences of an accident previously evaluated.

10 CFR 50, Appendix J, has been amended to include provisions

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

Option B allows plants with satisfactory Integrated Leak Rate

Testing (ILRT) performance history to reduce the Type A testing

frequency from three tests in ten years to one test in ten years.

For Type B and Type C tests, Option B allows plants to reduce

testing frequency based on the leak rate test history of each

component. In addition, Option B establishes controls to ensure

continued satisfactory performance of the affected penetrations

during the extended testing interval. To be consistent with the

requirements of Option B to 10 CFR 50, Appendix J, ComEd proposes to

include appropriate changes to the TSs that incorporate the

necessary revisions.

Some of the proposed changes represent minor curtailments to

current TS requirements, but are based on the requirements specified

by Option B to 10 CFR 50, Appendix J. Any such changes are

consistent with the current plant safety analyses and have been

determined to represent sufficient requirements for the assurance of

the reliability of equipment assumed to operate in the safety

analyses, or provide continued assurance that specified parameters

associated with containment integrity remain within their acceptance

limits. The other proposed changes maintain consistency with those

requirements specified by Option B to 10 CFR 50, Appendix J and are

consistent with the current plant safety analyses. Implementation of

these changes will provide continued assurance that specified

parameters associated with containment integrity will remain within

their acceptance limits, and as such, will not significantly

increase the probability or consequences of a previously evaluated

accident.

The associated systems affecting the leak rate integrity are not

assumed in any safety analyses to initiate any accident sequence;

therefore, the probability of occurrence of any accident previously

evaluated is not increased. In addition, the proposed changes to the

limiting conditions for operation and surveillance requirements for

such systems are consistent with the current 10 CFR 50, Appendix J,

requirements. The proposed changes maintain an equivalent level of

reliability and availability for all affected systems.

Maintaining allowable leakage within the analyzed limit assumed

for the accident analyses does not adversely affect either the

onsite or offsite dose consequences. Furthermore, containment

leakage is not an accident initiator. As such, there is no adverse

impact on the probability of accident initiators. Thus, there is no

significant increase in the probability or occurrence of any

previously analyzed accident, or increase the consequences of any

previously analyzed accident.

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

or different kind of accident from any accident previously

evaluated.

Option B of 10 CFR 50, Appendix J, specifies, in part, that a

Type A test may be conducted at a periodic interval based on the

performance of the overall containment system. Type A tests measure

both the containment system overall integrated leakage rate at the

containment pressure boundary and system alignments assumed during a

large break loss-of-coolant accident (LOCA), and demonstrate the

capability of the primary containment to withstand an internal

pressure load. The acceptable leakage rates are specified in the

TSs. For Type B and C tests, intervals are proposed for

establishment based on the performance history of each component.

Acceptance criteria for each component are based upon demonstration

that the leakage rates at design basis pressure conditions for

applicable penetrations are within the limits specified in the TSs.

The proposed changes reflect the requirements specified in the

amended 10 CFR 50, Appendix J, and are consistent with the current

plant safety analyses. Some minor curtailments of current TS

requirements are

[[Page 7548]]

based on generic guidance or similarly approved provisions for other

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

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

components at the plant; however, these are in accordance with the

current plant safety analyses and provide for appropriate testing or

surveillance that is consistent with Option B to 10 CFR 50, Appendix

J. The proposed changes will not introduce new failure mechanisms

beyond those already considered in the current plant safety

analyses.

No new modes of operation are introduced by the proposed

changes. Surveillance requirements are changed to reflect

corresponding changes associated with Option B to 10 CFR 50,

Appendix J. The proposed changes maintain at least the present level

of operability of any such system that affects plant containment

integrity. Therefore, the proposed changes do not create the

possibility of a new or different kind of accident from any

previously evaluated. The associated systems that affect plant leak

rate integrity related to the proposed amendment are not assumed to

initiate any accident sequence. In addition, the proposed

surveillance requirements for any such affected systems are

consistent with the current requirements specified within the TSs

and are consistent with the requirements of Option B to 10 CFR 50,

Appendix J. The proposed surveillance requirements maintain an

equivalent level of reliability and availability of all affected

systems and, therefore, do not affect the consequences of any

previously evaluated accident. As such, the probability of systems

associated with leak rate test integrity failing to perform their

intended function is unaffected by the proposed limiting conditions

for operation and surveillance requirements.

C. The proposed changes do not involve a significant reduction

in a margin of safety.

The provisions specified in Option B to 10 CFR 50 Appendix J,

allows changes to Type A, B, and C test intervals based upon the

performance of past leak rate tests. The effect of extending

containment leak rate test intervals is a corresponding increase in

the likelihood of containment leakage. The degree to which intervals

can be extended has a direct impact on the potential effect on

existing plant safety margins and the public health and safety that

can occur due to an increased likelihood of containment leakage.

Changing Type A, B, and C test intervals from those currently

provided in the TS to those provided for in 10 CFR 50, Appendix J,

Option B, slightly increases the risk associated with Type A, B, and

C specific accident sequences. Historical data suggest that

increasing the Type C test interval can slightly increase the

associated risk; however, this is compensated by the corresponding

risk reduction benefits associated with reduction in component

cycling, stress, and wear associated with increased test intervals.

In addition, when considering the total integrated risk, which

includes all analyzed accident sequences, the additional risk

associated with increasing test intervals is negligible.

The proposed changes are consistent with those provisions

specified in Option B of 10 CFR 50, Appendix J, and are consistent

with current plant safety analyses. In addition, these proposed

changes do not involve revisions to the design of the plant. As

such, the proposed individual changes will maintain the same level

of reliability of the equipment associated with containment

integrity, assumed to operate in the plant safety analysis, or

provide continued assurance that specified parameters affecting

plant leak rate integrity, will remain within their acceptance

limits. Therefore, the proposed changes provide continued assurance

of the leakage integrity of the containment without adversely

affecting the public health and safety and, as such, will not

significantly reduce existing plant safety margins.

The proposed changes are based on United States Nuclear

Regulatory Commission (USNRC) accepted provisions and maintain

necessary levels of system or component reliability affecting plant

containment integrity. The performance-based approach to leakage

rate testing concludes that the impact on public health and safety

due to revised testing intervals is negligible. The proposed changes

will not reduce the availability of systems associated with

containment integrity when they are required to mitigate accident

conditions; therefore, the proposed changes do 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

requested amendments involve no significant hazards consideration.

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.

Attorney for licensee: Michael I. Miller, Esquire; Sidley and

Austin, One First National Plaza, Chicago, Illinois 60603.

NRC Project Director: Robert A. Capra.

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

Plant, Middlesex County, Connecticut

Date of amendment requests: December 4, 19, 19, 20, 20, and 20,

1995.

Description of amendment request: Each proposed amendment would

change the surveillance requirement frequency from the current once per

18-month interval to once per 24-month which is the proposed length of

a Haddam Neck refueling cycle. The changes pertain to the following

equipment:

December 4, 1995, Reactivity control systems flow paths, rod

position indication system, and Rod drop time.

December 19, 1995, Containment Air Recirculation System.

December 19, 1995, Main steam line (MSL) Code Safety Valves self

actuation, auxiliary feedwater system, service water system, snubber

testing, feedwater isolation valve actuation, and primary auxiliary

building cleanup system.

December 20, 1995, reactor coolant system (RCS) interlock,

containment sump, High Pressure Safety Injection Pump and Low Pressure

Safety Injection autostart and alignment, containment spray, and PH

control.

December 20, 1995, Trip actuating devices and channel trips,

reactor trip system, reactor trip system instrumentation, and accident

monitoring instrumentation.

December 20, 1995, RCS flow indicators, Loop stop valve interlock,

Pressurizer code safety valves, Emergency power supply for the

pressurizer heaters, Containment main sump and volume control tank

(VCT) level monitoring system, RCS pressure boundary valves, Low

temperature overpressure protection (LTOP) system, and RCS vent path.

Basis for proposed no significant hazards consideration

determination: The Commission has made a proposed determination that

the amendment request involves 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. As required by 10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no significant hazards

consideration. The NRC staff has reviewed the licensee's analysis

against the standards of 10 CFR 50.92(c). The NRC staff's review is

presented below:

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

probability or consequences of an accident previously evaluated.

The proposed changes to surveillance requirements of the Haddam

Neck Plant Technical Specifications extend the frequency for checking

the operability of the affected components/equipment. The proposal

would extend the frequency from at least once per 18 months to at least

once each refueling interval (i.e., nominal 24-months).

[[Page 7549]]

Changing the frequency of surveillance requirements from at least

once per 18 months to at least once each refueling interval does not

change the basis for the frequency. The frequency was chosen because of

the need to perform this verification under the conditions that apply

during a plant outage, and to avoid the potential of an unplanned

transient if the surveillance were conducted with the plant at power.

The proposed changes do not alter the intent or method by which the

surveillance are conducted, do not involve any physical changes to the

plant, do not alter the way any structure, system, or component

functions, and do not modify the manner in which the plant is operated.

As such, the proposed changes in the frequency of surveillance

requirements will not degrade the ability of the equipment/components

to perform its safety function.

Additional assurance of the operability of the components/equipment

is provided by additional surveillance requirements (e.g., monthly or

quarterly surveillance).

Equipment performance over the last four operating cycles was

evaluated to determine the impact of extending the frequency of

surveillance requirements. This evaluation included a review of

surveillance results, preventive maintenance records, and the frequency

and type of corrective maintenance. It concluded that there is no

indication that the proposed extension could cause deterioration in the

condition or performance of any of the subject components.

In addition to the substantive changes, there are format changes

which are merely editorial and because format changes produce no

physical change they do not influence the probability or consequences

of accidents.

Since the proposed changes only affect the surveillance frequency

for safety systems that are used to mitigate accidents, the changes

cannot affect the probability of any previously analyzed accident.

While the proposed changes can lengthen the intervals between

surveillance, the increases in intervals has been evaluated and it is

concluded that there is no significant impact on the reliability or

availability of the safety system and consequently, there is no impact

on the consequences on any analyzed accident.

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

kind of accident from any accident previously evaluated.

The proposed changes to surveillance requirements of the Haddam

Neck Plant Technical Specifications extend the frequency for verifying

the operability of the affected components/equipment. The proposal

would extend the frequency from at least once per 18 months to at least

once each refueling interval (nominal 24 months).

Changing the frequency of surveillance requirements from at least

once per 18 months to at least once each refueling interval does not

change the basis for the frequency. The frequency was chosen because of

the need to perform this verification under the conditions that apply

during a plant outage, and to avoid the potential of an unplanned

transient if the surveillance were conducted with the plant at power.

In addition to the substantive changes, there are format changes

which are merely editorial and because format changes produce no

physical change they do not influence the probability of new or

different types of accidents.

The proposed changes do not alter the intent or method by which the

surveillance are conducted, do not involve any physical changes to the

plant, do not alter the way any structure, system, or component

functions, and do not modify the manner in which the plant is operated.

As such, the proposed changes cannot create the possibility of a new or

different kind of accident from any previously evaluated.

3. The changes do not involve a significant reduction in a margin

of safety.

The proposed changes to surveillance requirements of the Haddam

Neck Plant Technical Specifications extend the frequency for verifying

the operability of the components/equipment. The proposal would extend

the frequency from at least once per 18-months to at least once each

refueling interval (24-months).

In addition to the substantive changes, there are format changes

which are merely editorial and because format changes produce no

physical change they do not influence the margin of safety.

The proposed changes to surveillance frequency are still consistent

with the basis for the frequency, and the intent or method of

performing the surveillance is unchanged. Further, the current

inservice testing requirements and the previous history of reliability

of the system provides assurance that the changes will not affect the

reliability of the auxiliary feedwater system. Thus, it is concluded

that there is no impact on the margin of safety.

Based on this review, it appears that the three standards of 10 CFR

50.92(c) are satisfied. Therefore, the NRC staff proposes to determine

that the amendment request involves no significant hazards

consideration.

Local Public Document Room location: Russell Library, 123 Broad

Street, Middletown, CT 06457.

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

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

CT 06141-0270.

NRC Project Director: Phillip F. McKee.

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

Plant, Middlesex County and Northeast Nuclear Energy Company, et al.,

Docket Nos. 50-245, 50-336, and 50-423, Millstone Nuclear Power

Station, Units 1, 2, and 3, New London County, Connecticut

Date of amendment request: June 6, 1995 (published August 2, 1995,

60 FR 39434), as supplemented November 22, 1995.

Description of amendment request: The proposed amendments will

modify the size of the Plant Operations Review Committee (PORC) which

will collectively have the experience and expertise in various areas of

plant operation, and will clarify the composition of the Site

Operations Review Committee (SORC).

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 (SHC), which is presented below:

. . . These proposed changes do not involve an SHC because the

changes do not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The plant operations review committee (PORC) is an oversight

group and helps to ensure that the units are operated in a safe

manner. To accomplish this the PORCs provide their recommendations

on the safety related activities to the Vice President--Haddam Neck

Plant for Haddam Neck and to the respective Nuclear Unit Directors

for Millstone. Each Millstone Unit has its own PORC. It is proposed

that the members of the Millstone PORCs be selected by the

respective Nuclear Unit Director based on their knowledge and

expertise in specific key plant functions. The Millstone Station has

one site operations review committee (SORC). The SORC is also an

oversight group whose charter is to advise the Senior Vice

President--Millstone Station on all matters related to nuclear

safety at the Millstone site. The Haddam Neck Plant, being a single

unit site, has one PORC, which advises the Vice President--Haddam

Neck Plant. The members of the Haddam Neck Plant PORC will be

selected by the Vice President--Haddam Neck Plant based on their

knowledge and expertise in specific key

[[Page 7550]]

plant functions. The PORC and SORC add to the defense-in-depth concept

provided by the design, operation, maintenance, and quality

oversight by promoting excellence through the conduct of their

affairs and by maintaining a diligent watch over their

responsibilities.

These administrative changes will revise the composition section

of the technical specifications for the PORC members. Millstone Unit

individuals will be appointed by the Nuclear Unit Directors if the

individual meets one or more of the following areas of expertise:

Plant Operations, Engineering, Reactor Engineering, Maintenance,

Instrumentation and Controls, Health Physics, Chemistry, Work

Planning and Control, and Quality Services. The Haddam Neck Plant,

due to its broader scope of review also include an individual

experienced in Security and specific expertise in Electrical

Maintenance and Mechanical Maintenance. The individuals who will

serve on PORC shall continue to meet the criteria of ANSI N18.1-1971

along with the qualification requirements contained in the technical

specifications. This approach is consistent with the standard

technical specifications and NUREG 0800, Section 13.4. For SORC at

the Millstone Station, the method of identifying who shall serve as

Vice Chairperson has been modified for clarity. Finally, the

individual who shall represent Quality and Assessment Services shall

be modified to allow a qualified member of Quality and Assessment

Services to serve on SORC.

The remaining portions of the technical specifications related

to PORC and SORC are not being revised.

These modifications broaden the unit committee participation and

reflect current organizational positions and will not increase the

probability of occurrence or the consequences of an accident

previously evaluated.

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

from any accident previously evaluated.

The proposed administrative enhancements to the composition of

the PORC and Millstone Station SORC will not affect the way in which

the units are physically operated. These administrative changes to

PORC and SORC continue to meet the guidelines of ANSI N18.7-1976.

The modifications to PORC and SORC continue to allow these groups to

provide a thorough review of activities at the units.

The proposed modification does not impact any initiating events,

and therefore, cannot create the possibility of any new or different

kind of accident from any accident previously evaluated.

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

These proposed administrative changes will not impact the margin

of safety provided by PORC and SORC. The PORC and SORC will continue

to be staffed by qualified individuals experienced in the operation

of the plants. These administrative changes will modify how the

composition of the PORC and SORC members are presented in the

technical specifications, but will not adversely impact their

ability to review and comment on operations at the units.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Russell Library, 123 Broad

Street Middletown, Connecticut 06457, for the Haddam Neck Plant, and

the Learning Resources Center, Three Rivers Community-Technical

College, 574 New London Turnpike, Norwich, CT 06360, for Millstone 1,

2, and 3.

Attorney for licensee: Ms. L. M. Cuoco, Senior Nuclear Counsel,

Northeast Utilities Service Company, Post Office Box 270, Hartford, CT

06141-0270.

NRC Project Director: Phillip F. McKee.

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

Michigan

Date of amendment request: November 22, 1995 (NRC-95-0124).

Description of amendment request: The proposed amendment would

modify the allowed out-of-service time for one onsite alternating

current (ac) electrical power division from 72 hours to 7 days. The

proposed amendment would also eliminate accelerated testing and special

reports as a result of diesel generator surveillance failures in

accordance with Generic Letter 94-01, ``Removal of Accelerated Testing

and Special Reporting Requirements for Emergency Diesel Generators,''

dated May 31, 1994.

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. Changing the out-of-

service time, surveillance frequency and reporting requirements for

emergency diesel generators (EDGs) will not affect the initiation of

an accident, since EDGs are not associated with any accident

initiation mechanism. The proposed changes will not impact the plant

design or method of EDG operation. The increased out-of-service time

has been evaluated to have only a small impact on plant risk.

Performing the EDG inspections during plant operations will decrease

plant risk during plant outages. Deleting the accelerated testing

provisions will not affect the consequences of an accident since the

implementation of a maintenance and monitoring program for EDGs

consistent with the provisions of the maintenance rule will assure

EDG performance as discussed in Generic Letter 94-01. Deleting

reporting requirements has no impact on consequences of an accident

since reporting has no accident effect. Based on the amount of

electrical system redundancy, the small increase in plant risk

during operations and the decrease in plant risk during outages,

this change will not result in a significant increase in the

probability or consequences of an accident.

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

or different accident from any previously evaluated. The proposed

changes do not modify the plant design or method of diesel

operation. Therefore, no new accident initiator is introduced, nor

is a new type of failure created. For these reasons, no new or

different type of accident is created by these changes.

3. The proposed changes do not involve a significant reduction

in a margin of safety. Since implementation of a maintenance program

for the EDGs consistent with the Maintenance Rule will ensure that

high EDG performance standards are maintained, the accelerated

testing schedule is not needed to maintain the margin of safety.

Deleting reporting requirements has no impact on safety or margin of

safety. Increasing the allowed out-of-service time for one division

of onsite AC power will slightly increase EDG unavailability during

plant operation. However, this change does not impact the redundancy

of offsite power supplies, the allowed out-of-service time if both

divisions are inoperable, or the ability to cope with a station

blackout event. This request also does not change the Action

statement for AC electrical power systems required when the plant is

shutdown. The increase in core damage frequency was assessed to be

small by an evaluation using the plant PSA [probabilistic safety

assessment] for the operating condition. Enabling the diesel

generator inspections to be performed on-line will improve safety

while shutdown by reducing EDG out-of-service time during outages.

For these reasons, the proposed changes do 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: Monroe County Library System,

3700 South Custer Road, Monroe, Michigan 48161.

Attorney for licensee: John Flynn, Esq., Detroit Edison Company,

2000 Second Avenue, Detroit, Michigan 48226.

NRC Project Director: John N. Hannon.

[[Page 7551]]

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

Michigan

Date of amendment request: December 21, 1995 (NRC-95-0133).

Description of amendment request: The proposed amendment would

implement Option B of the recently revised 10 CFR Part 50 Appendix J in

a manner consistent with Regulatory Guide 1.163, ``Performance-Based

Containment Leak Test Program,'' and industry guidance contained in NEI

94-01, Revision 0, ``Industry Guideline for Implementing Performance-

Based Option of 10 CFR 50, Appendix J,'' with the exception of

previously approved exemptions which the licensee wishes to remain in

effect. The previously approved exemptions are for reduced pressure for

testing MSIVs [main steam isolation valves] and testing of LPCI [low

pressure coolant injection] isolation valves in accordance with

Technical Specification (TS) 4.4.3.2.2.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. This request does not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The proposed change implements the new Option B of 10 CFR Part

50 Appendix J on performance-based containment leakage testing. The

proposed change does not involve a change to the plant design or

operation. As a result, the proposed change does not affect any

parameters or conditions that contribute to the initiation of any

accidents previously evaluated. Thus, the proposed change cannot

increase the probability of any accident previously evaluated.

The proposed change potentially affects the leak-tight integrity

of the containment structure designed to mitigate the consequences

of a loss-of-coolant accident (LOCA). The function of the

containment is to maintain functional integrity during and following

the peak transient pressures and temperatures which result from any

loss-of-coolant accident (LOCA). The containment is designed to

limit fission product leakage following the design basis LOCA.

Because the proposed change does not alter the plant design, only

the frequency of measuring Type A, B, and C leakage, the proposed

change does not directly result in an increase in containment

leakage. However, decreasing the test frequency can increase the

probability that an increase in containment leakage could go

undetected for an extended period of time. Test intervals will be

established based on the performance history of components being

tested. The risk resulting from the proposed changes is

characterized as follows, based primarily on the results contained

in NUREG-1493 [''Performance-Based Containment Leakage Test

Program''], the principal Technical Support Document used by the NRC

as the basis for the Appendix J final rule (Reference 9 [of

application]) and the NRC's Final Regulatory Impact Analysis as

contained in SECY-95-181 [Final Regulatory Impact Analysis,

Performance-Based Containment Leakage-Test Program (Attachment 2 to

NRC Rulemaking Issue Affirmation, SECY-95-181 dated July 17, 1995,

Final Amendment to 10 CFR 50, Appendix J, ``Containment Leakage

Testing,'' to Adopt Performance-Oriented and Risk-Based Approaches)]

(Reference 10 [of application]):

Type A Testing

NUREG-1493 found that the effect of containment leakage on

overall accident risk is minimal since risk is dominated by accident

sequences that result in failure or bypass of the containment.

Industry wide, ILRTs [integrated leak rate tests] have only

found a small fraction of the leaks that exceed current acceptance

criteria. Only three percent of all leaks are detectable only by

ILRTs, and therefore, by extending the Type A testing intervals,

only three percent of all leaks have a potential for remaining

undetected for longer periods of time. In addition, when leakage has

been detected by ILRTs, the leakage rate has been only marginally

above existing requirements. The Fermi Type A testing confirms the

industry-wide experience that a majority of the leakage experienced

during Type A testing is through components tested by Type B and C

tests.

NUREG-1493 found that these observations, together with the

insensitivity of reactor accident risk to the containment leakage

rate, show that increasing the Type A leakage test intervals would

have a minimal impact on public risk.

Type B and C Testing

NUREG-1493 found that while Type B and C tests can identify the

vast majority (greater than 95 percent) of all potential leakage

paths, performance-based alternatives to current local leakage-

testing requirements are feasible without significant risk impacts.

The risk model used in NUREG-1493 suggests that the number of

components tested would be reduced by about 60 percent with less

than a three-fold increase in the incremental risk due to

containment leakage. Since, under existing requirements, leakage

contributes less than 0.1 percent of overall accident risk, the

overall impact is very small. In addition, the NRC's Final

Regulatory Impact Analysis concluded that while the extended testing

intervals for Type B and C tests led to minor increases in potential

offsite dose consequences, the beneficial expected decrease in

onsite (LLRT [local leak rate testing] & ILRT worker) dose exceeds

(by at least an order of magnitude) the potential off-site dose

consequences.

The editorial change to the bases has no impact on the

probability or consequence of an accident since it is strictly a

correction to achieve consistency between the bases and the

specifications.

Based on the above, DECO [the licensee] has concluded that the

proposed change will not result in a significant increase in the

probability or consequences of any accident previously evaluated.

2. The request does not create the possibility of occurrence of

a new or different kind of accident from any accident 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 involves the reduction of

Type A, B, and C test frequency. Except for the method of defining

the test frequency, the methods for performing the actual tests are

not changed. No new accident modes are created by extending the

testing intervals. No safety-related equipment or safety functions

are altered as a result of this change. Extending the test frequency

has no influence on, nor does it contribute to, the possibility of a

new or different kind of accident or malfunction from those

previously analyzed.

The editorial change to the bases has no effect on any kind of

accident since it is strictly a correction to achieve consistency

between the bases and the specifications.

Based on the above, DECO has concluded that the proposed change

will not create the possibility [of] a new or different kind of

accident previously evaluated.

3. The request does not involve a significant reduction in a

margin to safety.

The proposed change only affects the frequency of Type A, B, and

C testing. Except for the method of defining the test frequency, the

methods for performing the actual tests 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.

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 Fermi 2 Technical

Specifications to be 0.5 percent by weight of the containment air

per 24 hours at 56.5 psig (Pa). The limitation on containment

leakage rate is designed to ensure that total leakage volume will

not exceed the value assumed in the accident analyses at the peak

accident pressure (Pa). The margin to 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.

Except for the method of defining the test frequency, no change

in the method of testing is being proposed. The Type B and C tests

will continue to be done at full pressure (Pa) or greater with

the exception of the Main Steam Isolation Valves, which have an

approved exemption. Other programs are in

[[Page 7552]]

place to ensure that proper maintenance and repairs are performed

during the service life of the primary containment and systems and

components penetrating the primary containment.

The editorial change to the bases has no effect on the margin of

safety since it is strictly an editorial change to achieve

consistency between the bases and the specifications.

As a result, DECO has concluded that the proposed change will

not result in 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: Monroe County Library System,

3700 South Custer Road, Monroe, Michigan 48161.

Attorney for licensee: John Flynn, Esq., Detroit Edison Company,

2000 Second Avenue, Detroit, Michigan 48226.

NRC Project Director: John N. Hannon.

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: January 22, 1996.

Description of amendment request: The proposed amendment would

modify the steam generator tube plugging criteria in Technical

Specification 3/4.4.5, Steam Generators, and the allowable leakage in

Technical Specification 3/4.4.6.2, Operational Leakage, and the

associated Bases. The amendment would allow the implementation of

alternate steam generator tube plugging criteria for the tube support

plate (TSP)/tube intersections for Unit 1.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

Structural Considerations

Industry testing of model boiler and operating plant tube

specimens for free span tubing at room temperature conditions show

typical burst pressures in excess of 5000 psi for indications of

outer diameter stress corrosion cracking with voltage measurements

at or below the structural limit of 4.0 volts. One model boiler

specimen with a voltage amplitude of 19 volts also exhibited a burst

pressure greater than 5000 psi. Burst testing performed on one

intersection pulled from STP Unit 1 in 1993 with a 0.51 volt

indication yielded a measured burst pressure of 8900 psi at room

temperature. Burst testing performed on another intersection pulled

from STP Unit 1 in 1995 with a 0.48 volt indication yielded a

measured burst pressure of 9950 psi at room temperature.

The projected end-of-cycle (EOC) voltage compares favorably with

the 4.7 volt structural limit considering the EPRI [Electric Power

Research Institute] voltage growth rate for indications at STP.

Using the methodology of the NRC Generic Letter 95-05, the

structural limit is reduced by allowances for uncertainty and growth

to develop a beginning-of-cycle (BOC) repair limit which should

preclude EOC indications from growing in excess of the structural

limit. The non-destructive examination (NDE) uncertainty to be

applied per EPRI is approximately 20 percent. The EPRI recommended

growth allowance of 30 percent/EFPY [effective full power year] is

also to be applied. This growth value is conservative for STP Unit 1

based on previous inspection history. By adding NDE uncertainty

allowances and a crack growth allowance to the repair limit, the

structural limit can be validated. Therefore, the maximum allowable

BOC repair limit (RL) based on the structural limit of 4.7 volts can

be represented as:

RL + (0.20 x RL) + (0.45* x RL) = 4.7 volts, which yields RL of 2.85

volts.

* The 30% growth rate for 1 EFPY was scaled up to the cycle

length used at South Texas.

This repair limit (2.85 volts) reasonably could be applied for

APC [alternate plugging criteria] implementation to repair bobbin

indications greater than the 1.0 volt criterion specified by NRC

Generic Letter 95-05 and is independent of RPC [rotating pancake

coil-probe] confirmation of the indications. STP has chosen to use a

steam generator tube upper repair limit of 2.85 volts to assess tube

integrity for those bobbin indications which are above 1.0 volt but

do not have confirming RPC calls. This 2.85 volt upper limit for

non-confirmed RPC calls is consistent with the NRC Generic Letter

95-05. Since the upper bound for repair of non-confirmed RPC is

limited to a value far less than the structural limit associated

with a full alternate criteria, the establishment of the repair

limits are determined to be reasonable and conservative with respect

to the industry pulled tube data base used.

Leakage Considerations

As part of the implementation of APC, the distribution of EOC

cracking indications at the TSP intersections has been used to

calculate the primary-to-secondary leakage which is bounded by the

maximum leakage required to remain within applicable dose limits.

This limit was calculated using the Technical Specification RCS

[reactor coolant system] Iodine-131 transient spiking values

consistent with NUREG-0800. Application of the APC criteria requires

the projection of postulated MSLB [main steam line break] leakage

based on the projected EOC voltage distribution for the beginning of

cycle. Projected EOC voltage distribution is developed using the

most recent EOC eddy current results and a voltage measurement

uncertainty. Draft NUREG-1477 requires that all indications to which

APC is applied must be included in the leakage projection.

The projected MSLB leakage rate calculation methodology

prescribed in EPRI TR-100407 will be used to calculate the EOC

leakage. A Monte Carlo approach will be used to determine the EOC

leakage, accounting for all of the ECT [eddy current testing]

uncertainties, voltage growth, and an assumed probability of

detection (POD) of 0.6 for a 1.0 volt repair limit. The fitted

logarithmic function probability of leakage correlation will be used

to establish the STP MSLB leak rate used for comparison with a

bounding allowable leak rate in the faulted loop which would result

in radiological consequences which are within applicable dose

limits. Due to the relatively low voltage levels of indications at

STP and low voltage growth rates, it is expected that the actual

calculated leakage values will be far less than this limit.

Therefore, implementation of APC does not adversely affect steam

generator tube integrity and implementation will be shown to result

in acceptable dose consequences. The proposed amendment does not

result in any increase in the probability or consequences of an

accident previously evaluated.

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

kind of accident from any accident previously evaluated?

Implementation of the proposed steam generator tube alternate

plugging criteria for ODSCC [outer diameter stress corrosion

cracking] at the TSP intersections does not introduce any

significant changes to the plant design basis. Use of the criteria

does not provide a mechanism which could result in an accident

outside of the region of the TSP elevations since no ODSCC has been

identified outside the thickness of the TSPs. It is therefore

expected that for all plant conditions, neither a single or multiple

tube rupture event would occur in a steam generator where APC has

been applied.

Specifically, STP will implement, for Unit 1, a maximum leakage

rate of 150 gpd [gallons per day] per steam generator (SG) to help

preclude the potential for excessive leakage during all plant

conditions. The current technical specification limits on primary-

to-secondary leakage at operating conditions are 1 gpm [gallon per

minute] for all steam generators or 500 gpd for any one SG. The RG

[Regulatory Guide] 1.121 criterion for establishing operational

leakage rate limits governing plant shutdown is based upon leak-

before-break (LBB) considerations to detect a free span crack before

potential tube rupture as a result of faulted plant conditions. The

150 gpd limit is intended to provide for leakage detection and plant

shutdown in the event of an unexpected crack propagation resulting

in excessive leakage. RG 1.121 acceptance criteria for establishing

operating leakage limits are

[[Page 7553]]

based on LBB considerations such that plant shutdown is initiated if

the permissible crack is exceeded.

The predicted EOC leakage for STP is based on the calculated

growth rate and does not take credit for the TSP proximity during

normal operation. Thus, the 150 gpd limit provides for plant

shutdown prior to reaching critical crack lengths. Additionally,

this leak-before-break evaluation assumes that the entire crevice

area is uncovered during the secondary side blowdown of a MSLB.

Typically, it is expected for the vast majority of intersections

that only partial uncovery will occur. Thus, the proximity of the

TSP will enhance the burst capacity of the tube.

Steam generator tube integrity is continually maintained through

inservice inspection and primary-to-secondary leakage monitoring.

Any tubes falling outside the APC repair limits are removed from

service. Therefore, the possibility of a new or different kind of

accident from any accident previously developed is not created.

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

of safety?

The use of the voltage based bobbin probe for dispositioning

ODSCC degraded tubes within TSP intersections by APC is demonstrated

to maintain steam generator tube integrity in accordance with the

requirements of RG 1.121. RG 1.121 describes a method acceptable to

the NRC staff for meeting GDCs [General Design Criterion] 14, 15,

31, and 32 by reducing the probability or the consequences of steam

generator tube rupture. This is accomplished by determining the

limiting conditions of degradation of steam generator tubing, as

established by inservice inspection, for which tubes with

unacceptable cracking are removed from service. Upon implementation

of the criteria, even under the worst case conditions, the

occurrence of ODSCC at the TSP elevation is not expected to lead to

a steam generator tube rupture event during normal or faulted plant

conditions. The EOC distribution of crack indications at the TSP

elevations will be confirmed to result in acceptable primary-to-

secondary leakage during all plant conditions and that radiological

consequences are not adversely impacted.

In addressing the combined effects of loss of coolant accident

(LOCA) and safe shutdown earthquake (SSE) on the steam generator

component (as required by GDC 2), it has been determined that tube

collapse may occur in the steam generators at some plants. This is

the case at STP as the TSP may become deformed as a result of

lateral loads at the wedge supports at the periphery of the plate

due to the combined effects of the LOCA rarefaction wave and SSE

loadings. The resulting secondary-to-primary pressure differential

on the deformed tubes may cause some of the tube to collapse.

There are two concerns associated with steam generator tube

collapse. First, the collapse of steam generator tubing reduces the

RCS flow area through the tubes. The reduction in flow area

increases the resistance to flow of steam from the core during a

LOCA which, in turn, may potentially increase peak clad temperature

(PCT). Second, there is a potential that through wall cracks in

tubes could sufficiently enlarge during tube deformation or

collapse, causing sufficient in-leakage of secondary water back to

the core which dilutes the poisoning effect of boron injection from

the emergency cooling system. Again, an increase in core PCT may

result.

Consequently, since the LBB methodology is applicable to the STP

reactor coolant loop piping, the probability of breaks in the

primary loop piping is sufficiently low that they need not be

considered in the structural design of the plant. The analysis

identified tubes located adjacent to wedge regions that are subject

to potential collapse during combined LOCA and SSE. These tubes will

be excluded from application of APC. Thus, existing tube integrity

requirements apply to these tubes and the margin of safety is not

reduced.

Implementation practices using the bobbin probe voltage based

tube plugging criteria bounds RG 1.83 considerations by:

(1) Using enhanced eddy current inspection guidelines consistent

with those used by EPRI in developing the correlations. This

provides consistency in voltage normalization,

(2) Performing a 100 percent bobbin coil inspection for all hot

leg tube support plate intersections and all cold leg intersections

down to the lowest cold leg tube support plate with outer diameter

stress corrosion cracking (ODSCC) indications. The determination of

the tube support plate intersections having ODSCC indications shall

be based on the performance of at least a 20% random sampling of

tubes inspected over their full length, and

(3) Incorporating RPC inspection for all tubes with larger

indications left in service. This further establishes the principal

degradation morphology as ODSCC.

Implementation of APC at TSP intersections will decrease the

number of tubes which must be repaired. Since the installation of

tube plugs (to remove ODSCC degraded tubes from service) reduces the

RCS flow margin, APC implementation will help preserve the margin of

flow that would otherwise be reduced.

For each cycle the projected EOC primary-to-secondary leak rate

allowed is bounded by a leak rate which limits the radiological

consequences of a EOC MSLB to within applicable dose limits.

Therefore, this change does not involve a significant reduction in

the margin to safety.

It is therefore concluded that the proposed license amendment

request does not result in a significant reduction in the margin of

safety as defined in the plant Final Safety Analysis Report or

Technical Specifications.

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

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

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

request for amendments involves no significant hazards consideration.

Local Public Document Room location: Wharton County Junior College,

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

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

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

NRC Project Director: William D. Beckner.

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: January 22, 1996.

Description of amendment request: The proposed amendment would

modify the steam generator tube plugging criteria in Technical

Specification 3/4.4.5, Steam Generators, and the associated Bases, to

allow the implementation of alternate steam generator tube plugging

criteria for the tube-to-tubesheet joints (known in the industry as F*)

for Unit 1.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The proposed changes to the Steam Generator section of Technical

Specifications do not affect any accident initiators or precursors

and do not alter the design assumptions for the systems or

components used to mitigate the consequences of an accident. The

requirements approved by the NRC will not be reduced by this

request. Since F* utilizes the ``as rolled'' tube configuration that

exists as part of the original steam generator design, all of the

design and operating characteristics of the steam generator and

connected systems are preserved. The F* joint has been analyzed and

tested for design, operating and faulted condition loadings in

accordance with Regulatory Guide 1.121 safety factors. At worst

case, a tube leak would occur with the result being a primary to

secondary leak.

Should a tube leak occur, the impact is bounded by the ruptured

tube evaluation submitted by STP for the Unit 1 operating license.

No new or unreviewed accident conditions are created by the use of

F* criteria. The potential for a tube rupture is not increased from

the original submittal, thus there is no impact on accidents

evaluated as the design basis. Therefore use of the F* criteria will

not increase the probability of occurrence of an accident previously

evaluated.

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

or different kind of accident from any accident previously

evaluated.

[[Page 7554]]

The use of the proposed F* alternate plugging criteria will not

introduce significant or adverse changes to the plant design basis.

The failure of a tube which remained unplugged in accordance with

the F* criteria would result in a tube leak, which is a previously

analyzed condition. Since this leak would occur below the secondary

face of the tubesheet, its leak rate would be limited by the tube-

to-tubesheet interface. Qualification testing and previous

experience indicates that normal and faulted leakage would be well

below the technical specification limits creating no threat

associated with tube rupture type leakages. This conclusion is

consistent with previous F* programs approved and used at other

operating plants.

However, in the unlikely event the failed tube severed

completely at a point below the F* region, the remaining F* joint

would retain engagement in the tubesheet due to its length of

expanded contact within the tubesheet bore, preventing any

interaction with neighboring tubes. If the tube severs at a point

above the F* region, then it is covered by the tube rupture event as

a part of the UFSAR [Updated Final Safety Analysis Report]. Thus,

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

accident previously evaluated is not created.

3. The proposed change does not involve a significant reduction

in a margin of safety.

Based on previous responses (above), the protective boundaries

of the steam generator are preserved. A tube with degradation can be

kept in service through F* criteria which provided an un-degraded

expanded interface with the tubesheet and which satisfies all of the

necessary structural and leakage requirements in accordance with

Regulatory Guide 1.121 and the Technical Specifications. Since the

joint is constrained within the tubesheet bore there is no

additional risk associated with tube rupture. Since the UFSAR

analyzed accident scenarios remain bounding, the use of an F*

criteria does not reduce the margin of safety.

Thus, these changes do not involve a significant reduction in

the margin of safety. Therefore, based on the above evaluation, STP

has concluded that these changes do not involve any significant

hazards considerations.

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

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

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

request for amendments involves no significant hazards consideration.

Local Public Document Room location: Wharton County Junior College,

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

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

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

NRC Project Director: William D. Beckner.

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: January 12, 1996 (AEP:NRC:1233).

Description of amendment requests: The proposed amendments would

modify technical specification section 4.4.11 to eliminate the

surveillance requirement (SR) demonstrating operability of the

emergency power supply for the pressurizer power-operated relief valves

(PORVs) and block valves.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

hazards consideration if the change does not:

1. involve a significant increase in the probability or

consequence 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 proposed change is consistent with NUREG-1431 [Standard

Technical Specifications Westinghouse Plants]. Due to the high

reliability and continued testing of the Class 1E power supply, we

conclude that the elimination of the SR will not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

Criterion 2

The proposed change does not involve the addition of any new

plant operation or procedures, and the elimination of the SR is

consistent with NUREG-1431. For these reasons, we believe that the

proposed change will not create the possibility of a new or

different kind of accident from any accident previously evaluated.

Criterion 3

The proposed change is consistent with NUREG-1431, and it does

not affect the acceptance criteria of any of the other PORV and

block valve tests currently performed. For these reasons, we believe

that the proposed amendment will not involve a significant reduction

in a margin of safety.

The NRC staff has reviewed the licensee's analysis and the

applicable Bases of the Standard Technical Specifications Westinghouse

Plants. The Bases for the applicable surveillance, 3.4.11.4, states

``This Surveillance is not required for plants with permanent 1E power

supplies to the valves.'' 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.

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

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

Connecticut

Date of amendment request: January 22, 1996.

Description of amendment request: The proposed change relocates the

containment isolation valve (CIV) list, Table 3.6-2, from the Technical

Specifications to the Technical Requirements Manual (TRM). This change

affects Technical Specifications Sections 1.8.1a, 4.6.1.1a, 3.6.3.1,

4.6.3.1.1 and 4.6.3.1.2, and the Basis Section 3/4.6.3. A note at the

bottom of Table 3.6-2 regarding the CIVs that are subject to

administrative control is retained in the Technical Specifications by

relocating it to Sections 1.8.1a and 4.6.1.1a. This change is being

performed in accordance with Generic Letter 91-08, which provides

guidance for removal of component lists from the Technical

Specifications.

Additionally, a change to provide relief in the surveillance

requirement in Section 4.6.1.1a is included. The change allows valves,

blind flanges, and deactivated automatic valves located inside the

containment and are locked, sealed, or otherwise secured in the closed

position to be verified closed during each cold shutdown but not more

often than once per 92 days. The current requirements check the valve

position once per 31 days.

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 (SHC), which is presented below:

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

has reviewed the proposed changes. NNECO concludes that these

changes do not involve a significant hazards consideration (SHC)

since the proposed changes satisfy the criteria in 10CFR50.92(c).

That is, the proposed changes do not:

[[Page 7555]]

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed change to remove the Containment Isolation Valve

(CIV) list from the Technical Specifications will not result in any

hardware or operating changes. The proposed change is based upon NRC

Generic Letter (GL) 91-08 and merely removes the CIV table and all

references to the table from the technical specifications without

affecting the operability requirements of any of the listed valves.

The technical specifications will continue to require the CIVs to be

operable. Limiting Condition for Operation and surveillance

requirements for the valves will also remain in the technical

specifications. The CIV table will be relocated to the Millstone

Unit No. 2 Technical Requirements Manual (TRM) which is controlled

in accordance with 10CFR50.59.

This change is administrative in nature and does not involve an

increase in the probability or consequence of an accident previously

evaluated. Furthermore, the proposed change does not alter the

design, function, or operation of the valves involved, and therefore

does not affect the probability or consequences of any previously

evaluated accident.

The change to Section 4.6.1.1a that reduces the surveillance

requirement for valves, blind flanges, and deactivated automatic

valves located inside the containment provides consistency with

NUREG-1432, ``Standard Technical Specifications for Combustion

Engineering Plants'' as well as the Technical Specifications of

Millstone Unit No. 3, Haddam Neck Plant, and Seabrook. The

probability or consequences of any previously evaluated accidents

are not affected.

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

from any accident previously evaluated.

The change to relocate the CIV list from the technical

specifications to the TRM will not impose any different operational

or surveillance requirements, nor will the change remove any such

requirements. Adequate control of information will be maintained.

Furthermore, as stated above, the proposed change does not alter the

design, function, or operation of the valves involved, and therefore

no new accident scenarios are created.

The change to Section 4.6.1.1a that reduces the surveillance

requirement for valves, blind flanges, and deactivated automatic

valves located inside the containment does not alter the design,

function, or operation of the valves involved, and therefore no new

accident scenarios are created.

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

The proposed changes will not reduce the margin of safety since

it has no impact on any safety analysis assumption. The proposed

changes do not decrease the scope of equipment currently required to

be operable or subject to surveillance testing, nor does the

proposed change affect any instrument setpoints or equipment safety

functions.

The relocation of the valve list is consistent with the guidance

provided in GL 91-08. The change to the surveillance interval is

consistent with NUREG-1432, ``Standard Technical Specifications for

Combustion Engineering Plants'' as well as the Technical

Specifications of Millstone Unit No. 3, Haddam Neck Plant, and

Seabrook. The intent of the technical specification will be met

since the change will not alter function or operability requirements

for any CIV.

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: January 17, 1996.

Description of amendment request: The amendment request would

delete a license requirement to submit responses to and to implement

requirements of Generic Letter 83-28, because the requirement has been

completed. Generic Letter 83-28 pertains to the Salem anticipated

transient without scram (ATWS) event.

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 SHC because the

change would not:

1. Involve a significant increase in the probability or

consequences of an accident previously analyzed.

NNECO's proposal to delete License Condition 2.C(4) is an

administrative change. The NRC Staff has accepted Millstone Unit No.

3's responses regarding the actions required by GL 83-28, thus, the

license condition has been met and is no longer necessary. The

proposed change does not affect the configuration, operation, or

performance of any system, structure, or component. Additionally,

the limiting conditions for operation, limiting safety system

settings, and safety limits specified in the Millstone Unit No. 3

Technical Specifications are unchanged. Therefore, 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 NRC Staff has accepted Millstone Unit No. 3's responses

regarding the actions required by GL 83-28, thus, the license

condition has been met and is no longer necessary. The proposed

change to delete License Condition 2.C(4) does not affect the

configuration, operation, or performance of any system, structure,

or component. Additionally, the limiting conditions for operation,

limiting safety system settings, and safety limits specified in the

Millstone Unit No. 3 Technical Specifications are unchanged.

Therefore, this proposed change cannot create the possibility of a

new or different kind of accident from any previously analyzed.

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

The NRC Staff has accepted Millstone Unit No. 3's responses

regarding the actions required by GL 83-28, thus, the license

condition has been met and is no longer necessary. The proposed

change to delete License Condition 2.C(4) does not affect the

configuration, operation, or performance of any system, structure,

or component. Additionally, the limiting conditions for operation,

limiting safety system settings, and safety limits specified in the

Millstone Unit No. 3 Technical Specifications are unchanged.

Therefore, this 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.

Philadelphia Electric Company, Docket Nos. 50-352 and 50-353, Limerick

Generating Station, Units 1 and 2, Montgomery County, Pennsylvania

Date of amendment request: December 22, 1995.

Description of amendment request: The proposed changes will revise

Limerick Generating Station, Units 1 and 2, Technical Specification

3.6.1.8 ``Drywell and Suppression Chamber Purge System,'' increasing

the Drywell and Suppression Chamber Purge System operating time limit

from 90 hours each 365 days to 180 hours each 365 days.

Basis for proposed no significant hazards consideration

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

provided its analysis of the

[[Page 7556]]

issue of no significant hazards consideration, which is presented

below:

1. The proposed Technical Specification [TS] changes do not

involve a significant increase in the probability or consequences of

an accident previously evaluated.

These TS changes do not increase the probability of occurrence

of an accident previously evaluated in the SAR [Safety Analysis

Report]. This activity involves changing the allowable operating

limit for the Drywell and Suppression Chamber Purge System from 90

hours each 365 days to 180 hours each 365 days. This change

increases the probability that this system will be in service should

a LOCA [loss of coolant accident] occur, but does not increase the

probability that a LOCA will occur.

Increasing the operating limit for the Drywell and Suppression

Chamber Purge System from 90 hours to 180 hours each 365 days does

not increase the consequences of a LOCA as previously evaluated in

the SAR. These proposed TS changes increase the probability of a

LOCA occurring during the time the Drywell and Suppression Chamber

Purge System is in operation, and therefore, increase the

probability of the failure of the operating SGTS [Standby Gas

Treatment System] filter bank. However, the risk to containment

integrity was previously evaluated and found to be acceptable (UFSAR

[Updated Final Safety Analysis Report] Section 9.4.5.1.2.2 and

WASH--1400 ``Reactor Safety Study'').

Increasing the duration that the vent/purge line isolation

valves may be open does not increase the probability that these

valves will not perform as designed (i.e., close upon receipt of an

isolation signal) in response to a LOCA. However, the changes will

increase the likelihood that the vent and purge valves will be

called on to close. As discussed in UFSAR Section 6.2.4.2, the

containment purge valves have undergone extensive testing and

analyses to demonstrate the operability of these valves following a

LOCA.

In addition to the existing Safety Analysis Report (SAR)

evaluations, a Level 2 PSA [Probabilistic Safety Assessment]

Analysis (containment failure) was performed to determine the

additional risk associated with changing the operating limit from 90

to 180 hours each 365 days. The PSA evaluation conservatively

assumed a 200 hour vent/purge duration per a 365 day period. The

figure of merit evaluated is the large early release frequency

(LERF) which represents the likelihood of containment failure

following core damage that could significantly affect the public

(e.g., release of a large amount of radioactive material early

enough in the accident that evacuation of the public has not

occurred). The 200 hour vent/purge duration increased the LERF

approximately 3% from the base value of 2.57E-8 for all PSA

initiators. This analysis concluded that the increase in risk of

containment failure is well within the bounds of the EPRI

[Electrical Power Research Institute] PSA Applications Guideline for

permanent changes. The same relative increase applies to the large

Design Basis Accident LOCA LERF.

These changes do not directly or indirectly degrade the

performance of any other safety systems (assumed to function in the

accident analysis) below their design basis. The potential for other

equipment failures in the reactor enclosure due to duct-work impact,

impingement, and the resulting environmental conditions was

evaluated. It was concluded that the environmental qualifications

for the LGS equipment are sufficient to ensure operability under the

predicted environmental conditions, and there is no impact or

impingement-related damage to essential equipment. Although the

probability of occurrence of a malfunction of equipment important to

safety is increased, the existing SAR analysis and Level 2 PSA

Analysis demonstrate the increased risk and radiological

consequences are not significant.

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

new or different kind of accident from any accident previously

evaluated.

This activity does not change the function of the Drywell and

Suppression Chamber Purge System, the containment isolation system,

or SGTS as previously evaluated in the SAR. Changing the duration of

operation of the vent and purge system does not create an accident

initiator not considered in the SAR. Therefore, the possibility of

an accident of a different type is not created.

This activity does not create a failure mode not considered in

the SAR. All possible equipment failures that could occur as a

result of a LOCA during high volume purging have previously been

identified and evaluated in the SAR. Therefore, this activity does

not create the possibility of a different type of malfunction of

equipment important to safety.

Therefore, the proposed TS changes will not create the

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

previously evaluated.

3. The proposed TS changes do not involve a significant

reduction in a margin of safety.

The Bases of Technical Specification 3.6.1.8 states that the

intent of the 90 hour per 365 day operating limit for the Drywell

and Suppression Chamber Purge System is to protect the integrity of

the SGTS filters. As discussed above, the requirements specified in

ODCM paragraph 3.3.6 assure the availability of the backup SGTS

filter train during operation of the vent and purge system.

Furthermore, as discussed above, revising the operating limit from

90 hours to 180 hours each 365 days does not involve a significant

increase in risk. The margin of safety as defined in the Bases of

Technical Specification 3.6.1.8 is maintained.

Therefore, the implementation of the proposed TS changes will

not involve a significant reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Pottstown Public Library, 500

High Street, Pottstown, Pennsylvania 19464.

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

General Counsel, Philadelphia Electric Company, 2301 Market Street,

Philadelphia, Pennsylvania 19101.

NRC Project Director: John F. Stolz.

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 amendment request: February 6, 1996.

Description of amendment request: The amendments would change the

Technical Specifications to lower the 125 Volt Battery Charger

surveillance amperage from at least 200 amps to at least 170 amps.

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. Will not involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed amendment will permit replacement of aging battery

chargers while ensuring these replacement battery chargers will

restore the battery from the design minimum charge to its fully

charged state while supplying normal steady-state loads. This meets

the design basis for the 125V DC system and is consistent with Salem

Unit 1 and 2 commitment to IEEE 308-1971 in UFSAR Section 3A.

The 125V DC battery chargers are not addressed as a contributor

to any accident analyzed in the UFSAR, therefore, changes to the

battery charger output current will not increase the probability of

an accident occurring.

The limiting analyzed accident considered in this proposed TS

amendment is the Loss of Offsite Power coincident with a Loss of

Coolant Accident. This is currently the limiting design duty cycle

for the batteries. The 125V batteries are sized to maintain all

emergency loads for a period of 2 hours without battery chargers.

This is demonstrated by performing the surveillance specified in TS

4.8.2.3.2.f, which is not being changed. Since the chargers are not

required to be available during this 2 hour period, and since the

proposed charging rate will supply the necessary loads following

restoration of AC power, the proposed amendment will have no effect

on the consequences of this accident.

The current limiter is calculated to extend the recharging time

from 20 hours to 30 hours, but this is not considered significant

since two, sequential battery discharge events are not considered

plausible.

PSE&G calculation substantiates the capability of the chargers

to restore the battery from the design minimum charge to its fully

charged state while supplying

[[Page 7557]]

normal steady-state loads following a Station Blackout (SBO) Event

which exceeds the current design duty cycle.

In addition, a review of 125V DC Battery System load profiles

indicated that the battery chargers are capable of supplying

expected loads when restoring the battery from a design minimum

charge state to a fully charged state irrespective of the status of

the plant.

Therefore, the proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

2. Will not create the possibility of a new or different kind of

accident from any previously evaluated.

The proposed amendment does not result in any design or physical

configuration changes to the 125V DC system. This change supports

the installation of the replacement chargers and ensures the

chargers are surveilled within the bounds of limiting input

amperage. No changes are being made to the function, design basis,

or operation of the 125V DC system by this proposed change.

Therefore, the proposed amendment will not create the possibility of

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

3. Will not involve a significant reduction in a margin of

safety.

The proposed amendment to TS 4.8.2.3.2.e ensures that the

replacement battery chargers have sufficient capacity to restore

each 125V battery from the design minimum charge to its fully

charged state while supplying normal steady-state loads. A margin of

safety is maintained on both the AC input and DC output of the

chargers since the specified current is above that required to

support the 125V DC system and will result in AC current below the

ampacity rating of the battery charger input cables.

Testing to a charger output current of at least 170 amps will

maintain a margin of safety to the current required during actual

worst case normal loading on the 125V DC buses.

Therefore, the proposed amendment will not involve a significant

reduction in a margin of safety.

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

this review, it appears that the three standards of 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: Salem Free Public library, 112

West Broadway, Salem, New Jersey 08079.

Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston and

Strawn, 1400 L Street, NW, Washington, DC 20005-3502.

NRC Project Director: John F. Stolz.

Rochester Gas and Electric Corporation, Docket No. 50-244, R. E. Ginna

Nuclear Power Plant, Wayne County, New York

Date of amendment request: February 9, 1996.

Description of amendment request: The proposed amendment would

allow an installed overhead door assembly, to be used in lieu of the

equipment hatch closure, to isolate the hatch opening to the

containment building during fuel movement and core alterations.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Operation of Ginna Station in accordance with the proposed

changes does not involve a significant increase in the probability

or consequences of an accident previously evaluated. Containment

closure is used with respect to the mitigation of fuel handling

accidents, and as such, any change to these requirements will not

affect the probability of an accident. The proposed changes will

also not result in a significant increase in the consequences of an

accident previously analyzed since the technical specification

requirements remain bounded by the fuel handling accident assumption

of no containment closure.

2. Operation of Ginna Station in accordance with the proposed

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

of accident from any accident previously evaluated. The proposed

changes do not involve a physical alteration of the plant (i.e., no

new or different type of equipment will be installed) or changes in

the methods governing normal plant operation. The proposed changes

will not impose any new or different requirements. Thus, this change

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

accident from any accident previously evaluated.

3. Operation of Ginna Station in accordance with the proposed

changes does not involve a significant reduction in a margin of

safety. Containment closure is not assumed in the accident analyses

for Ginna Station. Also, the proposed change remains acceptable with

respect to SRP [NUREG-800, ``Standard Review Plan for the Review of

Safety Analysis Reports for Nuclear Power Plants, July 1981'']

15.7.4 and GDC [General Design Criterion] 19 requirements.

Therefore, no question of safety is involved, and the 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: Rochester Public Library, 115

South Avenue, Rochester, New York 14610.

Attorney for licensee: Nicholas S. Reynolds, Winston & Strawn, 1400

L Street, NW., Washington, DC 20005.

NRC Project Director: Ledyard B. Marsh.

Rochester Gas and Electric Corporation, Docket No. 50-244, R.E.

Ginna Nuclear Power Plant, Wayne County, New York

Date of amendment request: February 9, 1996.

Description of amendment request: The proposed amendment would

incorporate the methodology for determining the Low Temperature

Overpressure Protection (LTOP) limits into the Administrative Controls

Section 5.6.6 of the Ginna Technical Specifications (TS). The proposed

amendment will allow the licensee to perform future LTOP evaluations,

using NRC-approved methodology, without requiring changes to the TS.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Operation of Ginna Station in accordance with the proposed

changes does not involve a significant increase in the probability

or consequences of an accident previously evaluated. The proposed

changes only require that future LTOP limits be developed using NRC

approved methodology as specified within the Administrative Controls

section and do not involve any technical changes. As such, these

changes are administrative in nature and do not impact initiators or

analyzed events or assumed mitigation of accident or transient

events. Therefore, these changes do not involve a significant

increase in the probability or consequences of an accident

previously analyzed.

2. Operation of Ginna Station in accordance with the proposed

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

of accident from any accident previously evaluated. The proposed

changes do not involve a physical alteration of the plant (i.e., no

new or different type of equipment will be installed) or changes in

the methods governing normal plant operation. The proposed changes

will not impose any new or different requirements. Thus, this change

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

accident from any accident previously evaluated.

3. Operation of Ginna Station in accordance with the proposed

changes does not involve a significant reduction in a margin of

safety. The proposed changes will not reduce a margin of plant

safety because

[[Page 7558]]

the changes do not impact any safety analysis assumptions other than

requiring future evaluations of LTOP limits to be performed in

accordance with NRC approved methodology. These changes are

administrative in nature. As such, no question of safety is

involved, and the 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: Rochester Public Library, 115

South Avenue, Rochester, New York 14610.

Attorney for licensee: Nicholas S. Reynolds, Winston & Strawn, 1400

L Street, NW., Washington, DC 20005.

NRC Project Director: Ledyard B. Marsh.

Rochester Gas and Electric Corporation, Docket No. 50-244, R.E.

Ginna Nuclear Power Plant, Wayne County, New York

Date of amendment request: February 9, 1996.

Description of amendment request: The proposed amendment would

revise the Technical Specifications setpoints for steam generator (SG)

water level-high feedwater isolation function. It would take advantage

of a greater allowable operating band for SG water level afforded by

replacement SGs.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Operation of Ginna Station in accordance with the proposed

changes does not involve a significant increase in the probability

or consequences of an accident previously evaluated. The proposed

setpoint change does not degrade the performance of any plant

equipment. Therefore, the probability of an accident is not

increased. Since the revised trip setpoint and allowable value

remain bounded by the accident analysis value of 100% steam

generator narrow range level, the consequences of any accident are

not adversely affected.

2. Operation of Ginna Station in accordance with the proposed

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

of accident from any accident previously evaluated. The proposed

change does not involve a physical alteration to the plant (i.e., no

new or different types of equipment will be installed) or changes in

the methods 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.

3. Operation of Ginna Station in accordance with the proposed

changes does not involve a significant reduction in a margin of

safety. The revised setpoint and allowable value remain bounded by

the accident analysis assumptions. The existing values are based on

design considerations and not accident analysis parameters. The

replacement steam generators are not restricted by the same design

considerations with respect to the ESFAS [engineered safety features

actuation system] Steam Generator Water Level--High function.

Therefore, this change does not involve a 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: Rochester Public Library, 115

South Avenue, Rochester, New York 14610.

Attorney for licensee: Nicholas S. Reynolds, Winston & Strawn, 1400

L Street, NW., Washington, DC 20005.

NRC Project Director: Ledyard B. Marsh.

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

Callaway County, Missouri

Date of amendment request: February 9, 1996.

Description of amendment request: The proposed amendment would

change Technical Specification 5.3.1 to allow the use of Zirlo fuel

cladding material.

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 methodologies used in the accident analysis remain

unchanged. The proposed changes do not change or alter the design

assumptions for the systems or components used to mitigate the

consequences of an accident. Use of ZIRLO fuel cladding does not

adversely affect fuel performance or impact nuclear design

methodology. Therefore accident analyses are not impacted.

The operating limits will not be changed and the analysis

methods to demonstrate operation within the limits will remain in

accordance with NRC approved methodologies. Other than the changes

to the fuel assemblies, there are no physical changes to the plant

associated with this technical specification change. A safety

analysis will continue to be performed for each cycle to demonstrate

compliance with all fuel safety design bases.

VANTAGE 5 fuel assemblies with ZIRLO clad fuel rods meet the

same fuel assembly and fuel rod design bases as other VANTAGE 5 fuel

assemblies. In addition, the 10 CFR 50.46 criteria are applied to

the ZIRLO clad rods. The use of these fuel assemblies will not

result in a change to the reload design and safety analysis limits.

Since the original design criteria are met, the ZIRLO clad fuel rods

will not be an initiator for any new accident. The clad material is

similar in chemical composition and has similar physical and

mechanical properties as Zircaloy-4. Thus, the cladding integrity is

maintained and the structural integrity of the fuel assembly is not

affected. ZIRLO cladding improves corrosion performance and

dimensional stability. No concerns have been identified with respect

to the use of an assembly containing a combination of Zircaloy-4 and

ZIRLO clad fuel rods. Since the dose predictions in the safety

analyses are not sensitive to fuel rod cladding material, the

radiological consequences of accidents previously evaluated in the

safety analysis remain valid.

Therefore, the proposed changes do not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

VANTAGE 5 fuel assemblies with ZIRLO clad fuel rods satisfy the

same design bases as those used for other VANTAGE 5 fuel assemblies.

All design and performance criteria continue to be met and no new

failure mechanisms have been identified. The ZIRLO cladding material

offers improved corrosion resistance and structural integrity.

The proposed changes do not affect the design or operation of

any system or component in the plant. The safety functions of the

related structures, systems or components are not changed in any

manner, nor is the reliability of any structure, system or component

reduced. The changes do not affect the manner by which the facility

is operated and do not change any facility design feature, structure

or system. No new or different type of equipment will be installed.

Since there is no change to the facility or operating procedures,

and the safety functions and reliability of structures, systems or

components are not affected, the proposed changes do not create the

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

previously evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

Use of ZIRLO cladding material does not change the VANTAGE 5

reload design and safety limits. The use of these fuel assemblies

will take into consideration the normal core operating conditions

allowed in the Technical Specifications. For each cycle reload core,

the fuel assemblies will be

[[Page 7559]]

evaluated using NRC-approved reload design methods, including

consideration of the core physics analysis peaking factors and core

average linear heat rate effects.

The use of Zircaloy-4, ZIRLO or stainless steel filler rods in

fuel assemblies will not involve a significant reduction in the

margin of safety because analyses using NRC-approved methodologies

will be performed for each configuration to demonstrate continued

operation within the limits that assure acceptable plant response to

accidents and transients. These analyses will be performed using

NRC-approved methods that have been approved for application to the

fuel configuration.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Callaway County Public

Library, 710 Court Street, Fulton, Missouri 65251.

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

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

NRC Project Director: William H. Bateman.

Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339,

North Anna Power Station, Units No. 1 and No. 2, Louisa County,

Virginia

Date of amendment request: January 30, 1996.

Description of amendment request: The proposed amendments would

modify the Technical Specifications to increase the minimum allowable

reactor coolant system total flow rate from 284,000 gpm (for Unit 1)

and 275,300 gpm (for Unit 2) to 295,000 gpm for both units. Through the

1980's and into the 1990's the North Anna Unit 1 and 2 steam generators

experienced increasing levels of steam generator tube plugging. There

was a corresponding decrease in the reactor coolant flow rate. As a

result, the Commission issued several amendments in the 1989 to 1992

time frame to reduce the minimum reactor coolant flow rate.

Subsequently, the licensee replaced the steam generators in both units,

with steam generators having an increased number of tubes compared to

the replaced steam generators. With the increased number of tubes and

less flow resistance, a greater reactor coolant flow rate is

attainable. When the amendments were issued decreasing the minimum

required reactor coolant flow rate, the transmittal letters stated the

revision was temporary and would be increased when the steam generators

were replaced.

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 probability of occurrence or the consequences of an

accident or malfunction of equipment important to safety previously

evaluated in the safety analysis report would not increase. The

proposed Technical Specifications change only increases the minimum

allowable RCS total flow rate in the applicable Limiting Condition

of Operation. No other changes are being made to allowable operating

conditions defined by Technical Specifications, procedures, or to

any plant design feature by the implementation of this change. There

is no impact on the actual plant performance. Changes in the assumed

initial conditions for the accident have no bearing on the

probability of occurrence of the assumed accident or malfunction.

The RCS flow rate is an assumption in applicable safety analyses.

Existing analyses of record have assumed RCS flow rates which are

bounding with respect to expected actual plant behavior. Therefore,

the implementation of the proposed Technical Specifications change

does not affect the probability nor increase the consequences of an

accident previously evaluated.

2. The possibility for an accident or malfunction of a different

type than any evaluated previously in the safety analysis report

would not be created. The proposed change to North Anna Units 1 and

2 Technical Specifications Table 3.2-1 does not involve any

alterations to the physical plant which would introduce any new or

unique operational modes or accident precursors. Only the allowable

value for measured Reactor Coolant System Total Flow Rate will be

changed.

3. The margin of safety as defined in the basis for any

technical specifications is not reduced. The proposed Technical

Specifications change only increases the minimum allowable RCS total

flow rate in the applicable Limiting Condition of Operation. The RCS

flow rate is an assumption in applicable safety analyses. Existing

analyses of record have assumed RCS flow rates which are bounding

with respect to expected actual plant behavior. Therefore, the

margin of safety is not reduced by the proposed increase in the

allowable RCS Total Flow Rate.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: The Alderman Library, Special

Collections Department, University of Virginia, Charlottesville,

Virginia 22903-2498.

Attorney for licensee: Michael W. Maupin, Esq., Hunton and

Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond,

Virginia 23219.

NRC Project Director: David B. Matthews.

Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339,

North Anna Power Station, Units No. 1 and No. 2, Louisa County,

Virginia

Date of amendment request: January 31, 1996.

Description of amendment request: The amendments would revise the

Technical Specifications to reduce the minimum volume of fuel that must

be maintained in the diesel generator day tanks from 750 to 450

gallons. The amendments would also revise the surveillance requirements

for the diesel generators to permit some surveillances to be performed

while the reactor units are at power where the licensee considers it

safe to do so without compromising the availability of the diesel

generators to perform their intended function.

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 an increase in the probability of occurrence of an

accident previously evaluated.

The proposed changes do not result in any physical modifications

to any plant systems or components nor change the operation of any

plant equipment. The EDG [emergency diesel generator] fuel oil

supply system will continue to provide adequate fuel supply to the

EDGs in a manner consistent with applicable accident analyses.

Performing surveillance tests or portions of surveillance tests at

power that do not jeopardize stable plant operations does not

increase the probability of occurrence of previously analyzed

accidents.

Therefore, there is no increase in the probability of occurrence

of any accident.

2. Increase the consequences of an accident previously

evaluated.

The proposed changes do not result in any physical modifications

to any plant systems or components nor change the operation of any

plant equipment. The EDG fuel oil system remains capable of

supplying the EDGs with sufficient quantities of fuel oil to provide

power for long term loss of offsite power. The EDG surveillances

will continue to be performed in a manner that will ensure that the

EDGs will be capable of performing their intended safety functions.

The proposed changes to the electrical distribution system

surveillances will continue to ensure that the electrical

distribution system remains

[[Page 7560]]

operable to power the required safety systems.

Therefore, these proposed changes will not result in an increase

in the consequences of any evaluated accidents.

3. Create the possibility for an accident of a different type

than was previously evaluated.

The proposed changes do not result in any physical modifications

to any plant systems or components nor change the operation of any

plant equipment. Only those surveillance tests or portions of

surveillance tests that do not jeopardize stable plant operation

will be performed at power. Overlap testing to fully test the

electrical distribution system protection functions does not

introduce any unique accident precursors. The EDG fuel oil system

remains capable of supplying the EDGs with sufficient quantities of

fuel oil to provide power for long term loss of offsite power. The

EDG surveillances will continue to be performed in a manner that

will ensure that the EDGs will be capable of performing their

intended safety functions.

Therefore, there are no new precursors generated that would

result in the possibility of a different type of an accident than

was previously evaluated in the SAR [Safety Analysis Report].

4. Decrease the margin of safety as described in the bases

section of Technical Specifications.

The EDG fuel oil system will continue to provide adequate fuel

supply in a manner consistent with applicable accident analyses. The

EDG surveillances will continue to be performed in a manner that

will ensure that the EDGs are capable of performing their intended

safety functions. The proposed changes to the electrical

distribution system surveillances will continue to ensure that the

electrical distribution system remains operable to power the

required safety systems.

Therefore, the margin of safety as described in the Technical

Specifications is not reduced.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: The Alderman Library, Special

Collections Department, University of Virginia, Charlottesville,

Virginia 22903-2498.

Attorney for licensee: Michael W. Maupin, Esq., Hunton and

Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond,

Virginia 23219.

NRC Project Director: David B. Matthews

Previously Published Notices of Consideration of Issuance of

Amendments to Facility Operating Licenses, Proposed No Significant

Hazards Consideration Determination, and Opportunity for a Hearing

The following notices were previously published as separate

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

published as individual notices either because time did not allow the

Commission to wait for this biweekly notice or because the action

involved exigent circumstances. They are repeated here because the

biweekly notice lists all amendments issued or proposed to be issued

involving no significant hazards consideration.

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

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

of the original notice.

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

Linn County, Iowa

Date of application for amendment: July 21, 1995, August 8, 1995,

and December 15, 1995.

Brief description of amendment request: The proposed amendment

would modify the requirements for testing an emergency diesel generator

(EDG) when the other is inoperable. The amendment would correct an

editorial error in the Duane Arnold Energy Center Operating License and

would correct an erroneous reference in the Technical Specification.

Date of publication of individual notice in Federal Register:

February 2, 1996 (61 FR 3953).

Expiration date of individual notice: March 4, 1996.

Local Public Document Room location: Cedar Rapids Public Library,

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

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

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

Date of amendment request: January 18, 1995.

Description of amendment request: The proposed amendment would

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

Nuclear Power Plant, Unit Nos. 1 and 2, to allow operation of Unit 1 in

Mode 3 (Hot Standby) during replacement of nonvital auxiliary

transformer 1-1. Specifically, TS 3/4.8.1.1, ``Electrical Power

Systems--A.C. Sources--Operating,'' Action Statement (a), would be

revised to permit a one-time extension of the allowed outage time (AOT)

from 72 hours to 120 hours.

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

FR 3737).

Expiration of individual notice: March 4, 1996.

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 amendment request: February 5, 1996, as supplemented by

letter dated February 14, 1996.

Brief description of amendment request: The amendment changes

Technical Specifications 4.6.2.3.b, ``Suppression Pool Cooling'', and

TS 4.6.2.2.b, ``Suppression Pool Spray'', to include flow through the

RHR heat exchanger bypass line (in addition to the RHR heat exchanger)

in the Suppression Pool Cooling and Suppression Pool Spray flow path

used during RHR pump testing.

Date of publication of individual notice in Federal Register:

February 9, 1996 (61 FR 5040).

Expiration date of individual notice: March 11, 1996.

Local Public Document Room location: Pennsville Public Library, 190

S. Broadway, Pennsville, New Jersey 08070.

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.

Brief description of amendment request: The proposed amendment

would change the Technical Specification surveillance frequency for the

drywell bypass leakage rate test from 18 months to 120 months (10

years) with a more frequent testing requirement if performance

degrades. Additionally, specific leakage limits would be deleted for

the air lock seal and barrel tests. Also, surveillance frequencies for

the air lock interlock test and seal pneumatic system leak test would

be changed from 18 months to 24 months. Finally, the surveillance

frequencies for the air lock barrel test would be changed from ``each

COLD SHUTDOWN if not performed within

[[Page 7561]]

the previous 6 months'' to ``at least once per 24 months'' and from 18

months to 24 months. The licensee requested that this amendment be

approved for use during the current refueling outage which began on

January 27, 1996.

Date of publication of individual notice in Federal Register:

February 2, 1996 (61 FR 3951).

Expiration date of individual notice: March 4, 1996.

Local Public Document Room location: Perry Public Library, 3753

Main Street, Perry, Ohio 44081.

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.

Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318,

Calvert Cliffs Nuclear Power Plant, Unit No. 1, Calvert County,

Maryland

Date of application for amendments: December 7, 1995.

Brief description of amendments: The amendments add the convolution

analytical technique for the analysis of the pre-trip main steam line

break event to the list of approved core operating limits analytical

methods listed in Technical Specification 6.9.1.9, ``Core Operating

Limits Report.'' The convolution analytical technique was previously

reviewed and approved by the NRC staff and the supporting safety

evaluation was provided to Baltimore Gas and Electric Company by letter

dated May 11, 1995.

Date of issuance: February 5, 1996.

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

30 days.

Amendment Nos.: 210 and 188.

Facility Operating License No. DPR-53 and DPR-69: Amendment revised

the Technical Specifications.

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

177)

The Commission's related evaluation of these amendments is

contained in a Safety Evaluation dated February 5, 1996.

No significant hazards consideration comments received: No.

Local Public Document Room location: Calvert County Library, Prince

Frederick, Maryland 20678.

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

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

Date of application for amendments: November 2, 1994, as

supplemented by letters dated November 16 and December 14, 1995.

Brief description of amendments: The amendments delete the content

of the Appendix B, ``Environmental Protection Plan'' (Non-radiological)

Technical Specifications and modify License Condition 2.C.(2) so as to

delete that portion which refers to the Environmental Protection Plan.

Date of issuance: February 5, 1996.

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

30 days.

Amendment Nos.: Unit 1-164--Unit 2-146.

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

revised the Technical Specifications and License Conditions.

Date of initial notice in Federal Register: March 1, 1995 (60 FR

11131). The November 16 and December 14, 1995, letters provided

clarifying information that did not change the scope of the November 2,

1994, application and the initial proposed no significant hazards

consideration determination.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated February 5, 1996.

No significant hazards consideration comments received: No.

Local Public Document Room location: Atkins Library, University of

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

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

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

Date of application for amendments: January 13, 1995, as

supplemented by letter dated August 30, 1995.

Brief description of amendments: The amendments revise the

Technical Specifications to increase the surveillance test intervals

and allowed outage times for the Reactor Trip System and Engineered

Safety Features Actuation System. The NRC staff has reviewed the

proposed changes and finds that, with one exception as noted in the

enclosed Safety Evaluation, the amendments conform to WCAP-10271,

``Evaluation of Surveillance Frequencies and Out of Service Times for

the Reactor Protection Instrumentation Systems,'' with its revisions

and supplements, provides appropriate limiting conditions for operation

and action statements, and is, therefore acceptable.

Date of issuance: February 16, 1996.

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

30 days.

Amendment Nos.: Unit 1-165--Unit 2-147.

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

revised the Technical Specifications.

Date of initial notice in Federal Register: March 15, 1995 (60 FR

14019).

The August 30, 1995, letter provided clarifying information that

did not change the scope of the January 13, 1995, application and the

initial proposed no significant hazards consideration determination.

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

Safety Evaluation dated February 16, 1996.

No significant hazards consideration comments received: No.

Local Public Document Room location: Atkins Library, University of

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

[[Page 7562]]

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

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

Pennsylvania

Date of application for amendments: July 10, 1995.

Brief description of amendments: These amendments modify the

Technical Specifications to minimize the potential for boron dilution

of the reactor coolant system (RCS) during startup of an isolated RCS

loop. The changes permit RCS loop isolation only during Modes 5 and 6

and require the RCS loop isolation valves be open with power removed

from their valve operators during Modes 1, 2, 3, and 4. The changes

also require isolation of primary grade water from the RCS during Modes

4, 5, and 6, except during planned boron dilution or makeup activities.

Date of issuance: February 12, 1996.

Effective date: As of date of issuance, to be implemented within 60

days.

Amendment Nos.: 195 and 78.

Facility Operating License Nos. DPR-66 and NPF-73: Amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: August 16, 1995 (60 FR

42602).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated February 12, 1996.

No significant hazards consideration comments received: No.

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

663 Franklin Avenue, Aliquippa, Pennsylvania 15001.

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

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

Date of application for amendments: July 20, 1995, as supplemented

December 4, 1995.

Brief description of amendments: These amendments revise Technical

Specification 3/4.8.1.1, ``A.C. Sources-Operating,'' to incorporate

guidance provided in NRC Generic Letter (GL) 84-15, ``Proposed Staff

Actions to Improve and Maintain Diesel Generator Reliability,'' and GL

93-05, ``Line-Item Technical Specification Improvements to Reduce

Surveillance Requirements for Testing During Power Operation,'' which

includes (1) revised requirements for testing the operable emergency

diesel generators (EDGs) for various combinations of inoperable offsite

circuits and EDGs and (2) revised surveillance requirements for the

EDGs. The revised surveillance requirements include specifying

generator voltage, frequency limits, and diesel starting time. The

amendments also make several editorial changes to TS 3/4.8.1.1 to make

TS 3/4 8.1.1 consistent with the guidance provided in the NRC's

Improved Standard Technical Specifications (NUREG-1431).

Date of issuance: February 12, 1996.

Effective date: As of date of issuance, to be implemented within 60

days.

Amendment Nos.: 196 and 79.

Facility Operating License Nos. DPR-66 and NPF-73: Amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: August 16, 1995 (60 FR

42603).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated February 12, 1996.

No significant hazards consideration comments received: No.

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

663 Franklin Avenue, Aliquippa, Pennsylvania 15001.

Florida Power and Light Company, Docket Nos. 50-250 and 50-251,

Turkey Point Plant Units 3 and 4, Dade County, Florida

Date of application for amendments: November 22, 1995.

Brief description of amendments: The amendments consist of changes

relating to removal of the TS Bases from the TS index.

Date of issuance: February 13, 1996.

Effective date: February 13, 1996.

Amendment Nos.: 182 and 176.

Facility Operating Licenses Nos. DPR-31 and DPR-41: Amendments

revised the Technical Specifications.

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

FR 65678).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated February 13, 1996.

No significant hazards consideration comments received: No.

Local Public Document Room location: Florida International

University, University Park, Miami, Florida 33199.

Florida Power Corporation, et al., Docket No. 50-302, Crystal River

Unit No. 3 Nuclear Generating Plant, Citrus County, Florida

Date of application for amendment: May 31, 1995, as supplemented

November 28, 1995, and December 21, 1995. The supplementary submittals

did not affect the staff's proposed finding of no significant hazards

consideration.

Brief description of amendment: This amendment increases the

surveillance interval on various instruments from 18 to 24 months.

Date of issuance: February 13, 1996.

Effective date: February 13, 1996.

Amendment No.: 152.

Facility Operating License No. DPR-72. Amendment revised the

Technical Specifications.

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

35070).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated February 13, 1996.

No significant hazards consideration comments received: No.

Local Public Document Room location: Coastal Region Library, 8619

W. Crystal Street, Crystal River, Florida 32629.

Georgia Power Company, Oglethorpe Power Corporation, Municipal

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

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

Burke County, Georgia

Date of application for amendments: October 16, 1995, as

supplemented by letter dated December 22, 1995.

Brief description of amendments: The amendments add a footnote to

Technical Specification 4.6.1.2.d stating the Type B and C tests

scheduled for Unit 1's refueling outage, cycle 6 (1R6) will be

conducted in accordance with Option B of 10 CFR Part 50, Appendix J

(hereafter referred to as Option B) using the guidance of Regulatory

Guide 1.163, September 1995. This change only applies to Unit 1's

refueling outage 1R6 because implementation of Option B for Type A, B,

and C testing for both units is being incorporated into the Improved

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