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

UNITED STATES 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 November 10, 1995, through November 24,

1995. The last biweekly notice was published on November 27, 1995 (60

FR 58395).

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

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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 January 5, 1996, the licensee may file a request for a hearing

with respect to issuance of the amendment to the subject facility

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

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

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

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

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

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

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

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

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

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

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

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

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

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

Secretary or the designated Atomic Safety and Licensing Board will

issue a notice of a hearing or an appropriate order.

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

shall set forth with particularity the interest of the petitioner in

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

the proceeding. The petition should specifically explain the reasons

why intervention should be permitted with particular reference to the

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

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

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

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

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

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

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

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

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

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

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

the specificity requirements described above.

Not later than 15 days prior to the first prehearing conference

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

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

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

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

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

brief explanation of the bases of the contention and a concise

statement of the alleged facts or expert opinion which support the

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

contention at the hearing. The petitioner must also provide references

to those specific sources and documents of which the petitioner is

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

facts or expert opinion. Petitioner must provide sufficient information

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

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

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

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

petitioner who fails to file such a supplement which satisfies these

requirements with respect to at least one contention will not be

permitted to participate as a party.

Those permitted to intervene become parties to the proceeding,

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

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

hearing, including the opportunity to present evidence and cross-

examine witnesses.

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

determination on the issue of no significant hazards consideration. The

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

If the final determination is that the amendment request involves

no significant hazards consideration, the Commission may issue the

amendment and make it immediately effective, notwithstanding the

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

of the amendment.

If the final determination is that the amendment request involves a

significant hazards consideration, any hearing held would take place

before the issuance of any amendment.

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

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

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

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

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

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

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

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

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

should be given Datagram Identification Number N1023 and the following

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

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

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

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

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

attorney for the licensee.

Nontimely filings of petitions for leave to intervene, amended

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

not be entertained absent a determination by the Commission, the

presiding officer or the Atomic Safety and Licensing Board that the

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

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

[[Page 62487]]

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: May 2, 1995

Description of amendments request: The proposed change revises the

large- break loss-of-coolant accident (LOCA) dose consequences. The

large-break LOCA dose calculation is being changed to include an

additional release path through allowable steam generator tube leakage

to the atmospheric dump valves (ADVs) or turbine bypass valves (TBVs).

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 probability or consequences of an accident previously

evaluated are not significantly increased by this change to the

large break LOCA dose consequences. This change has no effect on the

LOCA safety analysis for emergency core cooling system performance,

which demonstrates conformance to the acceptance criteria of 10 CFR

50.46, as described in the PVNGS Updated Final safety Analysis

Section 6.3.3. This change has no effect on structures, systems or

components prior to a LOCA or any other accident. The new

radiological consequences of the revised large break LOCA dose

calculation are below 10 CFR 100 limits for the exclusion area

boundary (EAB) and low population zone (LPZ), and the 10 CFR 50,

Appendix A, GDC 19 limits for the control room, as shown in Table 1-

1, Column C. The NRC has previously approved changes to the PVNGS

LOCA dose consequences with the acceptance criteria that the doses

are still within the guidelines set forth in 10 CFR 100 and GDC 19.

This acceptance criteria is described in the Safety Evaluation

related to amendment Nos. 64, 50, and 37 to PVNGS Units 1, 2, and 3

respectively, dated September 8, 1992.

The LOCA dose calculation is being changed to include an

additional release path through allowable steam generator tube

leakage to the ADVs or TBVs. This change is necessary to reflect a

revised calculation assumption that, following a large break LOCA,

the secondary system pressure would fall below reactor coolant

system pressure and containment pressure when operators cooldown the

steam generators by using ADVs or the TBVs (in accordance with the

safety analysis and EOPs [emergency operating procedures]). It is

desirable to use the ADVs or TBVs to vent secondary system steam and

thus reduce heat input to the reactor coolant system following a

large break LOCA. No other LOCA analysis assumptions are being

changed, and no changes are being made to structures, systems,

components or procedures.

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

or different kind of accident from any accident previously

evaluated.

This change has no impact on any structures, systems,

components, or procedures. The only impact is the revised

radiological consequences of a large break LOCA to include an

additional release path, as discussed in the response to Standard 1

above. Therefore, the proposed change does not create the

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

previously evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

This change to the large break LOCA dose consequences does not

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

radiological consequences of the revised large break LOCA dose

calculation are below 10 CFR 100 limits for the EAB and LPZ, and the

10 CFR 50, Appendix A, GDC 19 limits for the control room, as

described in the response to Standard 1 above. The NRC has

previously approved changes to the PVNGS LOCA dose consequences with

the acceptance criteria that the doses are still within the

guidelines set forth in 10 CFR 100 and GDC 19. This acceptance

criteria is described in the Safety Evaluation related to amendment

Nos. 64, 50, and 37 to PVNGS Units 1, 2, and 3 respectively, dated

September 8, 1992. No equipment qualification is affected by the new

assumption of a release path through the secondary system following

a large break LOCA, and no post LOCA radiation zones will be

changed. This change has no impact on any structures, systems,

components, or procedures.

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: November 22, 1995

Description of amendment request: The current Technical

Specifications (TS) Section 3.3.4.2 describes the limiting condition

during which components in the Service Water (SW) system may be

inoperable. The TS Section 3.3.4.2 states, in part, ``During power

operation, the requirements of 3.3.4.1 may be modified to allow any one

of the following components to be inoperable provided the remaining

systems are in continuous operation.'' The proposed change will delete

the qualifying statement,''... provided the remaining systems are in

continuous operation,'' from TS Section 3.3.4.2. Currently, this

statement requires the ``remaining systems to be in continuous

operation'' while allowing one SW loop header, or one SW pump, or one

SW booster pump to be inoperable for a period of 24 hours.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The proposed change would remove the requirement for the

remaining SW system components to be in continuous operation while

one TS-required component is inoperable. Rather, the remaining

components would remain operable, and no change would be made in

normal system operation. The SW system provides an accident

mitigation function and is not involved in accident initiation

sequences. Therefore, the proposed change would not involve a

significant increase in the probability of an accident previously

evaluated.

The capacity of the SW system is such that its accident

mitigation function can be performed by operation of a maximum of

two SW pumps, one SW booster pumps, and one SW header. While a TS-

required component is inoperable, sufficient accident mitigation

capability is provided by the remaining operable components, rather

than requiring the remaining systems to be in continuous operation.

Therefore, the proposed change would not cause 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 would remove the requirement for the

remaining SW system

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components to be in continuous operation while one TS-required

component is inoperable. Rather, the remaining components would

remain operable. The proposed change would not change the normal

operation of the system, nor would any physical modifications result

from the change. The function and capability of the SW systems would

remain unchanged. Therefore, the proposed change would not create

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

accident previously evaluated.

3. The proposed change does not involve a significant reduction

in the margin of safety.

The proposed change would remove the requirement for the

remaining SW system components to be in continuous operation while

one allowed TS-required component is inoperable. Rather, the

remaining TS-required components would remain operable. Adequate

assurance of operability is maintained by performance of regular

surveillance testing. Maintaining operable status rather than

placing equipment in continuous operation does not result in a

change in the ability of the SW system to perform its intended

function, since the system provides an automatic response to

accident conditions, and the system possesses adequate capacity to

perform its normal operating function with one allowed TS-required

component inoperable. Therefore, the proposed change does not

involve a significant reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Hartsville Memorial Library,

147 West College Avenue, Hartsville, South Carolina 29550

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

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

NRC Project Director: David B. Matthews

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

Neck Plant, Middlesex County, Connecticut

Date of amendment request: October 24, 1995

Description of amendment request: The proposed amendment will

increase the trip setpoints and allowable values for the low power

block (P-7).

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

In accordance with 10CFR50.92, CYAPCO has reviewed the proposed

change and has concluded that it does not involve a significant

hazards consideration (SHC). The basis for this conclusion is that

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

The proposed change will relax the power level values for the P-

7 interlock by 2 percent. This change affects both the P-7 and P-7N

interlocks. The P-7 interlock affects reactor trips on 1) low flow

in more than one reactor coolant loop, 2) reactor coolant pump bus

under voltage, 3) more than one reactor coolant pump breaker open,

4) main steam line isolation valve closure, 5) turbine trip, and 6)

variable low pressure. The P-7 interlock automatically blocks these

reactor trips on decreasing power and automatically unblocks these

reactor trips on increasing power. The P-7N interlock affects the

reactor trip on wide range, neutron flux, high startup rate. P-7N

automatically enables this reactor trip on decreasing power level

and automatically blocks this reactor trip on increasing power

level. The Applicable Modes requirement and Action Statements for

the P-7 interlock and the reactor trips associated with both P-7 and

P-7N in the Instrumentation Channel and Surveillance Requirements of

Technical Specification 3/4.3.1 are being changed by 2 percent to be

consistent with the change to P-7. The interlock setpoint cannot

cause an accident. Also, the proposed 2 percent increase in the

power level still results in a power level well below the power

level at which the P-7 interlocked reactor trips are required for

accident mitigation, as well as maintaining the high startup rate

trip enabled at a higher power level. This proposed power level is

consistent with the technical specification requirement prior to the

conversion to standard format technical specifications and is also

consistent with the Standard Westinghouse technical specification

value. Therefore, the proposed change can neither increase the

consequences of the design basis accident nor the probability of

occurrence of the design basis accidents.

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

from any accident previously evaluated.

The proposed change only modifies the power level for the P-7

and P-7N interlocks. The proposed setpoint is a power level at which

stable plant conditions are easier to maintain while transferring

the power supply for the reactor coolant pumps between offsite power

and the main generator. The setpoint is also well below the power

level for which the reactor protection afforded by the trips that

are bypassed by P-7 is needed. This cannot create the possibility of

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

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

The proposed change maintains the power level for the P-7

interlock below the power level for which the reactor trips that are

blocked by the P-7 interlock are required. It also raises the power

level to a value at which it is easier to maintain stable plant

conditions. This will reduce the likelihood of an automatic reactor

trip during the transferring of power for the reactor coolant pumps

between offsite power and the main generator. The proposed change

will result in the high startup rate reactor trip being enabled at a

higher power level. This is conservative since it expands the range

of coverage for the trip. Therefore, the proposed change does not

impact the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Russell Library, 123 Broad

Street, Middletown, CT 06457.

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

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

CT 06141-0270.

NRC Project Director: Phillip F. McKee

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

Neck Plant, Middlesex County, Connecticut

Date of amendment request: November 1, 1995

Description of amendment request: The proposed amendment will

modifiy Surveillance Requirement 4.6.3.2, ``Containment Isolation

Valves,'' (CIVs) to change the surveillance interval from at least once

per 18 months to at least once per refueling interval.

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:

CYAPCO has reviewed the proposed change in accordance with

10CFR50.92 and concluded that the change does not involve a

significant hazards consideration (SHC). The basis for this

conclusion is that the three criteria of 10CFR50.92(c) are not

compromised. 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 evaluated.

The proposed change to Surveillance Requirement 4.6.3.2 of the

Haddam Neck Plant Technical Specifications extends the frequency for

verifying that each CIV actuates to its required position in

response to a safety injection actuation test signal. The proposal

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

least once per

[[Page 62489]]

refueling interval (24 months + 25% as allowed by Technical

Specification 4.0.2).

The proposed change to Surveillance Requirement 4.6.3.2 does not

alter the intent or method by which the surveillance is conducted,

does not involve any physical changes to the plant, does not alter

the way any structure, system, or component functions, and does not

modify the manner in which the plant is operated.

Additional assurance of CIV operability is provided by

Surveillance Requirement 4.6.3.3. Surveillance Requirement 4.6.3.3

requires the confirmation of the mechanical operability of the CIVs

by the inservice inspection program. The proposed change does not

modify these requirements.

Equipment performance over the last four operating cycles was

evaluated to determine the impact of extending the frequency of

Surveillance Requirement 4.6.3.2. This evaluation included a review

of surveillance results, preventive maintenance records, and

corrective maintenance records. It has been concluded that the CIVs

are highly reliable, and that there is no indication that the

proposed extension could cause deterioration in valve condition or

performance.

As such, the proposed change to the frequency of Surveillance

Requirement 4.6.3.2 will not degrade the ability of the CIVs to

perform their safety function.

Based on the above, the proposed change to Surveillance

Requirement 4.6.3.2 of the Haddam Neck Plant Technical

Specifications does not involve a significant increase in the

probability or consequences of an accident previously evaluated.

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

from any previously evaluated.

The proposed change to Surveillance Requirement 4.6.3.2 of the

Haddam Neck Plant Technical Specifications extends the frequency for

verifying that each CIV actuates to its required position in

response to a safety injection actuation test signal. The proposal

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

least once per refueling interval (24 months + 25% as allowed by

Technical Specification 4.0.2).

The proposed change does not alter the intent or method by which

the surveillance is conducted, does not involve any physical changes

to the plant, does not alter the way any structure, system, or

component functions, and does not modify the manner in which the

plant is operated. As such, the proposed change in the frequency of

Surveillance Requirement 4.6.3.2 will not degrade the ability of the

CIVs to perform their safety function.

Based on the above, the proposed change to Surveillance

Requirement 4.6.3.2 of the Haddam Neck Plant Technical

Specifications will not create the possibility of a new or different

kind of accident from any previously evaluated.

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

The proposed change to Surveillance Requirement 4.6.3.2 of the

Haddam Neck Plant Technical Specifications extends the frequency for

verifying that each CIV actuates to its required position in

response to a safety injection actuation test signal. The proposal

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

least once per refueling interval (24 months + 25% as allowed by

Technical Specification Section 4.0.2).

The proposed change does not alter the intent or method by which

the surveillance is conducted, does not involve any physical changes

to the plant, does not alter the way any structure, system, or

component functions, and does not modify the manner in which the

plant is operated. As such, the proposed change in the frequency of

Surveillance Requirement 4.6.3.2 will not degrade the ability of the

CIVs to perform their safety function.

Additional assurance of the operability of the CIVs is provided

by Surveillance Requirement 4.6.3.3.

Equipment performance over the last four operating cycles was

evaluated to determine the impact of extending the frequency of

Surveillance Requirement 4.6.3.2. This evaluation included a review

of surveillance results, preventive maintenance records, and

corrective maintenance records. It has been concluded that the CIVs

are highly reliable, and that there is no indication that the

proposed extension could cause deterioration in valve condition or

performance.

Based on the above, the proposed change to Surveillance

Requirement 4.6.3.2 of the Haddam Neck Plant Technical

Specifications does not involve a significant reduction in the

margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Russell Library, 123 Broad

Street, Middletown, CT 06457.

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

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

CT 06141-0270.

NRC Project Director: Phillip F. McKee

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

Charlevoix County, Michigan

Date of amendment request: November 8, 1995, as supplemented

November 17, 1995

Description of amendment request: The proposed amendment would

remove the prescriptive Type A containment leakage test rate frequency

of 40 plus or minus 10 months and add a reference to perform

containment leakage rate tests in accordance with the criteria

specified in Appendix J of 10 CFR Part 50 as amended by approved

exemptions. In addition, the proposed amendment would revise the test

pressure for Type B and C testing to correct a typographical error.

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:

Leakage test rate frequency

1) The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

This change is administrative in nature and does not impact

plant systems, structures or components. The proposed change will

allow the facility's technical specifications to be revised to allow

containment sphere leakage testing in accordance with Appendix J to

10 CFR Part 50 as modified by approved exemptions.

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

or different kind of accident from any accident previously

evaluated.

This change is administrative in nature and does not impact

plant syst ems, structures or components. The proposed change will

allow the facility's technical specifications to be revised to allow

containment sphere leakage testing in accordance with Appendix J to

10 CFR Part 50 as modified by approved exemptions.

3)The proposed change does not involve a significant reduction

in a margin of safety.

This change is administrative in nature and does not impact

plant systems, structures or components. The underlying purpose of

Appendix J is still achieved. Appendix J states that the leakage

test requirements provide for periodic verification testing of the

leak tightness integrity of the primary reactor containment. The

appendix further states that the purpose of the tests is to assure

that leakage through the primary containment shall not exceed the

allowable leakage rate values as specified in the technical

specifications or associated bases. As stated previously, for Big

Rock Point and a large percentage of other plants, the Appendix J

Type B and C testing programs provide the most significant and

meaningful assessment of containment leak tightness. The testing

history and structural capability of the containment establish that

there is significant assurance that the extended interval between

Type A tests will not adversely impact the integrity of the

containment.

Test pressure revision

As stated in the technical specification change request, this

revision is being performed to be consistent with accident pressure,

Pa, used for Big Rock Point. 20 psig is a typographical error.

23 psig has always been used for these tests.

The proposed change does not:

1) involve a significant increase in the probability or

consequences of an accident previously evaluated.

This change is administrative in nature and does not impact

plant systems, structures or components. The proposed change will

allow the facility's technical specifications to be revised to

reflect current containment sphere leakage testing in accordance

with Appendix J to 10 CFR Part 50.

[[Page 62490]]

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

from any accident previously evaluated.

This change is administrative in nature and does not impact

plant systems, structures or components. The proposed change will

allow the facility's technical specifications to be revised to

reflect current containment sphere leakage testing in accordance

with Appendix J to 10 CFR Part 50.

3) involve a significant reduction in a margin of safety.

This change is administrative in nature and does not impact

plant systems, structures or components.

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: North Central Michigan

College, 1515 Howard Street, Petoskey, Michigan 49770

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

Company, 212 West Michigan Avenue, Jackson, Michigan 49201

NRC Project Director: Brian E. Holian, Acting

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 amendment request: October 16, 1995

Description of amendment request: Appendix J of 10 CFR Part 50,

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

Reactors,'' has recently been revised to include Option B. This option

allows the implementation of a performance based Type B and C testing

program. The proposed change will add a footnote to Technical

Specification (TS) 4.6.1.2.d stating that the Type B and C tests

scheduled for Unit 1 refueling outage Cycle 6 (1R6) will be conducted

in accordance with Option B and using the guidance of Regulatory Guide

1.163, Revision 0. This option is being incorporated into the

licensee's request to implement the improved TS. However, the improved

TS are not scheduled to become effective until after the Unit 1

refueling outage 1R6.

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 change does not involve a significant increase in

probability or consequences of an accident previously evaluated. The

proposed change does not involve a change to structures, systems, or

components which would affect the probability or consequences of an

accident previously evaluated in the Vogtle Electric Generating

Plant (VEGP) Final Safety Analysis Report (FSAR). The proposed

change only provides a mechanism within the Technical Specifications

for implementing a performance-based method of determining the

frequency for leak rate testing which has been approved by the NRC

via a revision to 10 CFR 50, Appendix J.

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

or different kind of accident from any accident previously analyzed.

The amendment will not change the design, configuration, or method

of plant operation. It only allows for the implementation of Option

B of 10 CFR 50, Appendix J for Unit 1 refueling outage 1R6 without

violating the plant Technical Specifications.

3. Operation of VEGP, Unit 1, in accordance with the proposed

change will not involve a significant reduction in the margin of

safety. The proposed change does not affect a safety limit, an LCO

[limiting condition for operation], or the way plant equipment is

operated. The NRC is aware that changes similar to this proposed

change are required in order to implement Option B of 10 CFR 50,

Appendix J. In fact, the staff indicates in Paragraph V.B. of

Appendix J that Option B or parts thereof may be adopted by a

licensee 30 days after the rule becomes effective by submitting

notification of its implementing plan and a request for revision to

Technical Specifications. Since the NRC has approved the provision

for performance-based testing and must approve this Technical

Specification[] change before the performance-based Option B can be

implemented, the margin of safety will not be significantly reduced.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Burke County Public Library,

412 Fourth Street, Waynesboro, Georgia 30830

Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders,

NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta,

Georgia 30308

NRC Project Director: Herbert N. Berkow

Gulf States Utilities Company, Cajun Electric Power Cooperative,

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

Station, Unit 1, West Feliciana Parish, Louisiana

Date of amendment request: May 30, 1995 (noticed in the Federal

Register July 5, 1995, (60 FR 35080) as supplemented by letter dated

November 20, 1995

Description of amendment request: The proposed amendment would

revise the Technical Specifications as follows:

1. The Surveillance Frequency for the drywell bypass test is

changed from 18 months to 10 years with an increased testing frequency

required if performance degrades.

2. The following changes are requested for the drywell air lock

testing: (a) the leakage rate surveillance is moved from the air lock

Limiting Condition for Operation (LCO) to the drywell LCO, (b) the

requirement for the air lock to meet a specific overall leakage limit

is deleted, (c) the Note that an inoperable air lock door does not

invalidate the previous air lock leakage test is deleted, (d) the Note

which required that the air lock leakage test at 3 psid be preceded by

pressurizing the air lock to 19.2 psid is moved to the bases, and (e)

the Surveillance Frequency for the air lock leakage test and interlock

test is changed from 18 months to 24 months.

3. The Actions Notes in the drywell air lock LCO and the drywell

isolation valve LCO that identifies that the Actions required by the

drywell LCO must be taken when the drywell bypass leakage limit is not

met is deleted.

4. The requirement for the drywell air lock seal leakage rate to

meet a specific leakage limit is deleted.

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 for River Bend Station (RBS) and Grand Gulf Nuclear

Station (GGNS), which is presented below:

I. The proposed change does not significantly increase the

probability or consequences of an accident previously evaluated.

The requested changes are either administrative changes which

clarify the format of the requirement or change the requirement to

match the design bases of the plant, a change which relocates the

requirement to the Technical Specification Bases, or a change in

surveillance interval. Each of these types of change are discussed

below:

1. The administrative changes clarify the format of the

requirement or change therequirement to match the design bases of

the plant. Clarifying administrative format of

[[Page 62491]]

the Technical Specifications does not result in any changes to the

Technical Specification requirements and, as a result, does not

involve a significant increase in the probability or consequences of

an accident previously evaluated. Also, changing the requirements of

the Technical Specifications to more closely match the design bases

of the plant will continue to assure that the plant will respond as

assumed in the accident analyses and, as a result, does not involve

a significant increase in the probability or consequences of an

accident previously evaluated.

2. The proposed changes relocate information to the Technical

Specification Bases. In the Technical Specifications Bases the

relocated information will be maintained in accordance with 10 CFR

50.59 and subject to the change control provisions in Chapter 5 of

Technical Specifications. Since any changes to the Technical

Specifications Bases will be evaluated per the requirements of 10

CFR 50.59, no increase (significant or insignificant) in the

probability or consequences of an accident previously evaluated will

be allowed. Therefore, this change does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

3. The proposed changes in frequency for the drywell bypass

leakage and drywell air lock surveillances will continue to ensure

that no paths exist through passive drywell boundary components that

would permit gross leakage from the drywell to the primary

containment air space and result in bypassing the primary

containment pressure-suppression feature beyond the design basis

limit. The Mark III primary containment system satisfies General

Design Criterion 16 of Appendix A to 10 CFR Part 50. Maximum drywell

bypass leakage was determined previously by reviewing the full range

of postulated primary system break sizes. The limiting case was a

primary system small break loss of coolant accident (LOCA) and

yielded a design allowable drywell bypass leakage rate limit of

approximately 35,000 scfm for GGNS and 46,000 scfm (the Technical

Specification limit is based on a lower limit of 40,110 scfm) for

RBS. The Technical Specifications acceptable limit for the bypass

leakage following a surveillance is less than 10% of this design

basis value. The most recent bypass leakage value was approximately

2.5% for GGNS and .91% for RBS of the design allowable leakage rate

limit for the limiting event. EOI is committed to maintaining

programmatic and oversight controls that ensure that drywell bypass

leakage remains a small fraction of the design allowable leakage

limit.

The drywell is typically exposed to essentially 0 psig during

normal plant operation and 3 psig during drywell bypass leak rate

testing. These pressures are considerably lower than the structural

integrity test pressure and are less likely to initiate a crack or

cause an existing crack to grow. Visual inspections of the

accessible drywell surfaces that have been performed since the

structural integrity tests have not revealed the presence of

additional cracking or other abnormalities. Therefore, additional

cracking of the drywell structure is not expected due to testing or

operation and, similar to the justification for the ten year 10 CFR

50 Appendix J Type A test interval, it is not considered credible

for the passive drywell structure to begin to leak sufficiently to

impact the design drywell bypass leakage limit.

The primary containment's ability to perform its safety function

is fairly insensitive to the amount of drywell leakage, thereby

providing a margin to loss of the drywell safety function that is

not normally available for safety systems. This insensitivity is

demonstrated by the extremely high limiting event design basis

allowable leakage for the drywell (e.g., 35,000 scfm for GGNS and

46,000 scfm for RBS). The limiting leakage is almost an order of

magnitude higher for other events. Additionally, an even higher

allowable leakage can be realistically accommodated by the primary

containment due to the margins in the containment design. Because of

the margins available, it will take valves in multiple penetration

flow paths leaking excessively to cause the primary containment to

fail as a result of overpressurization, the probability that drywell

isolation valve leakage will result in primary containment failure

due to excessive drywell leakage is not considered significant and

this drywell/primary containment failure mode is not considered

credible.

The proposed Technical Specification changes have no significant

impact on the GGNS Individual Plant Examination (IPE) or the RBS IPE

conducted per NRC Generic Letter 88-20. The IPEs considered

overpressurization failure of primary containment as part of the

primary containment performance assessment. Due to the magnitude of

acceptable drywell leakage and the extremely low probabilities of

achieving such leakage, primary containment failure due to

preexisting excessive drywell leakage was considered a non

significant contributor to primary containment failure. Primary

containment overpressurization failure can occur with or without

preexisting excessive drywell leakage in a severe accident. This is

due to physical phenomena associated with potentially extreme

environmental conditions inside primary containment following a

severe accident. However, the calculated frequency of such extreme

conditions is very small. The proposed changes do not impact the IPE

evaluated phenomena causing primary containment overpressurization

failure nor significantly increase the probability that the drywell

has preexisting excessive leakage and therefore would not contribute

to these accident scenarios.

For the reasons discussed above, the proposed changes do not

have any significant risk impact to accidents previously evaluated

and do not significantly increase the consequences of an accident

previously evaluated. Additionally, drywell leakage is not the

initiator of any accident evaluated; therefore, changes in the

frequency of the surveillance for drywell leakage does not increase

the probability of any accident evaluated.

Therefore, the proposed changes do not significantly increase

the probability or consequences of an accident previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

The requested changes are either administrative changes which

clarify the format of the requirement or change the requirement to

match the design bases of the plant, a change which relocates the

requirement to the Technical Specification Bases, or a change in

surveillance interval. Each of these types of change are discussed

below:

1. The administrative changes in the Technical Specification

requirements do not involve a physical alteration of the plant (no

new or different type of equipment will be installed) nor does it

change 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.

2. The proposed relocation of requirements does not involve a

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

equipment will be installed) nor does it change the methods

governing normal plant operation. The proposed change will not

impose or eliminate any requirements. Adequate control of the

information will be maintained in the Technical Specification Bases.

Thus, the change proposed does not create the possibility of a new

or different kind of accident from any accident previously

evaluated.

3. The proposed change modifies the surveillance frequency for

drywell bypass leakage and drywell air lock surveillances. The

changes only impact the test frequency and do not result in any

change in the response of the equipment to an accident. The changes

do not alter equipment design or capabilities. The changes do not

present any new or additional failure mechanisms. The drywell is

passive in nature and the surveillance will continue to verify that

its integrity has not deteriorated. Therefore, the proposed change

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

accident from any accident previously evaluated.

Therefore, the proposed changes do not create the possibility of

a new or different kind of accident from any accident previously

evaluated.

III. The proposed change does not involve a significant

reduction in a margin of safety.

The requested changes are either administrative changes which

clarify the format of the requirement or change the requirement to

match the design bases of the plant, a change which relocates the

requirement to the Technical Specification Bases, or a change in

surveillance interval. Each of these types of changes are discussed

below:

1. The administrative changes in the Technical Specification

requirements do not involve a physical alteration of the plant (no

new or different type of equipment will be installed) nor does it

change the methods governing normal plant operation. Thus, this

change does not cause a significant reduction in the margin of

safety.

[[Page 62492]]

2. The relocation of requirements will not reduce a margin of

safety because it has no impact on any safety analysis assumptions.

In addition, the requirements to be transferred from the Technical

Specifications to the Technical Specifications Bases are the same as

the existing Technical Specifications. Since any future changes to

these requirements in the Technical Specifications Bases will be

evaluated per the requirements of 10 CFR 50.59, no reduction

(significant or insignificant) in a margin of safety will be

allowed.

3. The proposed change modifies the surveillance frequency for

drywell bypass leakage and associated air lock surveillances.

Reliability of drywell integrity is evidenced by the measured

leakage rate during past drywell bypass leakage surveillances.

Appropriate design basis assumptions will be upheld, even when

combined with the complementary bypass leakage surveillances as

proposed. Drywell integrity will continue to be tested by means of

the proposed periodic drywell bypass leakage test, performance of

the drywell air lock door latching and interlock mechanism

surveillance, and performance of additional surveillances including

excercising of drywell isolation valves. The combination of these

surveillances will provide adequate assurance that drywell bypass

leakage will not exceed the design basis limit. Margins of safety

would not be reduced unless leakage rates exceeded the design

allowable drywell bypass leakage limit. Therefore, the proposed

change does not cause a significant reduction in the margin of

safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Government Documents

Department, Louisiana State University, Baton Rouge, LA 70803

Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn,

1400 L Street, N.W., Washington, D.C. 20005

NRC Project Director: William D. Beckner

Gulf States Utilities Company, Cajun Electric Power Cooperative,

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

Station, Unit 1, West Feliciana Parish, Louisiana

Date of amendment request: October 26, 1995

Description of amendment request: The proposed amendment would

revise the technical specifications for sixteen editorial changes and

would delete the requirement for a program to prevent and detect

Asiatic Clams (Corbicula) in the service water system (SWS). The

editorial changes covers such things as removing systems or components

that do not exist in the River Bend Station, correcting typographical

errors, correcting to be consistent with the writers guide for Improved

Technical Specifications, adding descriptions for systems to make them

clear, and wording changes to be consistent with approved facility

operations. The Corbicula program is no longer needed because the

facility has been modified and SWS no longer takes water from the

Mississippi River; source of the larvae and infestation.

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:

EDITORIAL CHANGES

The purposed changes involves reformatting, renumbering and

rewording of the existing Technical Specifications. The

reformatting, renumbering and rewording process involves no

technical changes to existing Technical Specifications. As such,

these changes are administrative in nature and do not impact

initiators of 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 evaluated.

The proposed changes do not involve a physical alteration of the

plant (no new or different type of equipment will be installed) or

changes in methods governing normal plant operation. The proposed

changes will not impose or eliminate any new or different

requirements. Thus, these changes do not create the possibility of a

new or different kind of accident from any accident previously

evaluated.

The proposed changes will not reduce a margin of safety because

they have no impact on any safety analysis assumptions. These

changes are administrative in nature. As such, no question of safety

is involved, and the changes do not involve a significant reduction

in a margin of safety.

CORBICULA PROGRAM

The proposed change deletes the program associated with the

prevention and detection of Asiatic Clams (Corbicula) based upon

improvements to the non-safety related Normal Service Water System

(SWS). The source of makeup water to the SWS is no longer the

Mississippi River, which is the source of Asiatic Clams.

Demineralized water or well water is used eliminating the source of

asiatic clams. To prevent biofouling SWS is treated with chlorine/

bromine. This program is not considered as an initiator for any

previously evaluated accident. Therefore, the proposed change will

not increase the probability or consequences of any accident

previously evaluated.

The proposed change introduces no new mode of plant operation

and it does not involve a physical modification to the plant. The

possibility of the SES becoming contaminated by any other means is

highly unlikely since it is a ``closed-loop'' system. Therefore it

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

accident from any accident previously evaluated.

Prevention of Asiatic Clam infestation in the SWS and associated

safety-related equipment is ensured by the ``closed-loop'' design of

the SWS. Post Refuel Outage (RF-4) inspections of the safety-related

heat exchangers that interface with the ``closed-loop'' SWS have

shown no evidence of clam infestations. Therefore, 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: Government Documents

Department, Louisiana State University, Baton Rouge, Louisiana 70803

Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn,

1400 L Street, N.W., Washington, D.C. 20005

NRC Project Director: William D. Beckner

Gulf States Utilities Company, Cajun Electric Power Cooperative,

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

Station, Unit 1, West Feliciana Parish, Louisiana

Date of amendment request: November 20, 1995

Description of amendment request: The proposed amendment would

revise the technical specifications to eliminate the response time

testing requirements for selected Reactor Protection System

Instrumentation.

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 purpose of the proposed Technical Specification (TS) change

is to eliminate response time testing requirements for selected

components in the Reactor Protection System (RPS). The Boiling Water

reactors Owners' Group (BWROG) has completed an evaluating which

demonstrates that response time testing is redundant to the other

TS-required testing. These other tests, in conjunction with actions

taken in response to NRC Bulletin 90-01, ``Loss of Fill-Oil in

Transmitters Manufactured by Rosemount,'' and Supplement 1, are

sufficient to identify failure modes or degradation in instrument

response times and ensure operation of the associated systems within

acceptable limits. There are no known failure modes that can be

detected by response time testing that cannot also be detected by

the other TS-required testing. This evaluation was

[[Page 62493]]

documented in NEDO-32291, ``System Analyses for Elimination of Selected

Response Time Testing Requirements,'' January 1994. Entergy

Operations, Inc. (EOI) has confirmed the applicability of this

evaluation to River Bend Station (RBS). In addition EOI will

complete the actions identified in the NRC staff's safety evaluation

of NEDO-32291.

Because of the continued application of other existing TS-

required tests such as channel calibration, channel checks, channel

functional tests, and logic system functional tests, the response

time of these systems will be maintained within the acceptance

limits assumed in plant safety analyses and required for successful

mitigation of an initiating event. The proposed changes do not

affect the capability of the associated systems to perform their

intended function within their required response time, nor do the

proposed changes themselves affect the operation of any equipment.

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

involve a significant increase in the probability or the

consequences of an accident previously evaluated.

The proposed changes only apply to the testing requirements for

the components identified above and do not result in any physical

change to these or other components or their operation. As a result,

no new failure modes are introduced. Therefore, the proposed changes

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

from any accidents previously evaluated.

The current TS-required response times are based on the maximum

allowable values as assumed in the plant safety analyses. These

analyses conservatively establish the margin of safety. As described

above, the proposed changes do not affect the capability of the

associated systems to perform their intended function within the

allowed response time used as the basis for the plant safety

analyses. The potential failure modes for the components within the

scope of this request were evaluated for impact on instrument

response time. This evaluation confirmed that, with the exception of

loss of fill-oil of Rosemount transmitters, the remaining TS-

required testing is sufficient to identify failure modes or

degradation in instrument response times and ensure operation of the

instrument within the scope of this request is within acceptable

limits. The actions taken in response to NRC Bulletin 90-09 and

Supplement 1 are adequate to identify loss of fill-oil failures of

Rosemount transmitters. As a result, it has been concluded that

plant and systems response to an initiating event will remain in

compliance with the assumptions of the safety analysis.

Further, although not explicitly evaluated, the proposed changes

will provide an improvement to plant safety and operation by

reducing the time safety systems are unavailable, reducing the

potential for safety system actuations, reducing plant shutdown

risk, limiting radiation exposure to plant personnel, and

eliminating the diversion of key personnel resources to conduct

unnecessary testing. Therefore, EOI has concluded that this request

will result in an overall increase in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Government Documents

Department, Louisiana State University, Baton Rouge, LA 70803

Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn,

1400 L Street, N.W., Washington, D.C. 20005

NRC Project Director: William D. Beckner

North Atlantic Energy Service Corporation, Docket No. 50-443,

Seabrook Station, Unit No. 1, Rockingham County, New Hampshire

Date of amendment request: September 22, 1995

Description of amendment request: The proposed amendment would

modify a requirement of the Seabrook Station, Unit No. 1 Technical

Specifications. Specifically, the proposed amendment would change the

ACTION referenced in Table 3.3-3, Engineered Safety Features Actuation

System Instrumentation, for Functional Unit 8.b, Automatic Switchover

to Containment Sump/RWST Level Low-Low. The ACTION requirement would be

changed to ACTION 15 from ACTION 18. ACTION 15 requires an inoperable

channel to be placed in bypass (with no time limit specified) while

ACTION 18 requires an inoperable channel to be placed in the tripped

condition within 6 hours.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration. 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.

A. The change does not involve a significant increase in the

probability or consequences of an accident previously evaluated (10

CFR 50.92(c)(1)) because the proposed change would result in an

inoperable Functional Unit 8.b. protective channel being placed in

the bypassed condition vice tripped condition. Functional Unit 8.b.

is not involved in any accident initiation sequence; therefore, the

probability of a previously-analyzed accident is not increased.

Placing an inoperable Functional Unit 8.b. in bypass vice trip

reduces the probability of premature opening of the containment

building sump isolation valves thereby reducing the potential for

increasing the consequences of a previously-analyzed accident. Thus,

the consequences of a previously-analyzed accident is not increased.

B. The change does not create the possibility of a new or

different kind of accident from any accident previously evaluated

(10 CFR 50.92(c)(2)) because the change does not reduce the minimum

required number of channels of instrumentation to be operable. The

change does not alter the function of or affect the failure modes of

Functional Unit 8.b. instrumentation channels. The proposed change

does not otherwise affect the manner by which the facility is

operated, and it does not involve any changes to equipment or

features which affect the operational characteristics of the

facility.

C. The change does not involve a significant reduction in a

margin of safety (10 CFR 50.92(c)(3)) because the change does not

reduce the minimum required number of channels of instrumentation to

be operable, and it does not involve any changes to equipment or

features which affect the operational characteristics of the

facility. Therefore, the protection previously provided remains

unchanged.

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: Exeter Public Library,

Founders Park, Exeter, NH 03833.

Attorney for licensee: Lillian M. Cuoco, Esquire, Northeast

Utilities Service Company, Post Office Box 270, Hartford CT 06141-0270.

NRC Project Director: Phillip F. McKee

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

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

Connecticut

Date of amendment request: May 26, 1995, supplemented and revised

October 20, 1995.

Description of amendment request: The proposed changes would modify

TS 3.8.1.1., ``Electrical Power Systems, A.C. Sources, Operating,'' TS

3.8.1.2, ``Electrical Power Systems, Shutdown,'' TS 3.8.2.2,

``Electrical Power Systems, A.C. Distribution - Shutdown,'' and TS

3.8.2.4, ``Electrical Power Systems, D.C. Distribution - Shutdown,'' to

provide operational flexibility as well as consistency between action

statements and to eliminate certain surveillance requirements that are

not applicable in Modes 5 or 6.

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:

[[Page 62494]]

In accordance with 10 CFR 50.92, NNECO has reviewed the proposed

changes and has concluded that they do not involve an SHC. The basis

for this conclusion is that the three criteria of 10 CFR 50.92(c)

are not compromised. The proposed changes do not involve an SHC

because the change would not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed change to Surveillance Requirement 4.8.1.1.1 is

being made because presently, the surveillance requirement for

demonstrating offsite sources are operable states that ``two''

independent circuits are required. The surveillance requirement is

referenced for both operating and shutdown modes. While it is

accurate for operating modes, it is inconsistent with the limiting

condition for operation for shutdown. The proposed change is safe

because it renders the surveillance requirement consistent with the

applicable limiting condition for operation (i.e., operating or

shutdown) and eliminates a potential source of confusion.

The change to Surveillance Requirement 4.8.1.2 and Technical

Specification 3.8.2.2 merely clarifies the diesel generator

surveillance and operability requirements for Modes 5 and 6 and

renders action statements for related technical specification

sections consistent with and appropriate for operational Modes 5 and

6.

Regarding diesel generator surveillance requirements, automatic

A.C. power for LNP events in Modes 5 and 6 is not required. This is

validated by the fact that the undervoltage sensors are only

required to be operable in Modes 1, 2 and 3 to meet technical

specifications. Because the undervoltage sensors provide the logic

that results in actuation of the sequencer, it follows that the

sequencer need not be operable in Modes 5 and 6. Accordingly, the

sequencer is not required to support operability of the available

diesel generator in Modes 5 and 6. Further, because SIAS is blocked

in Modes 5 and 6, automatic start of the diesel generator upon

receipt of a SIAS is similarly not required to support operability

of the diesel generator in Modes 5 and 6.

Additionally, operation of the diesel generator in parallel with

the system during Modes 5 and 6 is not required to perform its

intended safety function. In fact, such operation may compromise

both sources as the result of a single event.

Since automatic A.C. power is not credited in the mitigation of

Mode 5 and 6 events and accidents, such as fuel handling accidents,

there is no increase in the probability or consequences of

previously evaluated accidents.

The action statement in Technical Specification 3.8.2.2 has been

revised to cite actions that are more appropriate for Modes 5 and 6

for Millstone Unit No. 2. This is due to the ability to maintain the

plant in a safe condition without needing to automatically load the

diesel generator through the sequencers in Modes 5 and 6. In

addition, the proposed change is consistent with the CE Owner's

Group Standard Technical Specification and with other Millstone Unit

No. 2 action statements. Consequently, there is no increase in the

probability or consequences of previously evaluated accidents.

The change to TS 3.8.2.4 merely renders the action statement

consistent with, and appropriate for, operational Modes 5 and 6.

Since D.C. power is not credited in the mitigation of Mode 5 and

6 events and accidents, such as fuel handling accidents, there is no

increase in the probability or consequences of previously evaluated

accidents.

The action statement in TS 3.8.2.4 has been revised to cite

actions that are more appropriate for Modes 5 and 6 for Millstone

Unit No. 2. This is due to the ability to maintain the plant in a

safe condition without D.C. power distribution available in Modes 5

and 6. In addition, the proposed change is consistent with the CE

Owner's Group Standard Technical

Specifications (NUREG-1432) and with other Millstone Unit No. 2

action statements. Consequently, there is no increase in the

probability or consequences of previously evaluated accidents.

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

from any previously evaluated.

The proposed changes do not alter or affect the design,

function, failure mode, or operation of the plant. Therefore, the

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

kind of accident from any previously evaluated.

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

The proposed changes to the technical specifications provides

greater consistency between the action statements and clarifies

which surveillance requirements are required in Modes 5 and 6. Since

the diesel generators are not required to be loaded automatically in

Modes 5 and 6, and since it is part of our shutdown risk management

program to assure that adequate cooling is able to be provided, and

since the diesel will still be verified to start and achieve rated

speed, the proposed changes to the technical specifications do not

reduce the margin of safety.

The proposed change to the TS provides greater consistency among

action statements during Modes 5 and 6. Since the D.C. distribution

system is not credited in the mitigation of Mode 5 and 6 events and

accidents, and since it is part of our shutdown risk management

program to assure that adequate fuel cooling is able to be provided,

the proposed change to the TS does not reduce the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: 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: June 27, 1995, as supplemented July 21,

1995

Description of amendment request: The amendment revises the

Technical Specifications (TS) to relocate TS requirements for the

containment purge exhaust and supply valves, and to remove a duplicate

testing requirement for the safety injection input from engineered

safety features from 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:

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

hazards consideration] because the changes would not:

1. Involve a significant increase in the probability or

consequences of an accident previously analyzed.

The first proposed change relocates the operability and

surveillance requirements for the containment high range radiation

monitors from Technical Specification Section 3.3.3 to Technical

Specification Section 3.3.2. The proposed changes are administrative

in nature. The proposed changes do not alter the way any structure,

system, or component functions and do not modify the manner in which

the plant is operated and do not involve any physical changes to the

plant.

The second proposed modification will delete the testing

requirement for functional unit 16, ``Safety Injection Input from

ESF,'' of Table 4.3-1 because the logic circuitry that processes

the safety injection signals and produces a reactor trip is

tested under functional unit 19 ``Automatic Trip and Interlock

Logic,'' and the testing is performed on a more frequent basis

(i.e., on a monthly staggered bases versus on an 18-month

frequency). In addition, the same logic testing is accomplished with

an 18-month TADOT of functional unit 1.a of Table 4.3-2 and with a

monthly staggered actuation logic testing of functional unit 16 of

Table 4.3-2. This testing ensures that operability of the logic

under functional unit 16 of Table 4.3-1 is verified. The other tests

will continue to verify the operability of the reactor trip system

and that a reactor trip will be initiated when required.

Therefore, there is no change in the potential for an increase

in the consequences of an accident previously analyzed.

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

from any previously analyzed.

[[Page 62495]]

The proposed changes do not affect the operation or response of

any plant equipment or introduce any new failure mechanisms. The

proposed elimination of the testing requirement line item does not

affect the test results since the logic circuitry that processes the

safety injection signal and produces a reactor trip will be tested

and is tested under functional unit 19 of Table 4.3-1. As such, the

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

accident previously evaluated.

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

The proposed changes do not have any adverse impact on the

protective boundaries nor do they affect the consequences of any

accident analyzed. The operability and surveillance requirements,

although relocated to other technical specifications, will still

ensure that the system (the radiation monitors) is tested and within

limits. The proposed elimination of the testing equipment will not

change the performance or operating conditions of the safety

systems. The operable reactor trip system instrumentation ensures

that the assumptions in the Bases of the Technical Specifications

are not affected and ensures that the margin of safety is not

reduced. Therefore, the proposed changes do not reduce the margin of

safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: 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

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

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

Obispo County, California

Date of amendment requests: November 14, 1994

Description of amendment requests: The proposed amendment would

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

Nuclear Power Plant, Unit Nos. 1 and 2, for the slave relay test

frequency from quarterly (Q) to refueling (R). The request would also

remove table notation 4 from Table 4.3-2. The associated Bases would

also be appropriately revised.

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 results of WCAPs 14117 and 13878 demonstrate that slave

relays are highly reliable. The WCAPs also provide guidance to

assure that slave relays remain highly reliable. The aging

assessment concludes that the age/temperature-related degradation of

all ND relays, and NE relays produced after May 1990, is

sufficiently slow such that a refueling frequency surveillance

interval will not significantly increase the probability of slave

relay failures. Finally, the evaluation of the interposing slave

relays in the emergency diesel generator start circuitry, control

room ventilation and auxiliary building ventilation realignments,

steam generator blowdown isolation and radwaste isolation systems

has concluded that based on the tests of the interposing relays

performed during other equipment testing, reasonable assurance is

provided that failures will be identified if the associated slave

relays are tested on a refueling frequency.

The removal of table notation 4 from TS Table 4.3-2 is an

administrative change that eliminates unnecessary redundancy from

the TS and does not affect plant operation.

Therefore, the proposed changes do not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

The proposed changes do not alter the performance of the ESFAS

mitigation systems assumed in the plant safety analysis. Changing

the interval for periodically verifying ESFAS slave relays (assuring

equipment operability) will not create any new accident initiators

or scenarios.

Therefore, the proposed changes do not create the possibility of

a new or different kind of accident from any accident previously

evaluated for DCPP.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The proposed changes do not affect the total ESFAS response

assumed in the safety analysis since the reliability of the slave

relays will not be significantly affected by the increased

surveillance frequency.

Therefore, the proposed changes do not involve a significant

reduction in a margin of safety.

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

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

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

amendment requests involve no significant hazards consideration.

Local Public Document Room location: California Polytechnic State

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

Department, San Luis Obispo, California 93407

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

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

NRC Project Director: William H. Bateman

South Carolina Electric & Gas Company, South Carolina Public

Service Authority, Docket No. 50-395, Virgil C. Summer Nuclear

Station, Unit No. 1, Fairfield County, South Carolina

Date of amendment request: August 18, 1995, as supplemented on

November 1, 1995

Description of amendment request: The proposed amendment would

revise the Operating License and Technical Specifications to allow for

a power uprate to 2900 MWt. The current maximum power level is 2775

MWt.

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 or consequences of an accident previously

evaluated is not significantly increased.

Implementation of uprate power operation does not contribute to

any accident evaluated in the FSAR [Final Safety Analysis Report].

The NSSS [Nuclear Steam Supply System] Components (RV [reactor

vessel], RCPs [reactor coolant pumps], CRDMs [control rod drive

mechanisms], SGs [steam generators], and piping) are compatible with

the revised operating conditions. These components have been

reanalyzed and the results show that ASME [American Society of

Mechanical Engineers] Code requirements remain satisfied and are

within the current Licensing Basis.

Interfacing Systems which are important to safety are not

adversely impacted and will continue to perform their design

function. Overall secondary plant performance is not significantly

altered by the proposed changes.

The revision to the Pressure Temperature Limits will not

adversely impact the RCS [reactor coolant system] Pressure Boundary.

The length of time these curves will be applicable, due to increased

neutron fluence, is being reduced. Before the 13 Effective Full

Power Years have elapsed, new curves will be generated to reflect

the analysis of the specimen capsule and will be derived utilizing

NRC approved methodology.

Therefore, since the Reactor Coolant pressure boundary integrity

and system functions are not adversely impacted, the probability of

occurrence of an accident evaluated in the VCSNS [Virgil C. Summer

Nuclear Station] FSAR will be no greater than the original design

basis of the plant.

An extensive analysis has been performed to evaluate the

consequences of the following accident types currently evaluated in

the VCSNS FSAR:

[[Page 62496]]

- Non-LOCA [loss-of-coolant accident] Events

- Large Break and Small Break LOCA

- Steam Generator Tube Rupture

With the [delta]75 SGs and revised operating conditions, the

calculated results (i.e., DNBR [departure from nucleate boiling

ratio], Primary and Secondary System Pressure, Peak Clad

Temperature, Metal Water Reaction, Challenge to Long Term Cooling,

Environmental Conditions Inside and Outside containment, etc.) for

the accidents are similar to those currently reported in the VCSNS

FSAR and remain within applicable Regulatory Acceptance Criteria.

Select results (i.e., Containment Pressure during a Steam Line

Break, Minimum DNBR for Rod Withdrawal from Subcritical, etc.) are

slightly more limiting than those currently reported in the FSAR due

to the use of the assumed operating conditions with the [delta]75

SGs and in some cases, use of an uprated core power of 2900 MWt.

However, in all cases, the calculated results do not challenge the

integrity of the primary/secondary/ containment pressure boundary

and remain within the regulatory acceptance criteria applied to

VCSNS's current licensing basis.

Given that calculated radiological consequences are not

significantly higher than current FSAR results and remain well

within 10 CFR 100 limits, it is concluded that the consequences of

an accident previously evaluated in the FSAR are not significantly

increased.

2. The proposed license amendment does not create the

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

previously evaluated.

Uprate power operation will not introduce any new accident

initiator mechanisms. Structural integrity of the RCS is maintained

during all plant conditions through compliance with the ASME code

and 10 CFR 50 Appendix G requirements. Design requirements of

auxiliary systems are met with the RSGs [replacement steam

generators] and uprate power operation. No new failure modes or

limiting single failures have been identified. Since the safety and

design requirements continue to be met and the integrity of the

reactor coolant system pressure boundary is not challenged, no new

accident scenarios have been created. Therefore, the types of

accidents defined in the FSAR continue to represent the credible

spectrum of events to be analyzed which determine safe plant

operation.

3. The proposed license amendment does not involve a significant

reduction in a margin of safety.

Although uprate power operation will require changes to the

VCSNS Technical Specifications, the proposed changes are supported

by extensive LOCA, NON-LOCA and SGTR [steam generator tube rupture]

analyses. These analyses show acceptable consequences with margin to

the applicable regulatory limits. All equipment required to function

during accident conditions has been shown to remain qualified and

thus will perform their design function, and all components remain

in compliance with the codes and standards in effect when VCSNS was

originally licensed (with the exception of the replacement steam

generators which use the 1986 ASME Code Section III Edition).

Low Temperature Overpressure transients which could challenge

RCS structural integrity are not impacted by the revision to the

Pressure Temperature Limitations Curves. The curves are not directly

impacted, the changes do not reduce any margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Fairfield County Library, 300

Washington Street, Winnsboro, SC 29180

Attorney for licensee: Randolph R. Mahan, South Carolina Electric &

Gas Company, Post Office Box 764, Columbia, South Carolina 29218

NRC Project Director: Frederick J. Hebdon

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 amendment request: August 29, 1995

Description of amendment request: The proposed amendment would

revise the Technical Specifications for allowable values and trip

setpoints for selected plant process instrumentation. The new allowable

values/setpoints are in accordance with the instrument setpoint

methodology accepted by the NRC staff in a letter dated July 18, 1995.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The proposed revised Trip Setpoints and Allowable Values are

more conservative than those currently approved in the Technical

Specifications. Therefore, any proposed system or component

actuations will occur earlier, resulting in a more conservative

plant response. Thus, the proposed change does not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

The proposed change to the Technical Specifications does not

introduce any new components nor does it modify the design of any

existing components. Other than making Trip Setpoints and Allowable

Values of existing instrumentation more conservative, the change

does not affect the design or function of any plant system,

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

are operated. Thus, the possibility of a new or different kind of

accident previously evaluated is not created.

3. The proposed change does not result in a significant

reduction in the margin of safety.

Since the proposed revised Trip Setpoints and Allowable Values

are more conservative than the existing values, the margin of safety

would be increased by issuance of the changes. Thus, the proposed

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

safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Perry Public Library, 3753

Main Street, Perry, Ohio 44081

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

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

NRC Project Director: Gail H. Marcus

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 amendment request: November 2, 1995

Description of amendment request: The proposed amendment would

revise the Technical Specifications to allow 120 volt AC buses EV-1-A

and EV-1-B to be energized from either their normal inverter power

supply or from their alternate power supply.

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:

These buses are not used as the initiator of any analyzed

accidents. Therefore, the probability of any previously evaluated

accident has not increased. If an accident were to occur while the

buses are supplied from the alternate power supply, there would

[[Page 62497]]

be no change in the analyzed accident scenario since even in the event

of a loss of offsite power event, the safety functions would be

completed. Thus, the consequences of any previously evaluated

accident have not increased.

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

or different kind of accident from any previously evaluated:

The proposed change introduces no new mode of plant operation

and it does not involve physical modification to the plant.

Therefore, it 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:

This change does not involve a significant reduction in a margin

of safety since the proposed change maintains a safety related,

diesel-backed power supply to these buses whether the power is

supplied from the inverters or from the alternate power supply. If a

loss of offsite power event were to occur while the buses were

supplied from the alternate power source, the safety functions being

performed by components supplied from these buses would occur. Thus,

there has been no reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Perry Public Library, 3753

Main Street, Perry, Ohio 44081

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

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

NRC Project Director: Gail H. Marcus

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 amendment request: November 2, 1995

Description of amendment request: The proposed amendment to the

Perry Nuclear Power Plant Technical Specifications revises those

specifications associated with handling irradiated fuel in Primary

Containment and the Fuel Handling Building, and selected specifications

associated with CORE ALTERATIONS. Specifically, analysis identifies

that only recently irradiated fuel contains sufficient fission

products to require OPERABILITY of accident mitigation features to meet

the accident analysis assumptions. Analyses also show that accident

mitigation features such as building INTEGRITY and engineered safety

feature (ESF) ventilation systems are not required for CORE ALTERATION

events.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The proposed requirements are imposed during specific

activities which can be postulated to result in significant

radioactive releases. The proposed APPLICABILITY requirements are

consistent with either the original design basis analyses or with

revised analyses performed to support this proposed amendment.

Because the equipment controlled by the revised Specifications is

not considered an initiator to any previously analyzed accident,

inoperability of the equipment cannot increase the probability of

any previously evaluated accident.

Consistent with the original design basis analysis, the reanalysis

concludes that radiological consequences of the fuel handling accident

are well within the 10 CFR 100.11 limits, as defined by acceptance

criteria in Standard Review Plan Section 15.7.4. The reanalysis has

previously been submitted to the Nuclear Regulatory Commission for

review, and NRC confirmatory calculations reached consistent results

(reference NRC Safety Evaluation for License Amendment No. 35). The

results of the CORE ALTERATION events other than the fuel handling

accident remain unchanged from the original design basis, which showed

that these events do not result in fuel cladding integrity damage or

radioactive releases. Therefore, the proposed changes do not

significantly increase the consequences of any previously evaluated

accident.

Based on the above, the proposed changes do not significantly

increase the probability or consequences of any accident previously

evaluated.

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

or different kind of accident from any accident previously

evaluated.

The proposed requirements are imposed when specific activities

represent situations where significant radioactive releases can be

postulated. The proposed APPLICABILITY requirements are consistent

with design basis analyses. The proposed changes do not introduce

any new modes of plant operation and do not involve physical

modifications to the plant. Therefore, the proposed changes do not

create the possibility of a new or different kind of accidident from

any previously evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The proposed change imposes controls to ensure that during

performance of activities which represent situations where

radioactive releases are postulated, the radiological consequences

are at or below the established licensing limit. Safety margins and

analytical conservatisms have been evaluated and are well

understood. Substantial conservatism is retained to ensure that the

analysis adequately bounds all postulated event scenarios. The

current margin of safety is retained.

Specifically, the margin of safety for the fuel handling

accident is the difference between the 10 CFR 100 limits and the

licensing limit defined by the Standard Review Plan (NUREG 0800),

Section 15.7.4. The licensing limit is defined by the Standard

Review Plan as being well within the 10 CFR 100 limits, with

``well within'' defined as 25% of the 10 CFR 100 limits for the fuel

handling accident. Excess margin is the difference between the

postulated doses and the corresponding licensing limit. In the NRCs

initial licensing review of the Perry Nuclear Power Plant (NUREG-

0887, Section 15.3.3), the NRC accepted the design and analyses

based on the results of the analyses being well within the guideline

values of 10 CFR 100.

The proposed APPLICABILITY requirements continue to ensure that

the whole-body and thyroid doses at the exclusion area and low

population zone boundaries as well as control room doses are at or

below the corresponding licensing limit. The margin of safety is

unchanged; therefore, the proposed changes do not involve a

significant reduction in a margin of safety.

The margin of safety for the CORE ALTERATION events other than

the fuel handling accident discussed above also remains the same as

in the original design basis analyses, since the proposed changes do

not impact on the Technical Specification requirements for systems

needed to prevent or mitigate such CORE ALTERATION events.

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: Perry Public Library, 3753

Main Street, Perry, Ohio 44081

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

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

NRC Project Director: Gail H. Marcus

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

[[Page 62498]]

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.

Commonwealth Edison Company, Docket Nos. 50-295 and 50-304, Zion

Nuclear Power Station, Units 1 and 2, Lake County, Illinois

Date of application for amendments: October 6, 1995, and

supplemented November 20, 1995

Brief description of amendments: The amendments revise the

Technical Specifications by incorporating a new acceptance criterion

for steam generator tubes with degradation in the tubesheet roll

expansion region.

Date of issuance: November 21, 1995

Effective date: November 21, 1995

Amendment Nos.: 172 and 159

Facility Operating License Nos. DPR-39 and DPR-48: The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: October 16, 1995 (60 FR

53648) The supplemental letter provided clarifying information that did

not affect the initial proposed no significant hazards consideration

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

contained in a Safety Evaluation dated November 21, 1995.

No significant hazards consideration comments received: No

Local Public Document Room location: Waukegan Public Library, 128

N. County Street, Waukegan, Illinois 60085

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

Neck Plant, Middlesex County, Connecticut

Date of application for amendment: August 10, 1995

Brief description of amendment: The amendment revises the Haddam

Neck Technical Specification Section 3/4.4.3, ``Pressurizer,'' to add a

footnote to allow the pressurizer level to be controlled, outside of

the programmed level, between 25 to 50 percent, plus or minus 5 percent

in Mode 3 when the reactor coolant system is borated to the required

Mode 5 concentrations.

Date of Issuance: November 14, 1995

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

within 60 days.

Amendment No.: 186

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

Technical Specifications.

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

52928) The Commission's related evaluation of this amendment is

contained in a Safety Evaluation dated November 14, 1995.

No significant hazards consideration comments received: No

Local Public Document Room location: Russell Library, 123 Broad

Street, Middletown, CT 06457

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: September 13, 1995, as

supplemented October 16,1995

Brief description of amendments: These amendments revise the

Administrative Controls section of the BVPS-1 and BVPS-2 TSs to make

them consistent with the requirements of the Offsite Dose Calculation

Manual (ODCM). The ODCM was recently updated to reflect the radioactive

liquid and gaseous effluent release limits and the liquid holdup tank

activity limit of BVPS-1 License Amendment No. 188 and BVPS-2 License

Amendment No. 70 which were issued June 12, 1995.

Date of issuance: November 21, 1995

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

within 10 days.

Amendment Nos.: 194 and 77

Facility Operating License Nos. DPR-66 and NPF-73: Amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: September 22, 1995 (60

FR 49292) The October 16, 1995, letter did not change the initial

proposed no significant hazards consideration determination or expand

the amendment request beyond the scope of the September 22, 1995,

Federal Register notice. The Commission's related evaluation of the

amendments is contained in a Safety Evaluation dated November 21, 1995.

No significant hazards consideration comments received: No

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

663 Franklin Avenue, Aliquippa, Pennsylvania 15001

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

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

Date of amendment request: February 14, 1994, as supplemented by

letters dated July 25, August 15, and August 29, 1995

Brief description of amendment: The amendment changes the Appendix

A Technical Specifications (TSs) to make them consistent with the

revised 10 CFR Part 20, Standards for Protection Against Radiation.

Date of issuance: November 17, 1995

Effective date: November 17, 1995

Amendment No.: 116

Facility Operating License No. NPF-38. Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: March 30, 1994 (59 FR

14888) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated November 17, 1995. The July 25,

August 15, and August 29, 1995 letters provided clarifying information

that did not change the initial propose no significance hazards

consideration determination.

No significant hazards consideration comments received: No

Local Public Document Room location: University of New Orleans

Library, Louisiana Collection, Lakefront, New Orleans, LA 70122

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: May 12, 1995, as supplemented

by letters dated July 6 and October 2, 1995.

[[Page 62499]]

Brief description of amendments: The amendments revise Technical

Specification Surveillance Requirement 4.6.1.2 to add the provision

that 10 CFR Part 50, Appendix J, applies, except as modified by NRC-

approved exemptions.

Date of issuance: November 17, 1995

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

30 days

Amendment Nos.: 91 and 69

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

revised the Technical Specifications.

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

35078) The July 6 and October 2, 1995, letters provided clarifying

information that did not change the scope of the May 12, 1995,

application and initial proposed no significant hazards consideration

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

contained in a Safety Evaluation dated November 17, 1995.

No significant hazards consideration comments received: No

Local Public Document Room location: Burke County Library, 412

Fourth Street, Waynesboro, Georgia 30830

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

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

Date of application for amendment: July 28, 1995, as supplemented

September 12, October 18, and October 31, 1995.

Brief description of amendment: In order to support a full-core

offload as a normal end-of-cycle event, the amendment adds License

Condition 2.C(6) and will require that: (1) the reactor be subcritical

for at least 100 hours prior to the start of reactor refueling

operations, (2) the spent fuel pool bulk temperature be maintained less

than or equal to 140F, and (3) two trains of shutdown cooling be

operable during reactor refueling operations.

Date of issuance: November 9, 1995

Effective date: As of the date of issuance.

Amendment No.: 89

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

license.

Date of initial notice in Federal Register: August 30, 1995 (60 FR

45180) The September 12, October 18, and October 31, 1995, submittals

provided additional information that did not change the initial

proposed no significant hazards consideration determination. The

Commission's related evaluation of the amendment and Final No

Significant Hazards Consideration Determination are contained in a

Safety Evaluation dated November 9, 1995.

No significant hazards consideration comments received: No public

comments received. A request for a hearing was received from We the

People, the Seacoast Anti-Pollution League, the New England Coalition

on Nuclear Pollution, and Donald Del Core of Uncasville, Connecticut.

Local Public Document Room location: Learning Resources Center,

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

Norwich, CT 06360

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

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

Connecticut

Date of application for amendment: October 6, 1995, supplemented

October 23, November 2, and November 15, 1995.

Brief description of amendment: The amendment adds footnotes to

Action Statement (AS) 3.8.1.1.a of the Technical Specification (TS) and

its bases to allow a one-time extension of the allowed outage time

(AOT) for an inoperable offsite power source from the current 72 hours

to 7 days.

Date of issuance: November 22, 1995

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

within 30 days.

Amendment No.: 192

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

Technical Specifications.

Date of initial notice in Federal Register: October 17, 1995 (60 FR

53812). The October 23, November 2, and November 15, 1995, letters

provided clarifying information and slight modifications to the

original request that were not outside the scope of the original notice

and did not change the initial proposed no significant hazards

consideration determination. The Commission's related evaluation of the

amendment is contained in a Safety Evaluation dated November 22, 1995.

No significant hazards consideration comments received: No

Local Public Document Room location: Learning Resources Center,

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

Norwich, CT 06360

Northern States Power Company, Docket No. 50-282, Prairie Island

Nuclear Generating Plant, Unit No. 1, Goodhue County, Minnesota

Date of application for amendment: January 10, 1995, as

supplemented August 9 and September 20, 1995.

Brief description of amendment: The amendments revise the Prairie

Island event monitoring instrumentation Technical Specifications and

associated Bases to conform to Standard Technical Specifications for

post-accident monitoring.

Date of issuance: November 9, 1995

Effective date: November 9, 1995, with full implementation within

30 days.

Amendment Nos.: 121/114

Facility Operating License No. DPR-42 and DPR-60. Amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: February 15, 1995 (60

FR 8753) The August 9 and September 20, 1995, letters provided updated

Technical Specification pages and clarifying information in response to

discussions with the staff during various teleconferences conducted

during the review process. This information was within the scope of the

original application and did not change the staff's initial proposed no

significant hazards consideration determination. The Commission's

related evaluation of the amendment is contained in a Safety Evaluation

dated November 9, 1995.

No Significant hazards consideration comments received: No

Local Public Document Room location: Minneapolis Public Library,

Technology and Science Department, 300 Nicollet Mall, Minneapolis,

Minnesota 55401

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

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

Pennsylvania

Date of application for amendments: December 2, 1994, as

supplemented May 12, 1995.

Brief description of amendments: These amendments relocate the fire

protection requirements from the Technical Specifications to the

Updated Final Safety Analysis Report in accordance with the guidance in

Generic Letter (GL) 86-10, ``Implementation of Fire Protection

Requirements,'' and GL 88-12, ``Removal of Fire Protection Requirements

from Technical Specifications.''

Date of issuance: November 20, 1995 Effective date: As of date of

issuance, both units, to be implemented within 30 days.

Amendment Nos.: 104 and 68

Facility Operating License Nos. NPF-39 and NPF-85. The amendments

revised the Technical Specifications and the License.

Date of initial notice in Federal Register: April 26, 1995 (60 FR

20524) The supplemental letter provided clarifying information and did

not

[[Page 62500]]

change the initial proposed no significant hazards consideration

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

contained in a Safety Evaluation dated November 20, 1995.

No significant hazards consideration comments received: No

Local Public Document Room location: Pottstown Public Library, 500

High Street, Pottstown, Pennsylvania 19464

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

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

Pennsylvania

Date of application for amendments: September 14, 1995 and

supplemented by letter dated October 27, 1995

Brief description of amendments: These amendments revise the

technical specifications by deleting Reactor Enclosure and Refueling

Area Secondary Containment Isolation Valve Tables 3.6.5.2.1-1 and

3.6.5.2.2-1, and references to them, in accordance with Generic Letter

91-08, ``Removal of Component lists from Technical Specifications.''

The TS have been modified to state requirements in general terms that

include the components listed in the tables removed from the TS.

Date of issuance: November 20, 1995

Effective date: As of date of issuance, to be implemented within 30

days.

Amendment Nos.: November 20, 1995

Facility Operating License Nos. NPF-39 and NPF-85. The amendments

revised the Technical Specifications.

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

52934) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated November 20, 1995.

No significant hazards consideration comments received: No

Local Public Document Room location: Pottstown Public Library, 500

High Street, Pottstown, Pennsylvania 19464

Tennessee Valley Authority, Docket No. 50-296, Browns Ferry Nuclear

Plant, Unit 3, Limestone County, Alabama

Date of application for amendments: October 4, 1995 (TS 368)

Brief description of amendment: The amendment delete requirements

for daily checks for certain instruments that do not have indications,

and provides editorial changes.

Date of issuance: November 13, 1995

Effective Date: November 13, 1995

Amendment No.: 202

Facility Operating License No. DPR-68: Amendment revised the

Technical Specifications.

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

52935) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated November 13, 1995.

No significant hazards consideration comments received: None

Local Public Document Room location: Athens Public library, South

Street, Athens, Alabama 35611

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

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

Date of application for amendments: August 7, 1995 (TS 95-03)

Brief description of amendments: The amendments address operation

with a rod urgent failure condition, including limited operation with

one control or shutdown bank inserted up to 18 steps below its

insertion point. In addition, the surveillance interval for rod

movement verifications has been increased from 31 to 92 days.

Date of issuance: November 21, 1995

Effective date: November 21, 1995

Amendment Nos.: 215 and 205

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

revise the technical specifications.

Date of initial notice in Federal Register: August 30, 1995 (60 FR

45186) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated November 21, 1995.

No significant hazards consideration comments received: None

Local Public Document Room location: Chattanooga-Hamilton County

Library, 1101 Broad Street, Chattanooga, Tennessee 37402

The Cleveland Electric Illuminating Company, Centerior Service

Company, Duquesne Light Company, Ohio Edison Company, Pennsylvania

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

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

Date of application for amendment: April 28, 1995

Brief description of amendment: The amendment removes the license

conditions for the Transamerica Delaval, Inc. emergency diesel

generators specified by paragraph 2.C.(9) and defined in Attachment 2

to the Operating License.

Date of issuance: November 16, 1995

Effective date: November 16, 1995

Amendment No.: 74

Facility Operating License No. NPF-58: This amendment revises the

license.

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

29889) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated November 16, 1995.

No significant hazards consideration comments received: No

Local Public Document Room location: Perry Public Library, 3753

Main Street, Perry, Ohio 44081

Toledo Edison Company, Centerior Service Company, and The Cleveland

Electric Illuminating Company, Docket No. 50-346, Davis-Besse

Nuclear Power Station, Unit No. 1, Ottawa County, Ohio

Date of application for amendment: June 23, 1995, and facsimile

transmission dated October 31, 1995

Brief description of amendment: This amendment relocates TS 3/

4.3.3.3, ``Seismic Instrumentation;'' TS 3/4.3.3.4, ``Meteorological

Instrumentation;'' and TS 3/4.4.11, ``Reactor Coolant System Vents;''

and the Bases for each of the three sections from the TS to the Updated

Safety Analysis Report, and eliminates the special reporting

requirements for inoperable seismic and meteorological monitoring

instrumentation from TS 6.9.2.

Date of issuance: November 14, 1995 Effective date: November 14,

1995, and shall be implemented not later than 90 days after issuance.

Amendment No.: 201

Facility Operating License No. NPF-3: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: August 2, 1995 (60 FR

39455) The October 31, 1995, facsimile transmission was clarifying in

nature and did not affect the initial no significant hazards

consideration determination. The Commission's related evaluation of the

amendment is contained in a Safety Evaluation dated November 14, 1995.

No significant hazards consideration comments received: No

Local Public Document Room location: University of Toledo, William

Carlson Library, Government Documents Collection, 2801 West Bancroft

Avenue, Toledo, Ohio 43606

Toledo Edison Company, Centerior Service Company, and The Cleveland

Electric Illuminating Company, Docket No. 50-346, Davis-Besse

Nuclear Power Station, Unit No. 1, Ottawa County, Ohio

Date of application for amendment: June 7, 1995

Brief description of amendment: This amendment revises Technical

Specification 3/4.9.4, Refueling Operations - Containment Penetrations;

[[Page 62501]]

Bases 3/4.9.4, Containment Penetrations; and Limiting Condition for

Operation (LCO) 3.9.4.b to allow both doors of the containment

personnel airlock to be open during core alterations or movement of

irradiated fuel within the containment, provided that certain specified

conditions are meet. Additional changes revise or clarify TS LCO

3.9.4.c, TS Action 3.9.4.a, and TS Surveillance Requirement 4.9.4, and

modify the associated Bases.

Date of issuance: November 17, 1995

Effective date: November 17, 1995

Amendment No.: 202

Facility Operating License No. NPF-3. Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: August 2, 1995 (60 FR

39454) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated November 17, 1995.

No significant hazards consideration comments received: No

Local Public Document Room location: University of Toledo, William

Carlson Library, Government Documents Collection, 2801 West Bancroft

Avenue, Toledo, Ohio 43606

TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak

Steam Electric Station, Unit Nos. 1 and 2, Somervell County, Texas

Date of amendment request: December 6, 1994

Brief description of amendments: These changes revise Technical

Specifications to allow appropriate remedial action for high

particulate levels in the diesel generator fuel oil inventory and other

out-of-limit properties in new diesel generator fuel oil that has been

added to the existing diesel generator fuel oil storage inventory.

Date of issuance: November 17, 1995

Effective date: November 17, 1995

Amendment Nos.: Unit 1 - Amendment No. 43; Unit 2 - Amendment No.

29

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

revised the Technical Specifications.

Date of initial notice in Federal Register: February 1, 1995 (60 FR

6311) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated November 17, 1995.

No significant hazards consideration comments received: No

Local Public Document Room location: University of Texas at

Arlington Library, Government Publications/Maps, 702 College, P.O. Box

19497, Arlington, Texas 76019.

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

County, Missouri

Date of amendment request: January 13, 1995

Brief description of amendment: The amendment revises Technical

Specifications (TS) 3.3.1 and 3.3.2 to relocate Tables 3.3-2 and 3.3-5,

which provide the response time limits for the reactor trip system and

the engineered safety features actuation system instruments, from the

TS to the updated Final Safety Analysis Report (FSAR). The amendment

also relocates the Bases discussion for TS 3.3.1 and TS 3.3.2 to

Section 16.3 of the updated FSAR.

Date of issuance: November 22, 1995

Effective date: November 22, 1995, to be implemented within 30 days

of issuance.

Amendment No.: 104

Facility Operating License No. NPF-30. The amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: February 15, 1995 (60

FR 8741) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated November 22, 1995.

No significant hazards consideration comments received: No

Local Public Document Room locations: Callaway County Public

Library, 710 Court Street, Fulton, Missouri 65251

Washington Public Power Supply System, Docket No. 50-397, Nuclear

Project No. 2, Benton County, Washington

Date of application for amendment: June 6, 1995

Brief description of amendment: The amendment modifies the Index of

the WNP-2 Technical Specifications by deleting reference to the Bases

pages.

Date of issuance: November 24, 1995

Effective date: November 24, 1995

Amendment No.: 143

Facility Operating License No. NPF-21: The amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: July 19, 1995 (60 FR

37102) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated November 24, 1995.

No significant hazards consideration comments received: No

Local Public Document Room location: Richland Public Library, 955

Northgate Street, Richland, Washington 99352

Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301,

Point Beach Nuclear Plant, Unit Nos. 1 and 2, Town of Two Creeks,

Manitowoc County, Wisconsin

Date of application for amendments: September 13, 1995, and October

19, 1995, as supplemented by letter dated October 25, 1995

Brief description of amendments: These amendments revise Technical

Specification (TS) Section 15.1, ``Definitions,'' TS Section 15.3.1.G,

``Operational Limitations'' (and basis), and TS Figure 15.2.1-2,

``Reactor Core Safety Limits, Point Beach Unit 2.'' The changes reduce

the reactor coolant system raw measured total flow rate limit and

reflect new reactor core safety limits for Unit 2.

Date of issuance: November 17, 1995

Effective date: November 17, 1995

Amendment Nos.: 165 and 169

Facility Operating License Nos. DPR-24 and DPR-27: Amendments

revised the Technical Specifications. Public comments requested as to

proposed no significant hazards consideration: Yes (60 FR 54527 dated

October 24, 1995). That notice provided an opportunity to submit

comments on the Commission's proposed no significant hazards

consideration determination. No comments have been received. The notice

also provided for an opportunity to request a hearing by November 24,

1995, but indicated that if the Commission makes a final no significant

hazards consideration determination any such hearing would take place

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

the amendment, finding of exigent circumstances, and final

determination of no significant hazards consideration is contained in a

Safety Evaluation dated November 17, 1995.

No significant hazards consideration comments received: No

Local Public Document Room location: Joseph P. Mann Library, 1516

Sixteenth Street, Two Rivers, Wisconsin 54241

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

Creek Generating Station, Coffey County, Kansas

Date of amendment request: September 14, 1995

Brief description of amendment: The amendment revised Technical

Specification 3/4.5.5 to increase the allowed outage time for

adjustment of boron concentration for the refueling water storage tank

from 1 hour to 8 hours.

Date of issuance: November 13, 1995

[[Page 62502]]

Effective date: November 13, 1995, to be implemented within 30 days

of issuance.

Amendment No.: 91

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

Technical Specifications.

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

52936) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated November 13, 1995.

No significant hazards consideration comments received: No

Local Public Document Room locations: Emporia State University,

William Allen White Library, 1200 Commercial Street, Emporia, Kansas

66801 and Washburn University School of Law Library, Topeka, Kansas

66621

Notice Of Issuance Of Amendments To Facility Operating Licenses And

Final Determination Of No Significant Hazards Consideration And

Opportunity For A Hearing (exigent public announcement or emergency

circumstances)

During the period since publication of the last biweekly notice,

the Commission has issued the following amendments. The Commission has

determined for each of these amendments that the application for the

amendment complies with the standards and requirements of the Atomic

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

and regulations. The Commission has made appropriate findings as

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

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

Because of exigent or emergency circumstances associated with the

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

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

Consideration of Issuance of Amendment, Proposed No Significant Hazards

Consideration Determination, and Opportunity for a Hearing.

For exigent circumstances, the Commission has either issued a

Federal Register notice providing opportunity for public comment or has

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

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

the Commission's proposed determination of no significant hazards

consideration. The Commission has provided a reasonable opportunity for

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

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

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

transcribed as appropriate and the licensee has been informed of the

public comments.

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

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

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

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

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

significant hazards consideration determination. In such case, the

license amendment has been issued without opportunity for comment. If

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

Commission may provide an opportunity for public comment. If comments

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

been consulted by telephone whenever possible.

Under its regulations, the Commission may issue and make an

amendment immediately effective, notwithstanding the pendency before it

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

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

significant hazards consideration is involved.

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

made a final determination that the amendment involves no significant

hazards consideration. The basis for this determination is contained in

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

been issued and made effective as indicated.

Unless otherwise indicated, the Commission has determined that

these amendments satisfy the criteria for categorical exclusion in

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

no environmental impact statement or environmental assessment need be

prepared for these amendments. If the Commission has prepared an

environmental assessment under the special circumstances provision in

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

it is so indicated.

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

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

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

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

available for public inspection at the Commission's Public Document

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

the local public document room for the particular facility involved.

The Commission is also offering an opportunity for a hearing with

respect to the issuance of the amendment. By January 5, 1996, the

licensee may file a request for a hearing with respect to issuance of

the amendment to the subject facility operating license and any person

whose interest may be affected by this proceeding and who wishes to

participate as a party in the proceeding must file a written request

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

hearing and a petition for leave to intervene shall be filed in

accordance with the Commission's ``Rules of Practice for Domestic

Licensing Proceedings'' in 10 CFR Part 2. Interested persons should

consult a current copy of 10 CFR 2.714 which is available at the

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

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

particular facility involved. If a request for a hearing or petition

for leave to intervene is filed by the above date, the Commission or an

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

the Chairman of the Atomic Safety and Licensing Board Panel, will rule

on the request and/or petition; and the Secretary or the designated

Atomic Safety and Licensing Board will issue a notice of a hearing or

an appropriate order.

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

shall set forth with particularity the interest of the petitioner in

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

the proceeding. The petition should specifically explain the reasons

why intervention should be permitted with particular reference to the

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

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

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

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

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

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

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

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

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

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

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

the specificity requirements described above.

[[Page 62503]]

Not later than 15 days prior to the first prehearing conference

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

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

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

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

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

brief explanation of the bases of the contention and a concise

statement of the alleged facts or expert opinion which support the

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

contention at the hearing. The petitioner must also provide references

to those specific sources and documents of which the petitioner is

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

facts or expert opinion. Petitioner must provide sufficient information

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

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

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

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

petitioner who fails to file such a supplement which satisfies these

requirements with respect to at least one contention will not be

permitted to participate as a party.

Those permitted to intervene become parties to the proceeding,

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

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

hearing, including the opportunity to present evidence and cross-

examine witnesses. Since the Commission has made a final determination

that the amendment involves no significant hazards consideration, if a

hearing is requested, it will not stay the effectiveness of the

amendment. Any hearing held would take place while the amendment is in

effect.

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

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

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

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

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

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

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

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

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

should be given Datagram Identification Number N1023 and the following

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

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

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

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

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

attorney for the licensee.

Nontimely filings of petitions for leave to intervene, amended

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

not be entertained absent a determination by the Commission, the

presiding officer or the Atomic Safety and Licensing Board that the

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

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

Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301,

Point Beach Nuclear Plant, Unit Nos. 1 and 2, Town of Two Creeks,

Manitowoc County, Wisconsin

Date of application for amendments: November 9, 1995, as

supplemented by letters dated November 13, 1995, and November 16, 1995

Brief description of amendments: These amendments revise Technical

Specification (TS) Section 15.4.2, ``In-Service Inspection of Safety

Class Components,'' to incorporate a new steam generator tube

acceptance criterion for the Unit 2 steam generators. This criterion

allows tubes that are degraded or defective in a location (within the

tubesheet) that does not affect the structural integrity of the tube to

remain in service. The applicable basis is also changed.

Date of issuance: November 22, 1995

Effective date: November 22, 1995

Amendment Nos.: 166 and 170

Facility Operating License Nos. DPR-24 and DPR-27. Amendments

revised the Technical Specifications. Public comments requested as to

proposed no significant hazards consideration: No The Commission's

related evaluation of the amendments, finding of emergency

circumstances, and final determination of no significant hazards

consideration are contained in a Safety Evaluation dated November 22,

1995.

No significant hazards consideration comments received: No

Local Public Document Room location: Joseph P. Mann Library, 1516

Sixteenth Street, Two Rivers, Wisconsin 54241.

Attorney for licensee: Ernest L. Blake, Jr., Shaw, Pittman, Potts &

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

NRC Project Director: Gail H. Marcus

Dated at Rockville, Maryland, this 29th day of November 1995.

For the Nuclear Regulatory Commission

Elinor G. Adensam,

Deputy Director, Division of Reactor Projects - III/IV, Office of

Nuclear Reactor Regulation

[Doc. 95-29540 Filed 12-5-95; 8:45 am]

BILLING CODE 7590-01-F