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

[[Page 494]]

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 December 12, 1994, through December 21,

1994. The last biweekly notice was published on December 21, 1994 (59

FR 65809).

Notice Of Consideration Of Issuance Of Amendments To Facility

Operating Licenses, Proposed No Significant Hazards Consideration

Determination, And Opportunity For A Hearing

The Commission has made a proposed determination that the following

amendment requests involve no significant hazards consideration. Under

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

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

involve a significant increase in the probability or consequences of an

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

or different kind of accident from any accident previously evaluated;

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

for this proposed determination for each amendment request is shown

below.

The Commission is seeking public comments on this proposed

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

publication of this notice will be considered in making any final

determination.

Normally, the Commission will not issue the amendment until the

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

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

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

the Commission may issue the license amendment before the expiration of

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

the amendment involves no significant hazards consideration. The final

determination will consider all public and State comments received

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

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

opportunity for a hearing after issuance. The Commission expects that

the need to take this action will occur very infrequently.

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

Directives Branch, Division of Freedom of Information and Publications

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

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

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

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

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

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

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

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

intervene is discussed below.

By February 3, 1994, 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 [[Page 495]] contention will

not be permitted to participate as a party.

Those permitted to intervene become parties to the proceeding,

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

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

hearing, including the opportunity to present evidence and cross-

examine witnesses.

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

determination on the issue of no significant hazards consideration. The

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

If the final determination is that the amendment request involves

no significant hazards consideration, the Commission may issue the

amendment and make it immediately effective, notwithstanding the

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

of the amendment.

If the final determination is that the amendment request involves a

significant hazards consideration, any hearing held would take place

before the issuance of any amendment.

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

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

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

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

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

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

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

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

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

should be given Datagram Identification Number N1023 and the following

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

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

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

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

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

attorney for the licensee.

Nontimely filings of petitions for leave to intervene, amended

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

not be entertained absent a determination by the Commission, the

presiding officer or the Atomic Safety and Licensing Board that the

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

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

For further details with respect to this action, see the

application for amendment which is available for public inspection at

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

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

the particular facility involved.

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

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

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

Date of amendment requests: November 30, 1994

Description of amendment requests: The proposed amendment would

relocate Table 3.3-2, ``Reactor Protective Instrumentation Response

Times,'' and Table 3.3-5, ``Engineered Safety Features Response

Times,'' of Technical Specifications (TS) 3/4.3.1 and 3/4.3.2,

respectively, to the Palo Verde Updated Final Safety Analysis Report

(UFSAR) in accordance with the guidance provided in Generic Letter (GL)

93-08. In addition, the proposed amendment would make administrative

changes to two previous TS amendment requests to reflect the deletion

of Tables 3.3-2 and 3.3-5. The amendment would also delete an obsolete

footnote on page 3/4 3-17 of the Palo Verde Unit 2's TS.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensees have

provided their analysis about the issue of no significant hazards

consideration, which is presented below:

Standard 1 -- Does the proposed change involve a significant

increase in the probability or consequences of an accident

previously evaluated?

The proposed change relocates two tables of instrument response

time limits from the TS to the UFSAR. The changes are in accordance

with the guidance provided by the NRC in Generic Letter 93-08. The

changes are administrative in nature and do not involve any

modifications to plant equipment or affect plant operation.

Therefore, the proposed change does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

Standard 2 - Does the proposed change create the possibility of

a new or different kind of accident from any accident previously

evaluated?

The proposed change relocates two tables of instrument response

time limits from the TS to the UFSAR. The changes are in accordance

with the guidance provided by the NRC in Generic Letter 93-08. The

changes are administrative in nature, do not involve any

modifications to plant equipment and cause no change in the method

by which any safety-related system performs its function. Therefore,

the proposed change does not create the possibility of a new or

different kind of accident from any accident previously evaluated.

Standard 3 - Does the proposed change involve a significant

reduction in a margin of safety?

The proposed change relocates two tables of instrument response

time limits from the TS to the UFSAR. The changes are in accordance

with the guidance provided by the NRC in Generic Letter 93-08. The

changes are administrative in nature, do not change or alter

regulatory requirements and do not affect the safety analysis. Plant

procedures contain response time testing acceptance criteria that

reflect the reactor trip and ESFAS [engineered safety feature

actuation system] response time limits in the tables being relocated

from the TS into the UFSAR. Therefore, the proposed change does not

involve a significant reduction in a margin of safety.

The NRC staff has reviewed the licensees' 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

amendment requests involve no significant hazards consideration.

Local Public Document Room location: Phoenix Public Library, 12

East McDowell Road, Phoenix, Arizona 85004

Attorney for licensees: 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: Theodore R. Quay

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

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

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

Date of amendment requests: December 7, 1994

Description of amendment requests: The proposed amendment would

change Table 4.3-1 of Technical Specification 3/4.3.1 to allow

verification of the shape annealing matrix elements used in the Core

Protection Calculators. This would provide the option to use generic

shape annealing matrix elements in the Core Protection Calculators.

Presently, cycle-specific shape annealing elements are determined

during startup testing after each core reload. Use of a generic shape

annealing matrix would eliminate approximately 2 to 3 hours of critical

path work during startup after a refueling outage.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensees have

provided their analysis [[Page 496]] about the issue of no significant

hazards consideration, which is presented below:

Standard 1 -- Does the proposed change involve a significant

increase in the probability or consequences of an accident

previously evaluated?

The proposed Technical Specification amendment does not involve

a significant increase in the probability or consequences of an

accident previously evaluated. The Technical Specification amendment

provides the option to use generic shape annealing matrix elements

in the Core Protection Calculators. The design basis of the Core

Protection Calculators is to provide the DNBR [departure from

nucleate boiling ratio] and linear heat rate trip functions for the

Reactor Protection System so that the Specified Acceptable Fuel

Design Limits on DNBR and fuel centerline melt are not exceeded

during normal operation or Anticipated Operational Occurrences, and

assist the Engineered Safety Features Actuation System in limiting

the consequences of postulated accidents. The generic shape

annealing matrix elements will be validated during startup testing

and will meet the same acceptance criteria as the cycle specific

shape annealing matrix elements. If the generic shape annealing

matrix elements are not valid, cycle specific shape annealing matrix

elements would be used in the Core Protection Calculators. This

change will not affect the Core Protection Calculators capability to

protect the plant by tripping the reactor, based on a conservative

calculation of minimum DNBR and peak linear heat rate, to ensure

that the Specified Acceptable Fuel Design Limits are not violated in

the event of an Anticipated Operational Occurrence. Therefore, the

generic shape annealing matrix elements will not affect the safety

analysis, since there is no change to the design basis of the Core

Protection Calculator System.

Standard 2 -- Does the proposed change create the possibility of

a new or different kind of accident from any accident previously

evaluated?

The proposed Technical Specification amendment does not create

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

accident previously evaluated. Since the generic shape annealing

matrix elements will still have to meet the same acceptance criteria

as the cycle specific shape annealing matrix elements, the Core

Protection Calculators will still generate axial power shapes that

fall within the required uncertainties. The Core Protection

Calculators will still trip the reactor, based on a conservative

calculation of minimum DNBR and peak linear heat rate, to ensure

that the Specified Acceptable Fuel Design Limits are not violated in

the event of an Anticipated Operational Occurrence.

Standard 3 -- Does the proposed change involve a significant

reduction in a margin of safety?

The proposed Technical Specification amendment will not involve

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

in the margin of safety, since the generic shape annealing matrix

elements will still have to meet the same acceptance criteria as the

cycle specific shape annealing matrix elements. Therefore, this

change will not affect the design basis of the Core Protection

Calculators. The Core Protection Calculators will still provide a

reactor trip based on a conservative calculation of minimum DNBR and

peak linear heat rate.

The NRC staff has reviewed the licensees' 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

amendment requests involve no significant hazards consideration.

Local Public Document Room location: Phoenix Public Library, 12

East McDowell Road, Phoenix, Arizona 85004

Attorney for licensees: 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: Theodore R. Quay

Arizona Public Service Company, et al., Docket No. STN 50-529, Palo

Verde Nuclear Generating Station, Unit No. 2, Maricopa County,

Arizona

Date of amendment request: November 30, 1994

Description of amendment request: The proposed amendment would

change the pressurizer code safety valve lift setting from 2500 psia to

2475 psia. The lift setting is being changed to permit Unit 2 to

operate with up to 1500 plugged tubes in each steam generator.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensees have

provided their analysis about the issue of no significant hazards

consideration, which is presented below:

Standard 1 -- Does the proposed change involve a significant

increase in the probability or consequence of an accident previously

evaluated?

The proposed Technical Specification amendment does not involve

a significant increase in the probability or consequences of an

accident previously evaluated. Chapters 6 and 15 of the [Palo Verde

Nuclear Generating Station] PVNGS [Updated Final Safety Analysis

Report] UFSAR have been reviewed to address the impact of these

changes (1500 plugged tubes and a pressurizer code safety valve

nominal lift setpoint of 2475 psia) on accident consequences. For

most of the events that were previously analyzed in the UFSAR, the

proposed change does not have a significant affect or adversely

impact the accident analysis. For RCS [reactor coolant system]

pressure peaking events, Loss of Condenser Vacuum (LOCV) and

Feedwater Line Breaks (FLB), a new analysis was performed to justify

the acceptability of the changes.

For the LOCV event (anticipated operational occurrence), the

reanalysis determined that the peak RCS pressure, assuming 1500

plugged tubes and a pressurizer code safety valve nominal lift

setpoint of 2475 psia, is 2728 psia. The maximum reactor coolant

system (RCS) pressure reached for this event as described in UFSAR

Section 15.2.3 is 2742 psia. Therefore, this change is bounded by

the reference cycle (UFSAR analysis) and remains below the 110%

(2750 psia) design pressure limit.

Several FLB scenarios are analyzed in support of PVNGS Unit 2

operation. The scenario with the highest system pressures is the

large FLB with a loss of alternating current (LOAC). For the large

FLB with a LOAC event (limiting fault event), assuming 1500 plugged

tubes and a pressurizer code safety valve nominal lift setpoint of

2475 psia, is 2813 psia. The maximum RCS pressure reached for this

event as described in UFSAR Section 15.2.8 is 2843 psia. The

analysis shows that the RCS peak pressure for the large FLB with a

LOAC (very low probability) event remains below the required value

of 120% (3000 psia) of design pressure. Therefore, the analyses and

reviews of the RCS pressure peaking events determined that the UFSAR

design pressure limit is still bounding with this change. That is,

the RCS design pressure limit will not be exceeded. Also, safety

valves are accident mitigating devices and do not contribute to the

probability of an event.

Standard 2 -- Does the proposed change create the possibility of

a new or different kind of accident from any accident previously

evaluated?

The proposed Technical Specification amendment does not create

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

accident previously evaluated. The analyses and reviews show that

the current licensing basis remains valid for this change (UFSAR

design pressure limit is still bounding with this change). Safety

valves are accident mitigating devices and do not contribute to the

possibility of an accident. The pressurizer code safety valves are

not manually or remotely operated, but are designed to automatically

open to provide overpressure protection for pressure peaking events.

The change in the pressurizer code safety valve setpoint to 2475

psia does not significantly increase the probability of a

pressurizer code safety valve opening, since the pressure is still

well above the Technical Specification Table 2.2-1 reactor trip

setpoint of 2383 psia for high pressurizer pressure.

Standard 3 -- Does the proposed change involve a significant

reduction in a margin of safety?

The proposed Technical Specification amendment does not involve

a significant reduction in a margin of safety. The analyses and

reviews show that the limits in the licensing and design basis are

still valid with this change. The analyses show that the RCS peak

pressure remains below the 110% (2750 psia) design pressure limit

for the LOCV event and remains below the required value of 120%

(3000 psia) of design pressure RCS peak pressure for the large FLB

with a LOAC (very low probability) event. The analyses

[[Page 497]] and reviews of the RCS pressure peaking events

determined that the UFSAR design pressure limit is still bounding

with this change. Therefore, the proposed Technical Specification

amendment maintains the margin of safety to the design pressure

limit.

The NRC staff has reviewed the licensees' 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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Phoenix Public Library, 12

East McDowell Road, Phoenix, Arizona 85004

Attorney for licensees: 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: Theodore R. Quay

Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-

324, Brunswick Steam Electric Plant, Units 1 and 2, Brunswick

County,North Carolina

Date of amendments request: November 16, 1994Description of

amendments request: The proposed revision to the Technical

Specifications (TS) would change the Technical Specification 3/4.6.2 to

remove the specific instrumentation requirements for monitoring of the

suppression chamber average water temperature. Also, the proposed

revision would change the TS Bases 3/4.6.2 to indicate the methods that

are acceptable for determining suppression chamber average water

temperature.Proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

The proposed change maintains the same number of monitored

locations from which an average suppression chamber water

temperature can be derived, while making available additional valid

RTD [resistance temperature detector] inputs from what was the

redundant channel. No safety-related equipment, safety function or

plant operation will be altered as a result of the proposed change.

The SPTMS [suppression chamber temperature monitoring system] is

neither an accident initiator nor does it provide any automatic

accident mitigation function. The change does not affect the design,

materials, or construction standards applicable to the suppression

chamber average water temperature monitoring instrumentation.

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

new or different kind of accident from any accident previously

evaluated.

The fundamental function and objective of the system is not

affected by the proposed change. As stated above, no safety-related

equipment, safety function or plant operations will be altered as a

result of the proposed change. The change does not affect the

design, materials, or construction standards applicable to the

suppression chamber average water temperature instrumentation.

3. The proposed amendment does not involve a significant

reduction in a margin of safety.

The proposed change allows the substitution of a qualified RTD

already installed at a monitored location to insure the suppression

chamber average water temperature remains valid. It does not involve

any changes to the plant design or operation, therefore, no margins

of safety, as defined by the plant's accident analyses, are

impacted. Deletion of the defined instrument channels will not

affect the ability to verify the suppression chamber ``average''

water temperature is being maintained below the maximum average

temperatures required by the specification. This will insure the

suppression chamber is Operable and able to perform its intended

safety function.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: University of North Carolina

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

Wilmington, North Carolina 28403-3297.

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

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

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: December 12, 1994

Description of amendment request: The requested change would revise

the containment spray (CS) nozzle surveillance interval from 5 to 10

years.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The requested change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The requested change extends the surveillance interval for

performance of qualitative flow testing of the CS nozzles. A

revision to this surveillance interval can in no way increase the

probability of any accident previously evaluated.

Containment spray nozzle testing is not intended to track

degradation of equipment by monitoring or trending performance.

Rather, this surveillance constitutes a test of the passive design

of the spray nozzles, i.e., it merely demonstrates whether the

nozzles are or are not blocked or clogged. Based upon industry and

plant-specific operating experience, a single failure rendering a

significant number of nozzles inoperable as a result of blockage is

considered highly unlikely. Since the reliability or functioning of

the spray nozzles will not be affected by the revised surveillance

interval, the consequences of any accident previously evaluated will

not be increased. The requested change does not affect the physical

design or operation of the plant, does not alter assumptions

contained within the Updated Final Safety Analysis Report, and will

not affect other Technical Specifications that preserve safety

analysis assumptions. Therefore, operation of the facility in

accordance with the requested change will not involve a significant

increase in the consequences of any accident previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

The requested change extends the surveillance interval for

performance of qualitative flow testing of the CS nozzles. This

change in the spray nozzle surveillance interval will not change or

affect the physical plant or the modes of plant operation defined

within the facility Operating License. This change does not involve

the addition or modification of plant equipment, nor does it alter

the design or operation of plant systems. Therefore, operation of

the facility in accordance with the requested change will not create

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

accident previously evaluated.

3. The requested change does not involve a significant reduction

in the margin of safety.

The requested change extends the surveillance for performance of

qualitative flow testing of the CS nozzles. This revised

surveillance interval will not change or otherwise influence the

degree of operability assumed for the CS system within the plant

safety analyses. As demonstrated by plant-specific and industry

experience, an operational failure of the containment spray nozzles

is considered highly unlikely. Since prior testing has demonstrated

proper functioning of the CS spray nozzles, and operational single-

failures are considered highly unlikely, a reduction in testing

frequency should not affect the ability of the CS system to mitigate

the affects of a large loss-of-coolant or steam release accident.

[[Page 498]] Therefore, operation of the facility in accordance with

the requested change will not result in a significant reduction in

the margin or 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: William H. Bateman

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

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

Date of amendment request: October 24, 1994

Description of amendment request: The proposed amendments would

restructure the primary containment integrity and primary containment

leakage technical specifications (TS) to reduce the repetition of those

requirements contained in NRC regulations such as Appendix J to 10 CFR

50. The amendments also support proposed exemptions from Appendix J

requirements related to the scheduling of containment integrated leak

rate tests (CILRT). In addition to the restructuring and scheduling

changes, the proposed amendments incorporate (1) the relocation of the

list of primary containment isolation valves in accordance with Generic

Letter 91-08, ``Removal of Component Lists from Technical

Specifications,'' and (2) a revision of the interval for functional

testing of hydrogen recombiners from 6 months to 18 months in

accordance with Generic Letter 93-05, ``Line-Item Technical

Specifications Improvements to Reduce Surveillance Requirements for

Testing During Power Operation.''

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1) Involve a significant increase in the probability or

consequences of an accident previously evaluated because of the

following:

a. The relocation of Technical Specification 3/4.6.1.2, Primary

Containment Leakage, and Surveillance Requirements 4.6.1.1.a,

4.6.4.3, and 4.6.6.1.d to specification 3/4.6.1.1, Primary

Containment Integrity, as Surveillance Requirement 4.6.1.1.b

continues to assure that Primary Containment leakage is maintained

within the analyzed limit assumed for accident analysis by testing

in accordance with 10 CFR part 50, Appendix J as modified by

approved exemptions.

The requirement to be less than 0.75 La for as-left Type A

test and less than 0.60 La for Type B and C tests prior to

first unit startup following testing performed in accordance with 10

CFR part 50, Appendix J, as modified by approved exemptions,

provides margin for degradation between tests and thus primary

containment integrity is maintained during the time period between

required leakage testing. The current Limiting Condition for

Operation 3.6.1.2 in conjunction with Surveillance Requirements

4.6.1.2 basically require the same leakage limits as proposed

Surveillance Requirement 4.6.1.1.b. The Limiting Condition for

Operation (LCO) is required to be less than 1.0 La and is

applicable during a fuel cycle for the Type A test. The LCO for Type

B and C combined leakage total is currently required to be less than

0.60 La. The proposed Surveillance Requirement maintains the

following:

1.The current LCO for Overall Containment leakage (as determined

by a Type A test) and for the Type B and C combined leakage during

the cycle by requiring overall containment leakage to be less than

1.0 La and Type B and C leakage total less than 0.60 La.

2. The associated limits specified in the current Action

Statements are maintained by verifying Overall Containment leakage

to be less than 0.75 La and Type B and C leakage total less

than 0.60 La prior to startup from an outage in which the

applicable leakage testing is conducted.

Therefore, there is no change to the consequences of an accident

previously evaluated, because maintaining leakage within the

analyzed limit assumed for accident analysis does not change either

the onsite or offsite dose consequences resulting from an accident.

In addition to this, containment leakage is not an accident

initiator, so there is no effect on the probability of accident

initiators. Thus there is no increase in the probability of an

accident previously analyzed.

b. Relocation of Technical Specification table of Primary

Containment Isolation Valves, Table 3.6.3-1, to the LaSalle UFSAR is

an administrative change to remove the component list of Primary

Containment Isolation Valves, Table 3.6.3-1, from the Technical

Specifications. The Limiting Condition for Operation (LCO), 3.6.3,

is being revised to define which components the LCO applies to. The

wording of the revised LCO encompasses all of the components listed

in the current Technical Specification Table 3.6.3. Removal of this

component list does not change the probability of any accident

initiators or change any other relevant initial assumptions. Also,

there is no change to the consequences of an accident previously

evaluated, because removing this list from Technical Specifications

does not change either the onsite or offsite dose consequences

resulting from the event. The component list will be controlled by

an Administrative Procedure and can only be changed by the 10 CFR

50.59 change process with review and approval per the Onsite Review

and Investigative Function. Therefore, there is no increase in

either the probability or consequences of an accident previously

evaluated.

c. The change in the functional test interval for the Drywell

and Suppression Chamber Hydrogen Recombiner systems from ``once per

6 months'' to ``once per 18 months'' was determined by the NRC in

NUREG 1366 and Generic Letter 93-05 to be acceptable by evaluation

of the industry Licensing Event Reports (LERs) to assess the

reliability of hydrogen recombiners. The conclusion was that the

interval should be changed, because of the redundancy and apparent

high reliability. A review of LaSalle LERs has shown only one LER

that involved the operability of the hydrogen recombiner system and

that was due to a Part 21 issue regarding circuit breaker

environmental qualification. The breakers were replaced with

qualified breakers. Therefore, the LaSalle Hydrogen Recombiner

reliability is consistent with or better than that found by the NRC

in determining this surveillance interval extension based on all

LERs. Also, redundancy is the same as that assumed by the NRC;

because, LaSalle has two hydrogen recombiner subsystems that are

shared by Unit 1 and Unit 2. Both hydrogen recombiners subsystems

are required to be Operable for either or both units in Operational

Conditions 1 and 2. Based on LaSalle operating experience, the

hydrogen recombiner subsystems are expected to continue to be

demonstrated operable when the functional test is performed at an 18

month frequency.

Therefore, there is minimal or no change to the consequences of

an accident previously evaluated, because at least one of the

hydrogen recombiner subsystems is expected to be available to meet

its design function to reduce the potential for hydrogen explosion

or hydrogen burn in the primary containment. By preserving the

integrity of the primary containment, there is no change to either

the onsite or offsite dose consequences resulting from an accident.

In addition to this, control of hydrogen concentration by use of a

hydrogen recombiner subsystem is not an accident initiator, so there

is no effect on the probability of accident initiators. Thus there

is no significant increase in the probability of an accident

previously analyzed.

d. The first exemption request is from the requirements of

paragraph III.A.6(b) of Appendix J to allow LaSalle County Station

Unit Two to return to or resume a Type A test schedule of three

times in ten years (40 plus or minus 10 months). Due to consecutive

failures, 10 CFR 50 Appendix J requires that Type A tests be

performed every refueling outage on Unit Two until two consecutive

Type A tests are satisfactory. 10 CFR Part 50 has an exemption

process and is specified in 10 CFR Part 50.12(a), which states:

``The Commission may, upon application by any interested person

or upon its own [[Page 499]] initiative, grant exemptions from the

requirements of the regulations of this part,...''

The exemption process requires showing that the granting of the

exemption is authorized by law, will not present an undue risk to

the public health and safety, and is consistent with the common

defense and security. Also, special circumstances are required to be

present for the granting of an exemption. One of the special

circumstances that would apply in this instance is 10 CFR part

50.12(a)(2)(ii) which states:

``Application of the regulation in the particular circumstances

would not serve the underlying purpose of the rule or is not

necessary to achieve the underlying purpose of the rule''.

This requires that it be shown that unacceptable containment

leakage will be identified and corrected, by alternative methods.

The alternative method is specifically Type B and C tests, which

will identify any local penetration leakage. This is acceptable,

because Type C test failures have been the cause for failures of as-

found Type A tests in the LaSalle Unit 2 first, third, and fourth

refueling outages.

Exceeding the allowable leakage rate during the performance of

the Type A test is indicative of either a passive or a structural

component that is leaking or that there is an inadequacy in the

Local Leak Rate Test (Type B and C tests) program. When the failure

of a Type A test is due to a passive or structural component, the

only test for adequate repair would be the Type A test. For a Local

Leak Rate Test program inadequacy, the Type A test would serve as a

means of verification of the results of the test program. The Type A

tests have not found new significant Type B or C tested local

penetration leakage that has not been identified by Type B or C

testing alone. Therefore, the LaSalle Local Leak Rate Test program

is adequate to find and correct Type B and C containment penetration

leakage.

When it is determined that Type A tests failed as a direct

result of as-found Type B and C minimum path leakage penalty

additions and not due to a non Type B or C tested components or

structures, then performance of the Type A test more frequently as

required by 10 CFR Part 50, Appendix J, due only to Type B and C

test failures is redundant to the performance of Type B and C tests.

Therefore, Type B or C tested penetration leakage that can be

determined by Type B or C tests is evaluated and corrected, as

applicable, to maintain overall containment leakage within limits,

without an additional Type A test.

Primary Containment leakage which includes the minimum path

Primary Containment Isolation Valve leakage is an assumption in any

analyzed accident which could involve an offsite radioactive

release. Because performance of Type B and C tests will find and

allow correction/repair of leaking valves/penetrations, verification

of as-found and as-left local leakage assures that Primary

Containment leakage will be within the analyzed limit assumed for

accident analysis.

Therefore, for this one-time exemption for LaSalle Unit 2, there

is little or no increase in the consequences of an accident

previously evaluated involving the dose previously calculated either

onsite or offsite at the site boundary due to any analyzed accident.

In addition to this, containment leakage is not an accident

initiator, so there is no effect on the probability of accident

initiators. Thus there is no significant increase in the probability

of an accident previously analyzed.

e. The request for a partial exemption from paragraph III.D of

Appendix J to 10 CFR 50 involves a deletion of the requirement to

perform the third Type A test for each 10-year service period during

the shutdown for the 10-year plant inservice inspections. There is

no significant benefit in coupling these two surveillances (i.e.,

the Type A test and the 10-year ISI program). Each of the two

surveillances is independent of the other and provides assurance of

different plant characteristics. The Type A test assures the

required leak-tightness for the reactor containment building be less

than Appendix J acceptance criteria. This demonstrates compliance

with the guidelines of 10 CFR Part 100 based on the assumptions used

in the UFSAR which conform to NRC Safety Guide 4. The 10-year ISI

program provides assurance of the integrity of the plant structures,

systems, and components in compliance with 10 CFR 50.55(a). There is

no safety-related concern necessitating their coupling to the same

refueling outage. As a result, this change cannot increase the

consequences (i.e., offsite dose) of any accident previously

evaluated. Furthermore, since the decoupling of the test schedules

has no affect on the test's effectiveness, decoupling their

schedules will not increase the probability of an accident.

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

from any accident previously evaluated because:

a. Technical Specification 3/4.6.1.2, Primary Containment

Leakage, and Surveillance Requirements 4.6.1.1.a, 4.6.4.3, and

4.6.6.1.d are being relocated to specification 3.4.6.1.1, Primary

Containment Integrity, as Surveillance Requirement 4.6.1.1.b. The

proposed Surveillance Requirement 4.6.1.1.b assures that Primary

Containment leakage is maintained within the analyzed limit assumed

for accident analysis by testing in accordance with 10 CFR part 50,

Appendix J as modified by approved exemptions. Primary containment

leakage is an assumption in accident analyses, and is maintained by

both the current specifications and the proposed specification. The

leakage does not cause an accident and no new failure modes are

created. Therefore this request for exemption does not create the

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

previously evaluated.

b. This is an administrative change to control the list of

Primary Containment Isolation Valves outside the LaSalle Unit 1 and

Unit 2 Technical Specifications. The administrative controls

provided to control this component list assure that the design and

operation of the plant will continue to be in accordance with the

UFSAR, Facility License and the associated Technical Specifications.

Therefore, the possibility of a new or different kind of accident

from any previously evaluated is not created.

c. The change in the functional test interval for the Drywell

and Suppression Chamber Hydrogen Recombiner systems from ``once per

6 months'' to ``once per 18 months'' is based on good equipment

performance on a 6 month frequency. The expected outcome of the 18

month surveillances, based on the low failure rate at a six month

frequency, is to show the hydrogen recombiner subsystems Operable.

This system is for mitigating the consequences of an accident that

causes generation of hydrogen and oxygen in the primary containment.

No new failure modes are created by this change in surveillance

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

accident from any previously evaluated is not created.

d. The first exemption is from the requirements of paragraph

III.A.6(b) of Appendix J to allow LaSalle County Station Unit Two to

return to or resume a Type A test schedule of three times in ten

years (40 plus or minus 10 months). Containment leakage testing,

including both Type B and C testing and Type A testing as specified

in the LaSalle County Station Safety Analysis Report were evaluated

in Section 6.2.6 of Safety Evaluation Report, NUREG-0519, and found

to be acceptable. Since Type B and C testing will find and verify

correction of penetration leakage when Type B and C test as-found

penalties are specifically what caused the failure of the as-found

Type A tests, then Type B and C testing will provide adequate

assurance of the continued integrity of the Primary Containment

without increasing the frequency of Type A tests. As a result, the

Primary Containment will continue [to] be maintained as designed and

previously evaluated.

Based on this, the requirement of two acceptable as-found Type A

tests prior to returning to the Appendix J paragraph III.D frequency

of three times in ten years (40 plus or minus 10 months) is not

necessary to assure that the primary containment remains within the

analyzed leakage limits. Containment leakage is an assumption for

the dose consequences of accident analyses, and not an accident

initiator. Also, no new failure modes are created by this exemption.

Therefore this Amendment does not create the possibility of a new or

different kind of accident.

e. The request for a partial exemption from paragraph III.D of

Appendix J to 10 CFR 50 involves a deletion of the requirement to

perform the third Type A test for each 10-year service period during

the shutdown for the 10-year plant inservice inspections. The

proposed exemption does not involve any change to the plant design

or operation. As discussed above, this change cannot increase the

consequences of any accident previously evaluated. As a result, no

new failure modes are created. Therefore, this proposed change

cannot create the possibility of any new or different kind of

accident from any accident previously evaluated.

3) Involve a significant reduction in the margin of safety

because:

a. Technical Specification 3/4.6.1.2, Primary Containment

Leakage, and Surveillance Requirements 4.6.1.1.a, 4.6.4.3,

[[Page 500]] and 4.6.6.1.d are being relocated to specification 3/

4.6.1.1, Primary Containment Integrity, as proposed Surveillance

Requirement 4.6.1.1.b. The proposed Surveillance Requirement

4.6.1.1.b continues to assure that Primary Containment leakage is

maintained within the analyzed limit assumed for accident analysis

by testing in accordance with 10 CFR part 50, Appendix J as modified

by approved exemptions.

As stated in 1)a. above, the proposed Surveillance Requirement

4.6.1.1.b maintains the acceptance criteria and limits for continued

operation of the current specification for primary containment

leakage. Therefore, the margin of safety is not reduced by this

change. Also, the proposed addition of a definition for the maximum

allowable primary containment leakage rate assures that the margin

of safety is maintained.

The leakage limits for MSIVs and hydrostatically tested valves

are maintained by relocating the current surveillance requirements

to specification 3/4.6.3, with the acceptance criteria of the

current specification retained. Thus preserving the current margin

of safety by maintaining the leakage rates as assumed in the

accident analyses.

b. The Limiting Condition for Operation for Technical

Specification 3.6.3, Primary Containment Isolation Valves, is

revised by this Technical Specification change to specifically

define the components to which the LCO applies. Therefore, removal

of Technical Specification Table 3.6.3-1, which lists the specific

components to which the LCO applies does not change the scope or

applicability of the specification. The component list will be

controlled administratively with any changes to the list made in

accordance with the 10 CFR 50.59 change process. Therefore, this is

an administrative change only and there is no reduction in the

margin of safety.

c. The change in the functional test interval for the Drywell

and Suppression Chamber Hydrogen Recombiner systems from ``once per

6 months'' to ``once per 18 months'' is based on good equipment

performance on a 6 month frequency. The expected outcome of the 18

month surveillances, based on the low failure rate at a six month

frequency, is to show the hydrogen recombiner subsystems Operable.

The change in frequency has no affect on the hydrogen or oxygen

generation assumptions or the recombination rate of the hydrogen

recombiner subsystems. Therefore, the margin of safety is not

reduced or changed by this surveillance interval change.

d. The first exemption is from the requirements of paragraph

III.A.6(b) of Appendix J to allow LaSalle County Station Unit Two to

return to or resume a Type A test schedule of three times in ten

years (40 plus or minus 10 months). The limit of total leakage

determined from Type B and C tests will remain the same, providing a

margin of 40 percent to the maximum allowable containment leakage

rate (La) at the design basis accident pressure specified in

proposed Technical Specification definition of La. This 40

percent is as specified by 10 CFR Part 50, Appendix J. In addition

to this, administrative guidelines have been set for each

penetration/valve, so that any abnormal leakage will be corrected by

adjustment or repair as needed. Any postponement of repairs is based

on a technical evaluation and then only if the total Type B and Type

C leakage is maintained at less than 0.60 La. Repairs will be

required to restore the leakage rate to less than the administrative

limit at the next refueling outage.

This request for exemption is based the fact that Type B and C

testing minimum path leakage rate penalties are the direct cause of

the failure of as-found Type A tests. The leakage through Type B and

C tested penetrations is best measured and corrected via a local

leak test. Therefore, verification of an adequate margin of safety

is assured by conducting Type B and C tests, and not another

increased frequency Type A test.

e. The request for a partial exemption from paragraph III.D of

Appendix J to 10 CFR 50 involves a deletion of the requirement to

perform the third Type A test for each 10-year service period during

the shutdown for the 10-year plant inservice inspections. The

proposed exemption does not change the acceptance criteria that must

be met for inservice inspections, does not relax the condition of

containment that must be met prior to plant restart, and does not

change the requirements that must be met between plant refueling

outages. Therefore, 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

requested amendments involve no significant hazards consideration.

Local Public Document Room location: Public Library of Illinois

Valley Community College, Rural Route No. 1, Oglesby, Illinois 61348

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

Austin, One First National Plaza, Chicago, Illinois 60690

NRC Project Director: Robert A. Capra

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

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

Date of amendment request: November 21, 1994

Description of amendment request: The proposed amendment would

revise the Technical Specifications to allow a one-time extension of

the allowed outage time for an inoperable reserve source of offsite

power.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed amendment does not involve a significant

increase in the probability of occurrence or consequences of an

accident previously evaluated.

The proposed changes will extend the allowed outage time for the

Reserve source of off-site power, on a one time basis, to allow the

installation of high speed protective relays on the unit system

auxiliary transformers which will increase the level of protection

from ground faults on the low voltage (secondary) side of the

transformers. Operation of Zion, Units 1 and 2, in accordance with

the proposed requirements will not affect the initiators or

precursors of any accident previously evaluated. Operation in

accordance with the proposed requirements will not increase the

likelihood that a transient initiating event will occur because

transients are initiated by equipment malfunction and/or

catastrophic system failure. As a result, the probability of

occurrence of accidents previously evaluated is not significantly

increased.

During the [system auxiliary transformer] SAT outage, power to

the shut down unit will be provided by backfeeding off-site power

through the unit main power transformers and the UAT to supply the

unit non-essential 4-KV service buses. Emergency on-site power will

be available to the shut down unit from at least one unit specific

[emergency diesel generator] EDG when fuel is in the reactor core.

This will ensure that at least one train of Residual Heat Removal

(RHR) will have an emergency source of AC power at all times. RHR

Train A is powered by ESF bus 149(249) which can be energized by the

1B(2B) EDG during a loss of off-site power. RHR Train B is powered

by bus 148(248) which can be energized by the 1A(2A) EDG. Because

the 'O' EDG must be operable for the operating unit, it will also be

available to energize the Division 7 ESF bus on the shut down unit.

The 'O' EDG can supply buses 147 and 247 simultaneously if the need

should arise during an emergency.

Power to the operating unit (opposite unit) will be provided by

the SAT and the UAT in the normal at-power configuration. Emergency

on-site will be provided by the two unit specific EDGs (A and B) and

the common 'O' EDG. In accordance with the proposed requirements,

the Reserve source of off-site power will not be removed from

service unless all three EDGs are operable and the normal source of

off-site power is operable. Administrative controls will be in place

to limit activities in the switchyard that could impact the

reliability of the remaining source of off-site power to the unit.

The Zion PRA was used to compare the impact of extending the

action time versus the impact of manual reactor shutdown on core

damage probability. The PRA result concluded that the risk of

continuing to operate the operating unit for an additional 11 days

with the shutdown unit's SAT out of service is not significantly

greater than the risk of manually shutting down the operating unit

at the expiration of the current 72 hour action statement and is not

significant when compared to the total core damage probability in a

year.

The revised surveillance requirements will provide additional

assurance that redundant sources of power are maintained operable

while the reserve source of off-site power is

[[Page 501]] unavailable. The ability to safely shut down the

operating unit and mitigate the consequences of all accidents

previously evaluated will be maintained. The reserve source of off-

site power is not relied upon in any design basis accident.

Therefore, based on the previous discussion, the proposed changes do

not involve a significant increase in consequences of any accident

previously evaluated.

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

new or different kind of accident from any previously analyzed.

The proposed changes to the Technical Specifications do not

involve the addition of any new or different types of safety-related

equipment, nor does it involve the operation of equipment required

for safe operation of the facility in a manner different from those

addressed in the safety analysis. No safety related equipment or

function will be altered as a result of the proposed changes. Also,

the procedures governing normal plant operation and recovery from an

accident are not changed by the proposed Technical Specification

changes. The proposed changes will extend the allowed outage time

for the Reserve source of off-site power, on a one-time basis, to

allow the installation of high speed protective relays on the unit

system auxiliary transformers which will increase the level of

protection from ground faults on the low voltage (secondary) side of

the transformers. The addition of the high speed relaying has been

evaluated pursuant to 10 CFR 50.59, and no unreviewed safety

questions were identified.

Requirements will be modified to require additional assurance

that the remaining off-site source of AC power and the on-site

source of emergency (emergency diesel generators) are OPERABLE.

Since no new failure modes or mechanisms are added by the proposed

changes, the possibility of a new or different kind of accident is

not created.

3. The proposed changes do not involve a significant reduction

in a margin of safety.

The proposed changes will extend the allowed outage time for the

reserve source of off-site power, on a one-time basis, to allow for

installation of high speed protective relays on the unit system

auxiliary transformers which will increase the level of protection

from ground faults on the low voltage (secondary) side of the

transformers.

During the SAT outage, power to the operating unit (opposite

unit) will be provided by the unit SAT and the UAT in the normal

configuration. Emergency on-site power will be provided by the two

unit specific EDGs (A and B) and the common 'O' diesel generator.

Because the accident analyses take no credit for offsite power

availability, this temporary degradation will not impact the

analysis results.

No safety system setpoints are changed by this proposal. There

is no impact on any physical design margins, and no analytical

results are affected by this change. The revised surveillance

requirements will provide additional assurance that redundant

sources of power are maintained operable while the Reserve source of

off-site power is unavailable.

Based on the above discussion, the ability to safely shut down

the operating unit and mitigate the consequences of all accidents

previously evaluated will be maintained. Therefore, the margin of

safety is not significantly affected.

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: Waukegan Public Library, 128

N. County Street, Waukegan, Illinois 60085

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

Austin, One First National Plaza, Chicago, Illinois 60690

NRC Project Director: Robert A. Capra

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

Buren County, Michigan

Date of amendment request: October 5, 1994

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

revise primary coolant system (PCS) pressure-temperature (P-T) limits,

power-operated relief valve (PORV) setting limits, and primary coolant

pump starting limits to accommodate reactor vessel fluence for an

additional 4 effective full power years (up to 2.192 x 1019nvt).

The existing P-T limit curves are calculated for a fluence of 1.8 x

1019 could be reached as early as March 1, 1995; (2) require the

high pressure safety injection (HPSI) pumps to be ``rendered incapable

of injection into the PCS'' when the PCS is below 300 deg.F, rather

than the existing requirement to render both HPSI pumps ``inoperable''

when the PCS is below 260 deg.F. This change supports the assumption in

the P-T limit analyses that HPSI injection would not occur below

300 deg.F; and (3) establish a more restrictive limit on pressurizer

heatup rate to achieve consistency between design assumptions and

technical specification (TS) limits. The limit in the existing TS is

less restrictive than used in design calculations. Neither the design

heatup rate nor the TS heatup rate limit is achievable with installed

equipment.

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 following evaluation supports the finding that operation of

the facility in accordance with the proposed Technical

Specifications would not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The revision of the Primary Coolant Pump [PCP] starting limits, PCS

P-T curves, and PORV setting limits would not cause any changes to the

capability or operation of plant systems that would affect the

probability of occurrence or consequences of an accident. These

revisions simply update the existing requirements to account for

additional reactor vessel fluence.

The reduction of the allowable pressurizer heatup rate would

have no effect on operation of the plant. The current limit is

physically unobtainable with installed equipment. The proposed

change better aligns the Technical Specification limits with the

design analysis. The change in the pressurizer heatup rate limit

will not increase the probability or consequences of an accident.

Requiring the HPSI pumps to be operable when above 325 deg.F,

rather than when above 300 deg.F does not affect the probability or

consequences of any accident previously evaluated. Neither the

existing 300 deg.F requirement nor the proposed 325 deg.F

requirement has an analytical base. This requirement was recently

changed from 325 deg.F to 300 deg.F simply for uniformity. With the

revised P-T limit analysis requirement to assure that inadvertent

HPSI injection will not occur below 300 deg.F, it is necessary to

revert to the former limit of 325 deg.F to provide time to

transition between these two contrasting HPSI pump requirements.

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

from any previously evaluated.

The revised specifications, PCP starting limits, PCS P-T limits,

pressurizer heatup rate, PORV setting limits, and HPSI pump

restrictions, all are directly related to, and intended to prevent,

a previously analyzed event, failure of the Reactor Coolant Pressure

Boundary. Revision of these limits would not create the possibility

of a new or different kind of accident.

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

The revised PCP starting limits, PCS P-T limits, and PORV

setting limits are calculated using a similar methodology as the

limits which they replace. Therefore they provide the same margin of

safety.

The revised pressurizer heatup rate reduces the currently

allowable limit which is in the direction of increased margin of

safety. Since there is no equipment installed which would cause

either the existing or the proposed limit to be reached, there will

be no change on the operation of the plant equipment. Therefore

reducing the limit on the pressurizer heatup rate will not involve a

significant reduction in the margin of safety.

Requiring the HPSI pumps to be operable when above 325 deg.F,

rather than when above 300 deg.F does not involve a significant

reduction in any margin of safety. Neither the existing 300 deg.F

requirement nor the proposed [[Page 502]] 325 deg.F requirement has

an analytical base. This requirement was recently changed from

325 deg.F to 300 deg.F simply for uniformity. With the revised P-T

limit analysis requirement to assure that inadvertent HPSI injection

will not occur below 300 deg.F, it is necessary to revert to the

former limit of 325 deg.F to provide time to transition between

these two contrasting HPSI pump requirements.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Van Wylen Library, Hope

College, Holland, Michigan 49423.

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

Company, 212 West Michigan Avenue, Jackson, Michigan 49201

NRC Project Director: John N. Hannon

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

One,Unit No. 1, Pope County, Arkansas

Date of amendment request: November 8, 1994

Description of amendment request: The proposed amendment revises

technical specifications (TSs) associated with requirements for

performing the containment integrated leak rate test (ILRT). The

proposed change describes the ILRT test frequency by referencing the

test frequency requirements included in 10 CFR Part 50, Appendix J. The

existing specifications paraphrase the Appendix J requirements, but

include differences that result in interpretation problems.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

Criterion 1 - Does Not Involve a Significant Increase in the

Probability or Consequences of an Accident Previously Evaluated.

The proposed change revises Technical Specification 4.4.1.1.4 to

reference the testing frequency requirements of 10 CFR 50, Appendix

J, and to state that NRC approved exemptions to the applicable

regulatory requirements are permitted. The current requirements of

TS 4.4.1.1.4 paraphrase the requirements of Section III.D.1.(a) of

Appendix J. The proposed administrative revision simply deletes the

paraphrased language and directly references Appendix J. No new

requirements are added, nor are any existing requirements deleted.

An approved exemption to Section III.D.1.(a) of Appendix J would not

necessarily affect the requirements of TS 4.4.1.1.4, unless the

proposed clarification phrase permitting the use of approved

exemptions is added. Any specific changes to the requirements of

Section III.D.1(a) will require a submittal from Entergy Operations

under 10CFR50.12 and subsequent review and approval by the NRC prior

to implementation. The proposed change is stated generically to

avoid the need for further TS changes if different exemptions are

approved in the future.

The proposed change, in itself, does not affect reactor

operations or accident analysis and has no radiological

consequences. The change provides clarification so that TS changes

will not be necessary in the future to correspond to applicable NRC

approved exemptions from the requirements of Appendix J. Therefore,

this change does not involve a significant increase in the

probability or consequences of any accident previously evaluated.

Criterion 2 - Does Not Create the Possibility of a New or

different Kind of Accident from any Previously Evaluated.

The proposed change provides clarification to a specification

which paraphrases a codified requirement. Since the proposed

amendment would not change the design, configuration or method of

operation of the plant, it would not create the possibility of a new

or different kind of accident from any previously evaluated.

Criterion 3 - Does Not Involve a Significant Reduction in the

Margin of Safety.

The proposed change is administrative and clarifies the

relationship between the requirements of TS 4.4.1.1.4, Appendix J,

and any approved exemptions to Appendix J. It does not, in itself,

change a safety limit, an LCO, or a surveillance requirement on

equipment required to operate the plant. The NRC will directly

approve change proposed exemption to III.D.1.(a) of Appendix J prior

to implementation. Therefore, this change does not involve a

significant reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Tomlinson Library, Arkansas

Tech University, Russellville, Arkansas 72801

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

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

NRC Project Director: William D. Beckner

Georgia Power Company, Oglethorpe Power Corporation, Municipal

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

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

Appling County, Georgia

Date of amendment request: December 2, 1994

Description of amendment request: The proposed amendments would

replace Appendix B, ``Environmental Technical Specifications'' with an

Environmental Protection Plan (Nonradiological) and revise the

Operating Licenses to reflect these changes. The proposed changes are

administrative in nature, altering only the format and location of

programmatic controls and procedural details relative to

nonradiological environmental monitoring.

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 amendments do not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The proposed changes to the Environmental Technical

Specifications (ETS) are administrative in nature, altering only the

format and location of programmatic controls and procedural details

relative to nonradiological environmental values. The proposed

Environmental Protection Plan (EPP) (Nonradiological) contains the

programmatic controls now residing in the ETS, with appropriate

plant procedures serving as implementing documents. The proposed

changes to the operating licenses are also administrative in nature

and change the Appendix B reference from ETS to EPP. Compliance with

applicable regulatory requirements will be maintained. In addition,

the proposed changes do not alter the conditions or assumptions in

any of the accident analyses. Therefore, these proposed changes do

not involve a significant increase in the probability or

consequences of an accident previously evaluated.

2) The proposed amendments do not create the possibility of a

new or different kind of accident from any accident previously

evaluated.

The proposed changes to the ETS do not involve any change to the

configuration or method of operation of any plant equipment. These

proposed changes are administrative in nature and consist of

replacing the ETS with an EPP. The proposed changes to the operating

licenses are also administrative in nature and change the Appendix B

reference from ETS to EPP. Accordingly, no new failure modes have

been identified for any plant system or component important to

safety nor has any new limiting single failure been identified as a

result of the proposed changes. Therefore, the proposed changes do

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

from any accident previously evaluated.

3) The proposed amendments do not result in a significant

reduction in the margin of safety.

The proposed changes to the ETS relate primarily to matters

involving recordkeeping, reporting, and administrative procedures or

requirements. No significant change in the type or quantity of any

effluent release will result from this action. These changes replace

[[Page 503]] the ETS with an EPP. The proposed EPP contains the

programmatic controls now residing in the ETS, with appropriate

plant procedures serving as implementing documents to ensure

compliance with applicable regulatory requirements. The proposed

changes to the operating licenses are also administrative in nature

and change the Appendix B reference from ETS to EPP. Therefore, the

proposed changes do not involve a significant reduction in a margin

of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Appling County Public Library,

301 City Hall Drive, Baxley, Georgia 31513

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

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

NRC Project Director: Herbert N. Berkow

Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388

Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County,

Pennsylvania

Date of amendment request: November 21, 1994

Description of amendment request: The proposed amendment would

eliminate the Main Steam Isolation Valve (MSIV) - Leakage Control

System (LCS) including the primary containment isolation valves

associated with the MSIV - LCS, along with increasing the allowable

MSIV leakage rates.

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:

I. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed changes to TS Section 3.6.1.2 do not involve a

change to structures, components, or systems that would affect the

probability of an accident previously evaluated. The TS limits for

MSIVs are increased from 46 scf per hour for all four main steam

lines to less than or equal to 100 scf per hour for any one MSIV and

a combined maximum pathway leakage rate of less than or equal to 300

scf per hour for all four main steam lines. The consequences of an

accident are affected as discussed in this section.

The proposed changes to TS Section 3.6.1.4 eliminate the Main

Steam Isolation Valves (MSIVs) Leakage Control System (LCS)

requirements from the TS. As described in Section 6.7 of the FSAR,

the LCS is manually initiated in about 20 minutes following a design

basis Loss of Coolant Accident (LOCA). Since the LCS is operated

only after an accident has occurred, these proposed changes have no

effect on the probability of an accident.

Since MSIV leakage and operation of the LCS are included in the

radiological analysis for the design basis LOCA as described in

Section 15.6.5 of the FSAR, the proposed changes do not affect the

precursors of other analyzed accidents. Analysis of the effects of

the proposed changes do, however, result in acceptable radiological

consequences for the design basis LOCA previously evaluated in

Section 15.6.5 of the FSAR.

SSES, Units 1 and 2 have an inherent MSIV leakage treatment

capability as discussed below. We propose to use the drain lines

associated with the main steam lines and main turbine condenser as

an alternative to the guidance in Regulatory Guide 1.96, ``Design of

Main Steam Isolation Valve Leakage Control System For Boiling Water

Nuclear Power Plants'', Revision 0, May 1975, for MSIV leakage

treatment. If approved, we will incorporate this alternate method in

the appropriate operational procedures and Emergency Operating

Procedures.

The Boiling Water Reactor Owners' Group (BWROG) has evaluated

the availability of main steam system piping and main condenser

alternate pathways for processing MSIV leakage, and has determined

that the probability of a near coincident LOCA and a seismic event

is much smaller than for other plant safety risks. Accordingly, this

alternate MSIV leakage treatment pathway is available during and

after a LOCA. Nevertheless, the BWROG has also determined that main

steam piping and main condenser design are extremely rugged, and the

design requirements applied to SSES Unit 1 and Unit 2 main steam

system piping and main condenser contain substantial margin, based

on the original design requirements. Therefore, the alternate

treatment method has been evaluated for its capability to mitigate

the consequences of a LOCA, and has been evaluated to assure its

availability considering a seismic event.

In order to determine the capability of the main steam piping

and main condenser alternate treatment pathway, the BWROG has

reviewed earthquake experience data on the performance of non-

seismically designed piping and condensers during past earthquakes.

The data is summarized in General Electric (GE) Report, ``BWROG

Report for Increasing MSIV Leakage Rate Limits and Elimination of

Leakage Control Systems,'' NEDC 31858P, Revision 2, submitted to the

NRC by BWROG letter dated October 4, 1993. This study concluded that

the possibility of a failure that could cause a loss of steam or

condensate in Boiling Water Reactor (BWR) main steam piping or

condensers in the event of a design basis (i.e., safe shutdown)

earthquake is highly unlikely, and that such a failure would also be

contrary to a large body of historical earthquake experience data,

and thus unprecedented.

A verification has been performed of the seismic adequacy of the

Unit 1 and Unit 2 main steam piping and main condenser consistent

with the guidelines discussed in Section 6.7 of NEDC-31858P,

Revision 2, to provide reasonable assurance of the structural

integrity of these components. An evaluation, including the walkdown

report outliers, ``MSIV Leakage Alternate Treatment Method Seismic

Evaluation,'' for Unit 1 and Unit 2, is attached. The results of the

evaluation clearly demonstrate that the MSIV Leakage Alternate

Treatment Method meets the intent of 10CFR100 Appendix A, with

regards to seismic qualification. Except for the requirement to

establish a proper flow path from the MSIVs to the condenser, the

proposed method is passive and does not require any additional logic

control and interlocks. The method proposed for MSIV leakage

treatment is consistent with the philosophy of protection by

multiple barriers used in containment design for limiting fission

product release to the environment.

A plant-specific radiological analysis has been performed in

accordance with NEDC-31858P, Revision 2, to assess the effects of

the proposed increase to the allowable MSIV leakage rate in terms of

control room and off-site doses following a postulated design basis

LOCA. This analysis utilizes the hold-up volumes of the main steam

piping and condenser as an alternate method for treating the MSIV

leakage. As discussed earlier, there is reasonable assurance that

the main steam piping and condenser remain intact following a design

basis earthquake. The radiological analysis uses standard

conservative assumptions for the radiological source term consistent

with Regulatory Guide (RG) 1.3, Assumptions Used for Evaluating the

Potential Radiological Consequences of a Loss-Of-Coolant Accident

for Boiling Water Reactor, Revision 2, dated April 1974.

The analysis results demonstrate that dose contributions from

the proposed MSIV leakage rate limit of 100 scfh per steam line, not

to exceed a total of 300 scfh for all four main steam lines, and

from the proposed deletion of the LCS, result in an insignificant

increase to the LOCA doses previously evaluated against the

regulatory limits for the off-site doses and control room doses

contained in 10CFR100 and 10CFR50, Appendix A, General Design

Criterion (GDC) 19, respectively. The off-site and control room

doses resulting from a LOCA are discussed in Section 15.6.5 of the

FSAR. The off-site and control room doses resulting from a LOCA

associated with the proposed changes are the sum of LOCA doses

evaluated in the power uprate revision to the design basis DBA-LOCA

calculation (EC-RADN-1009) and the additional doses calculated using

the alternate MSIV leakage treatment method. Enclosure 3 [of

application dated November 21, 1994] summarizes the off-site and

control room doses and compares the alternate treatment method doses

to the original MSIV-LCS treatment method doses.

The 30-day whole body doses at the Low Population Zone (LPZ) did

not change and remained at .37 rem for the alternate treatment

method. The 30-day control room whole body doses increased slightly

from .38 [[Page 504]] rem to .76 rem for the alternate treatment

method. The increase in control room dose is not significant since

the revised doses are well below the regulatory limits, i.e., .76

rem calculated versus the limit of 5 rem in the control room. The

two-hour whole body dose at the Exclusion Area Boundary (EAB)

decreased slightly from 2.47 rem to 2.217 rem.

The 30-day thyroid dose at the LPZ increased from 30.4 rem for

the MSIV-LCS treatment method to 41.74 rem for the alternate

treatment method. This increase is not significant since the revised

dose of 41.74 rem is well within the regulatory limit of 300 rem.

The two-hour thyroid dose at the EAB decreased slightly from 127.8

rem to 125.61 rem. The 30-day control room thyroid dose increased

from 14.19 rem for the MSIV-LCS treatment method to 18.55 rem for

the alternate treatment method. The increased control room thyroid

dose is not significant since the revised dose remains well below

the regulatory limit of 30 rem.

The 30-day control room beta dose increased insignificantly from

12 rem for the MSIV-LCS treatment method to 12.17 rem for the

alternate treatment method, remaining a small fraction relative to

the limit of 75 rem.

In summary, the proposed changes discussed above do not result

in a significant increase in the radiological consequences of a LOCA

when the same assumptions and methods specified in the FSAR are

used, recognizing that radiological consequences calculated in the

FSAR and for these proposed changes are significantly higher than

those using more realistic assumptions and methods. Nevertheless,

the calculated off-site and control room doses resulting from a LOCA

remain well below the regulatory limits.

The proposed change to TS Table 3.6.3-1 deletes the LCS valves

from the list of primary containment isolation valves. This proposed

change is consistent with the proposed deletion of the LCS. The LCS

lines that are connected to the main steam piping are welded and/or

capped closed to assure primary containment integrity is maintained.

The welding and post weld examination procedures will be in

accordance with American Society of Mechanical Engineers (ASME)

Code, Section III requirements. These welds and/or caps will be

periodically tested as part of the Containment Integrated Leak Rate

Test (CILRT). This proposed change does not involve an increase in

the probability of equipment malfunction previously evaluated in the

FSAR. In fact, this proposed change reduces the probability of

equipment malfunction since, upon implementation of these proposed

changes, the plant will be operated with less primary containment

isolation valves subjected to postulated failure. This proposed

change has no effect on the consequences of an accident since the

LCS lines will be welded and/or cap closed, thus assuring that the

containment integrity, isolation and leak test capability are not

compromised.

The proposed change to TS Table 3.8.4.2.1-1 deletes the LCS

motor operated valves from the list of ``Motor Operated Valves

Thermal Overload Protection - Continuous.'' The proposed change has

no effect on the probability or consequences of an accident since

the valves are eliminated and not performing a safety function.

Therefore, as discussed above, the proposed changes do not

involve a significant increase in the probability or consequences

from any accident previously evaluated.

II. Create the possibility of a new or different kind of

accident from any accident previously evaluated.

As stated in Section I, the proposed changes do not involve a

change to structures, components, or systems that would affect the

probability of an accident previously evaluated , nor would these

changes create any new or different kind of accident from any

previously evaluated. The proposed changes will introduce and take

credit for a new level of operational performance for existing plant

systems and components to mitigate the consequences of the accident.

The effect on this equipment has been evaluated and found to provide

an acceptable level of reliability resulting in the required level

of protection. This conclusion is based on the evaluation performed

in NEDC 31858P, Revision 2, and the plant specific seismic

evaluation provided in the Enclosure 2 [of application dated

November 21, 1994], ``MSIV Leakage Alternate Treatment Method

Seismic Evaluation.'' The Leakage Control System has been installed

to direct any leakage past the MSIVs during the LOCA; acting after

the accident has occurred. The resulting consequences of the

evaluated accidents have been affected as discussed in Section I

resulting in no significant increase in the probability or

consequences of said accident. Therefore, reliance on different

equipment than previously assumed to mitigate the consequences of an

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

of accident from any accident previously evaluated.

The BWROG evaluated MSIV performance and concluded that MSIV

leakage rates up to 200 scfh per valve will not inhibit the

capability and isolation performance of the MSIVs to effectively

isolate the primary containment. Implementation of the proposed

changes does not result in modifications which could adversely

impact the operability of the MSIVs. The LOCA has been analyzed

using the main steam piping and main condenser as a treatment method

to process MSIV leakage at the proposed maximum rate of 100 scfh per

main steam line, not to exceed 300 scfh total for all four main

steam lines. Therefore, the proposed TS Section 3.6.1.2 change to

increase the allowed MSIV leakage rate does not create any new or

different kind of accident from any accident previously evaluated.

The proposed TS Section 3.6.1.4 change to eliminate the LCS does

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

from any accident previously evaluated because the removal of the

LCS does not affect any of the remaining SSES Unit 1 and Unit 2

systems, and the LOCA has been re-analyzed using the proposed

alternate method to process MSIV leakage. The associated proposed

change to delete the LCS isolation valves from TS Table 3.6.3-1 and

Table 3.8.4.2.1-1 does not create the possibility of a new or

different kind of accident. The affected main steam piping will be

welded and/or capped closed to assure that the primary containment

integrity, isolation, and leak testing capability are not

compromised. The affected LCS motor operated valves will be

eliminated so their thermal overloads will not need to be bypassed.

Therefore, as discussed above, the proposed changes do not

create the possibility for any new or different kind of accident

from any accident previously evaluated.

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

The proposed change to TS Section 3.6.1.2 to increase the MSIV

allowable leakage does not involve a significant reduction in the

margin of safety. As discussed in the current Bases for TS Section

3/4.6.1.2, the allowable leak rate limit specified for the MSIVs is

used to quantify a maximum amount of leakage assumed to bypass

primary containment in the LOCA radiological analysis. Accordingly,

results of the re-analysis supporting these proposed changes are

evaluated against the dose limits contained in 10CFR100 for the off-

site doses, and 10CFR50, Appendix A, GDC 19, for the control room

doses. As discussed above, sufficient margin relative to the

regulatory limits is maintained even when assumptions and methods

(e.g., RG 1.3) that are considered highly conservative relative to

more realistic assumptions and methods are used in the analysis.

Results of the radiological analysis demonstrate that the

proposed changes do not involve a significant reduction in the

margin of safety. Whole body doses, in terms of margin of safety,

are insignificantly reduced by .38 rem in the control room. The

margin of safety remains constant for the LPZ whole body dose or

actually increases by .253 rem for the EAB whole body dose. The

margin of safety for thyroid dose category is reduced by 11.34 rem

at the LPZ and 4.36 rem in the control room. The margin of safety is

found to increase for the EAB thyroid dose by 2.19 rem. The margin

of safety for beta dose is insignificantly reduced by .17 rem in the

control room. The reductions in the margin of safety are not

significant since the revised calculated doses are highly

conservative yet remain well below the regulatory limits, and

therefore, a substantial margin to the regulatory limits is

maintained.

The proposed change to eliminate the LCS from TS Section 3.6.1.4

does not reduce the margin of safety, in fact, the overall margin of

safety is increased. The function of the LCS for MSIV leakage

treatment will be replaced by alternate main steam drain lines and

condenser equipment. This treatment method is effective in reducing

the dose consequences of MSIV leakage over an expanded operating

range compared to the capability of the LCS and will, thereby,

resolve the safety concern that the LCS will not function at MSIV

leakage rates higher than the LCS design capacity. Except for the

requirement to establish a proper flow path from the MSIVs to the

condenser, the proposed method is passive and does not require any

new logic control and interlocks. This proposed method is consistent

with the [[Page 505]] philosophy of protection by multiple barriers

used in containment design for limiting fission product release to

the environment. Furthermore, as previously identified, based on the

evaluations discussed in NEDC-31858P, Revision 2, and the seismic

evaluation provided in the Enclosure 2 [of application dated

November 21, 1994] report, ``MSIV Leakage Alternate Treatment Method

Seismic Evaluation,'' the design of the MSIV leakage alternate drain

pathway, meets the intent of the 10CFR100, Appendix A requirement

for seismic qualification. Therefore, the proposed method is highly

reliable and effective for MSIV leakage treatment.

The revised calculated LOCA doses remain within the regulatory

limits for the off-site and the control room. Therefore, the

proposed method maintains a margin of safety for mitigating the

radiological consequences of MSIV leakage for the proposed TS

leakage rate limit of 100 scfh per main steam line, not to exceed a

total of 300 scfh for all four main steam lines.

The proposed change to delete LCS isolation valves from TS Table

3.6.3-1 and Table 3.8.4.2.1-1 does not reduce the margin of safety.

Welded and/or capped closure of the LCS lines assures that the

primary containment integrity and leak testing capability are not

compromised. These welds and/or caps will be periodically leak

tested as part of the CILRT. The LCS motor operated valves will be

eliminated so their thermal overloads will not need to be bypassed.

Therefore, the proposed deletion of the LCS isolation valves does

not involve a 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: Osterhout Free Library,

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

Pennsylvania 18701

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

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

NRC Project Director: John F. Stolz

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

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

County, New Jersey

Date of amendment request: November 18, 1994

Description of amendment request: The proposed change would revise

the Reactivity Control System Technical Specification Limiting

Conditions for Operation for boration flow paths and charging pumps by

reducing the number of operable charging pumps required for boron

addition in Mode 4 from two to one.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Will not involve a significant increase in the probability or

consequences of an accident or malfunction of equipment important to

safety previously evaluated.

The Emergency Core Cooling System (ECCS) requirements assume

that only one charging pump will be available below 350 deg.F

without single failure considerations on the bases of the stable

reactivity condition of the reactor and limited core cooling

requirements. Therefore, the Mode 4 Applicability has been deleted

from LCOs 3.2.1.2 and 3.2.1.4, and was added to LCOs 3.2.1.1 and

3.2.1.3 consistent with the requirements of LCO 3.5.3.

The current Bases for the Unit 2 Technical Specification for

boration system flow paths via the charging pumps supports the use

of a similar LCO for Salem Unit 1.

The limitation for a maximum of one centrifugal charging pump to

be operable when the RCS temperature is less than or equal to

312 deg.F has been added to LCO 3.1.2.3 for clarity and is

consistent with the Cold Overpressure Protection (POPS) analysis and

the requirements of Technical Specification 3.5.3.

The requirements for Boric Acid Transfer Pump operability are

adequately addressed in Technical Specifications 3.1.2.1 and 3.1.2.2

which specify the boron injection flow paths to be operable and the

components required to perform this function. This includes the

availability of the transfer pumps to meet this Technical

Specification requirement.

Therefore, the proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated in the UFSAR.

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

accident from any previously evaluated.

As discussed in response to Question 1 above, the proposed

amendment to the number of charging pumps required to be operable in

Mode 4 is consistent with the current Technical Specification

requirements for the ECCS LCO and the POPS. The current bases for

the Unit 2 Technical Specification for boration system flow paths

via the charging pumps supports the use of a similar LCO for Salem

Unit 1. The requirements for Boric Acid Transfer Pump operability

for Unit 1 are adequately addressed in Technical Specifications

3.1.2.1 and 3.1.2.2 which specify the boron injection flow paths to

be operable and the components required to be available to perform

this function including the transfer pumps. Therefore, the proposed

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

of accident from any previously evaluated.

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

safety. The proposed amendment to the number of charging pumps

required to be operable in Mode 4 will not result in any changes to

the assumptions or conditions for the current ECCS analysis and POPS

analysis. The current bases for the Unit 2 Technical Specification

for boration system flow paths via the charging pumps supports the

use of a similar LCO for Salem Unit 1 (i.e., the Bases are

essentially the same). The requirements for Boric Acid Transfer Pump

operability for Unit 1 are adequately addressed in Technical

Specifications 3.1.2.1 and 3.1.2.2 which specify the boron injection

flow paths to be operable and the components required to be

available to perform this function including the transfer pumps.

Therefore, the proposed amendment 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: Salem Free Public library, 112

West Broadway, Salem, New Jersey 08079

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

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

NRC Project Director: John F. Stolz

Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and

50-364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston

County, Alabama

Date of amendments request: December 19, 1994Description of

amendments request: The proposed change to Table 3.7-3 of the Technical

Specifications includes the revision to the main steam safety valve

(MSSV) setpoint tolerance from plus or minus 1 percent to plus or minus

3 percent and modifies the bases to 3/4.7.1.1 to increase the relieving

capacity of the MSSVs to at least 12,984,660 pounds per hour which

corresponds to approximately 112 percent of total secondary steam flow

at 100 percent rated thermal power. In addition, modifications to Table

3.7-1 are proposed to reduce the allowable power range neutron flux

high setpoints for multiple inoperable steam generator safety valves.

The proposed amendment includes an editorial correction to Bases 3/

4.7.1.2 to indicate required auxiliary feedwater flow at ``1133 psia''

rather than ``1133 psig.''

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 license amendment does not involve a significant

increase in the [[Page 506]] probability or consequences of an

accident previously evaluated.

These proposed changes to the Farley Technical Specifications do

not result in a condition where the design, material and

construction standards of the MSSVs that were applicable prior to

the proposed change are altered. The valves will continue to

function as designed. All applicable safety analyses have been

reviewed, evaluated or reanalyzed and all applicable safety criteria

continue to be met. No accident sequences are altered because of the

proposed amendment. The radiological consequences for the Steam

Generator Tube Rupture were reanalyzed and 10 CFR 100 criteria

continue to be met. All other FSAR radiological analyses remain

bounding. Analyses have been performed to justify the proposed high

nuclear flux setpoint changes. All acceptance criteria for these

analyses continue to be met. Therefore, the proposed amendment does

not result in a significant increase in the probability or

consequences of an accident previously evaluated.

2. The proposed license amendment does not create the

possibility of a new or different accident from any accident

previously evaluated.

The MSSVs continue to have the required pressure relieving

capacity to ensure that system design pressure remains below 110% of

shell design pressure. The proposed changes are not accident

initiators nor do they create any new accident scenarios or any new

limiting single failures. The ability of the MSSVs to respond to an

accident condition is not impaired by the proposed changes. The

proposed high nuclear flux setpoints for multiple valves out of

service ensure all applicable safety criteria for accident analyses

are met. No new accident scenarios are created by these proposed

changes. Therefore, the proposed amendment does not create the

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

previously evaluated.

3. The proposed license amendment does not involve a significant

reduction in the margin of safety.

Acceptance criteria for accident analysis continue to be met.

Radiological consequences for the affected Chapter 15 analysis

remain within 10 CFR 100 acceptance criteria. No safety limits or

safety system setpoint requires modification due to the proposed

changes. The current secondary side over-pressure limit of 100% of

steam generator shell design pressure is not violated. Analysis for

the high nuclear flux setpoints have verified that there is no

reduction in margin for the events analyzed. Therefore, there is not

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: Houston-Love Memorial Library,

212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302

Attorney for licensee: M. Stanford Blanton, Esq., Balch and

Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham,

Alabama 35201

NRC Project Director: William H. Bateman

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

Callaway County, Missouri

Date of amendment request: September 9, 1994

Description of amendment request: The proposed amendment would

revise Technical Specification 3/4.8.1 and its associated Bases to

improve emergency diesel generator reliability and availability.

Several surveillance requirements would be revised or eliminated, and

guidance provided in Regulatory Guide 1.9, Revision 3, and Generic

Letter 93-05 would be incorporated.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

The proposed changes do not involve a significant hazards

consideration because operation of Callaway Plant with these

changes would not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

These proposed changes do not involve a change in the

operational limits or physical design of the emergency power system.

Emergency diesel generator operability and reliability will continue

to be assured while minimizing the number of required emergency

diesel generator starts. Also, emergency diesel generator

reliability will be enhanced by minimizing service test conditions

which can lead to premature failures.

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

from any previously evaluated.

These proposed changes do not involve a change in the

operational limits or physical design of the emergency power system.

The performance capability of the emergency diesel generator will

not be affected. Emergency diesel generator reliability and

availability will be improved by the implementation of the proposed

changes. There is no actual impact on accident analysis.

3. Involve a Significant Reduction in the Margin of Safety.

These proposed changes do not involve a change in the

operational limits or physical design of the emergency power system.

The performance capability of the emergency diesel generator will

not be affected. Emergency diesel generator reliability and

availability will be improved by the implementation of the proposed

changes. No margin of safety is reduced.

Based on the above discussions, it has been determined that the

requested technical specification revision does not involve a

significant increase in the probability or consequences of an

accident or other adverse condition over previous evaluations; or

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

condition over previous evaluations; or involve a significant

reduction in a margin of safety. The requested license amendment

does not involve a significant hazards consideration.

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: Callaway County Public

Library, 710 Court Street, Fulton, Missouri 65251.

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

& Trowbridge, 2300 N Street, N.W., Washington, DC 20037

NRC Project Director: Leif J. Norrholm

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

Callaway County, Missouri

Date of amendment request: September 9, 1994

Description of amendment request: The proposed amendment would

revise Technical Specification 3.8.2.1 and 3.8.2.2, 125-volt D.C.

busses for battery bank and chargers and provides for the installation

of swing chargers during the next refueling outage. Technical

Specifications 3.8.3.1 and 3.8.3.2 would be revised to address the 120-

volt A.C. Vital Busses.

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 to the Technical Specifications do not

involve a significant hazards consideration because operation of

Callaway Plant in accordance with these changes would not:

1) Involve a significant increase in the probability or

consequences of an accident previously evaluated.

These proposed Technical Specification changes do not involve

any hardware changes nor do they affect the probability of any event

initiators. There will be no change to normal plant operating

parameters or accident mitigation capabilities. There will be no

increase in the consequences of any accident or equipment

malfunction. [[Page 507]]

2) Create the possibility for accident or malfunction of

equipment of a different type than previously evaluated in the FSAR.

The proposed Technical Specification changes do not involve any

design changes nor are there any changes to the method by which any

safety-related plant system performs its safety function. The normal

manner of plant operation is unaffected. No new accident scenarios,

transient precursors, failure mechanisms, or limiting single

failures are introduced as a result of these changes.

Involve a significant reduction in the margin of safety.

There will be no affect [SIC] on the manner in which safety

limits or limiting safety system settings are determined, nor will

there be any effect in those plant systems necessary to assure the

accomplishment of protection functions. There will be no impact on

DNBR limits, FQ, F-delta-H, LOCA PCT, peak local power density

or any other margin of safety.

Based on the information presented above, the proposed amendment

does not involve a significant increase in the probability or

consequences of an accident previously evaluated, create the

possibility of a new or different kind of accident from any

previously evaluated, or involve a significant reduction in a margin

of safety. Therefore, it is concluded that the proposed changes meet

the requirements of 10 CFR 50.92(c) and does [SIC] not involve a

significant hazards consideration.

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: Callaway County Public

Library, 710 Court Street, Fulton, Missouri 65251.

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

& Trowbridge, 2300 N Street, N.W., Washington, DC 20037

NRC Project Director: Leif J. Norrholm

Vermont Yankee Nuclear Power Corporation, Docket No. 50-271,

Vermont Yankee Nuclear Power Station, Vernon, Vermont

Date of amendment request: August 27, 1993

Description of amendment request: The proposed amendment would

revise the Technical Specifications (TSs) to be consistent with recent

revisions to 10 CFR Part 20 and 10 CFR 50.36a. Administrative changes

are also proposed.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change will not involve a significant increase

in the probability or consequences of an accident previously

evaluated. The changes as proposed consist of revisions to the

Technical Specifications to meet new regulatory requirements as

contained in 10CFR20 and 10CFR50.36a, and other related changes of

an administrative nature. There is no change in the types and

amounts of effluents released, nor will there be any increase in

individual or cumulative occupational radiation exposures. None of

the changes proposed will affect any plant hardware, plant design,

safety limit settings, or plant system operation, and therefore do

not modify or add any initiating parameters that would significantly

increase the probability or consequences of any previously analyzed

accident.

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

or different kind of accident from any accident previously

evaluated. The changes as proposed do not physically alter the plant

nor do they change the operation of the plant.

3. The proposed change does not involve a significant reduction

in the margin of safety. The changes will not increase the amount or

types of effluents that may be released offsite, nor do they

significantly increase individual or cumulative occupational

radiation exposures. These changes will not alter any of the

requirements or responsibilities for protection of the public and/or

employees against radiation hazards.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Brooks Memorial Library, 224

Main Street, Brattleboro, Vermont 05301

Attorney for licensee: John A. Ritsher, Esquire, Ropes and Gray,

One International Place, Boston, Massachusetts 02110-2624

NRC Project Director: Walter R. Butler

Vermont Yankee Nuclear Power Corporation, Docket No. 50-271,

Vermont Yankee Nuclear Power Station, Vernon, Vermont

Date of amendment request: March 31, 1994

Description of amendment request: The proposed amendment would

modify the requirements for avoidance and protection from thermal

hydraulic instabilities to be consistent with the Boiling Water Reactor

(BWR) Owner's Group long-term solution Option 1-D described in the

Licensing Topical Report, ``BWR Owner's Group Long-Term Stability

Solutions Licensing Methodology, NEDO-31960 June 1991'' and NEDO-31960,

Supplement 1, dated March 1992. NEDO-31960 and NEDO-31960, Supplement

1, were accepted by the NRC staff in a letter to L.A. England (BWR

Owner's Group) dated July 12, 1993.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed amendment will not involve a significant

increase in the probability or consequences of an accident

previously evaluated. The implementation of BWR Owner's Group long

term stability solution Option 1-D at Vermont Yankee does not modify

the assumptions contained in the existing accident analysis. The use

of an exclusion region and the operator actions required to avoid

and minimize operation inside the region do not increase the

possibility of an accident. Conditions of operation outside of the

exclusion region are within the analytical envelope of the existing

safety analysis. The operator action requirement to exit the

exclusion region upon entry minimizes the possibility of an

oscillation occurring. The actions to drive control rods and/or to

increase recirculation flow to exit the region are maneuvers within

the envelope of normal plant evolutions. The flow biased scram has

been analyzed and will provide automatic fuel protection in the

event of an instability. Thus, each proposed operating requirement

provides defense in depth for protection from an instability event

while maintaining the existing assumptions of the accident analysis.

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

new or different kind of accident from an accident previously

evaluated. As stated in 1), the proposed operating requirements

either mandate operation within the envelope of existing plant

operating conditions of force specific operating maneuvers within

those carried out in normal operation. Since operation of the plant

with all of the proposed requirements are within the existing

operating basis, an unanalyzed accident will not be created through

implementation of the proposed change.

3. The proposed amendment will not involve a significant

reduction in a margin of safety. Each of the proposed requirements

for plant thermal hydraulic stability provides a means for fuel

protection. The combination of avoiding possible unstable conditions

and the automatic flow biased reactor scram provides an in depth

means for fuel protection. Therefore, the individual or combination

of means to avoid and suppress an instability supplements the margin

of safety.

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

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

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

amendment request involves no significant hazards consideration.

[[Page 508]]

Local Public Document Room location: Brooks Memorial Library, 224

Main Street, Brattleboro, Vermont 05301

Attorney for licensee: John A. Ritsher, Esquire, Ropes and Gray,

One International Place, Boston, Massachusetts 02110-2624

NRC Project Director: Walter R. Butler

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

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

Date of amendment request: November 29, 1994

Description of amendment request: Virginia Electric and Power

Company plans to insert fuel assemblies containing fuel rods, guide

thimble tubes, instrumentation tubes, and mid-span grids fabricated

with Westinghouse Electric Corporation's (Westinghouse's) advanced

zirconium alloy material, ZIRLO, into the Surry Units 1 and 2 reactors,

beginning with Cycle 14 at each unit. In the current fuel design, these

components are fabricated from Zircaloy-4.

Because the Technical Specifications define the fuel rod cladding

material as Zircaloy-4, implementation of this material change requires

changes to the Technical Specifications. Technical Specification

5.3.A.1 is being modified to allow the use of either Zircaloy-4 or

ZIRLO fuel rod cladding, and an additional reference for the

calculation of the heat flux hot channel factor for loss-of-coolant-

accident evaluations of fuel with ZIRLO cladding is being defined in

Technical Specification 6.2. The use of the ZIRLO fabricated guide

thimble tubes, instrumentation tubes, and mid-span grids does not

require changes to the Technical Specifications.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

Specifically, operation of Surry Power Station in accordance

with the Technical Specifications changes will not:

1. Involve a significant increase in the probability or

consequence of an accident previously evaluated. The Surry fuel

assemblies containing fuel rods, guide thimble tubes,

instrumentation tubes and mid-span grids fabricated with ZIRLO alloy

meet the same fuel assembly and fuel rod design bases as the current

fuel assemblies fabricated with Zircaloy-4 components. In addition,

the 10 CFR 50.46 criteria will be applied to the fuel rods, guide

thimble tubes, instrumentation tubes and mid-span grids fabricated

with ZIRLO alloy. The use of these fuel assemblies will not result

in a change to the Surry Units 1 and 2 reload design and safety

analysis limits. The ZIRLO alloy is similar in chemical composition

to Zircaloy-4, and also has physical and mechanical properties

similar to those of Zircaloy-4. Thus the cladding integrity is

maintained and the structural integrity of the fuel assembly is not

affected. The ZIRLO clad fuel rods improve corrosion resistance and

dimensional stability. Since the dose predictions in the safety

analyses are not sensitive to the fuel rod cladding material changes

as specified in this report, the radiological consequences of

accidents previously evaluated in the safety analyses remain valid.

Therefore, neither the probability of occurrence nor the

consequences of any accident previously evaluated is significantly

increased.

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

from any accident previously identified, since the Surry Units 1 and

2 fuel assemblies containing fuel rods, guide thimble tubes,

instrumentation tubes and mid-span grids fabricated with ZIRLO alloy

will satisfy the same design bases used for previous fuel regions

containing Zircaloy-4 components. Since the original design criteria

are being met, the fuel rods, guide thimble tubes, instrumentation

tubes and mid-span grids fabricated with ZIRLO alloy will not be

initiators for any new accident. Applicable design and performance

criteria will continue to be met and no single failure mechanisms

have been created. In addition, the use of these fuel assemblies

does not involve any alteration to plant equipment or procedures

which would introduce any new or unique operational modes or

accident precursors. Therefore, the possibility for a new or

different kind of accident from any accident previously evaluated is

not created.

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

Surry Units 1 and 2 fuel assemblies containing fuel rods, guide

thimble tubes, instrumentation tubes and mid-span grids fabricated

with ZIRLO alloy do not change the Surry Units 1 and 2 reload design

and safety analysis limits. The use of fuel assemblies containing

fuel rods, guide thimble tubes, instrumentation tubes and mid-span

grids fabricated with ZIRLO alloy will take into consideration the

normal core operating conditions allowed in the Technical

Specifications. For each cycle reload core these fuel assemblies

will be specifically evaluated using approved reload design methods

and approved fuel rod design models and methods. This will include

consideration of the core physics analysis peaking factors and core

average linear heat rate effects. Analyses or evaluations will be

performed each cycle to confirm that the 10 CFR 50.46 criteria will

be met for the use of fuel with fuel rods, guide thimble tubes,

instrumentation tubes and mid-span grids fabricated with ZIRLO

alloy. Therefore, the margin of safety as defined in the Bases to

the Surry Units 1 and 2 Technical Specifications is not

significantly reduced.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Swem Library, College of

William and Mary, Williamsburg, Virginia 23185.

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

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

Virginia 23219.

NRC Project Director: Mohan C. Thadani, Acting

Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee

Nuclear Power Plant, Kewaunee County, Wisconsin

Date of amendment request: December 2, 1994

Description of amendment request: The proposed amendment would

revise Kewaunee Nuclear Power Plant (KNPP) Technical Specification (TS)

3.2 by eliminating the requirements for the charging pumps, high

concentration boric acid in the boric acid storage tanks (BASTs), the

boric acid transfer pumps, and boric acid heat tracing. Changes to TS

3.3 and Table TS 3.5.3 are also being proposed to add requirements

associated with the emergency core cooling system (ECCS) accumulators,

remove the requirements associated with the boric acid storage tanks,

and to increase the minimum required boron concentration in the

refueling water storage tank (RWST). Additionally, the surveillance

requirements involving the BASTs, associated valves and heat tracing

located in Table TS 4.1-1, Table TS 4.1-2 and Section 4.5 would be

eliminated. Supporting analysis for the limiting design basis accident

conditions have been performed using the proposed minimum RWST boron

concentration of 2400 ppm.

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:

Significant Hazards Determination for Proposed Changes to

Technical Specification (TS) 3.2 and Table TS 3.5-3.

The proposed changes were reviewed in accordance with the

provisions of 10 CFR 50.92 to show no significant hazards exist. The

proposed changes will not:

1) Involve a significant increase in the probability or

consequences of an accident previously evaluated;

Neither the charging pumps, the high concentration boric acid,

the BASTs, the boric acid transfer pumps nor the boric acid heat

tracing system are accident initiators. Therefore, a change to these

systems will not significantly increase the probability of an

accident previously evaluated. The effect of a reduction in initial

safety injection boron concentration on the accident analysis was

evaluated. The limiting accidents were the Large-Break Loss-of-

Coolant Accident [[Page 509]] (LOCA) and the Steam Line Break (SLB)

event. A decrease in the initial safety injection boron

concentration from 20,000 ppm to 2400 ppm will not adversely affect

the Large or Small-Break Loss-of-Coolant Accident analysis because

the evaluation models used in analyzing these accidents do not take

credit for the high concentration boric acid stored in the BASTs.

However, the evaluation models did take credit for boron in

maintaining the long term post LOCA reactor core sub-critical. An

analysis was performed which concluded that the inventory contained

in the BASTs would not be required provided the minimum RWST boron

concentration was increased to 2400 ppm. The SLB event is the other

design basis event that could be affected by the proposed

elimination of the high boron concentration BASTs as a source of

safety injection fluid. Analyses have been performed which conclude

that the BASTs are not required and that a minimum RWST boron

concentration of only 1950 ppm is sufficient to provide adequate

protection for the SLB event although 2400 ppm will be maintained to

address post-LOCA subcriticality thus providing further safety

margin. The results of these analyses indicate that the departure

from nucleate boiling (DNB) design basis continues to be met. (A

minimum Departure from Nucleate Boiling Ratio (DNBR) of 1.45 can be

maintained throughout the event.) Finally, the containment pressure

and temperature remains within the acceptable containment design

limits. Since these criteria have been satisfied, there will be no

adverse effect on the health and safety of the public and the

consequences of any accident previously evaluated have not

significantly increased.

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

from any accident previously evaluated;

Neither the charging pumps, the removal of the BASTs from

initial SI pump injection, nor the elimination of both the boric

acid transfer pumps and the boric acid heat tracing system as

safety-related components would create the possibility of a new or

different kind of accident from any accident previously evaluated.

Furthermore, the reactivity control function of the boron in the

CVCS and SI systems is not being changed. Therefore, the proposed

changes will not adversely affect the health and safety of the

public or create the possibility of a new or different kind of

accident from any accident previously evaluated.

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

The reduction in the initial concentration of boron injected

into the reactor coolant system for accident mitigation has been

analyzed. These analyses conclude that all applicable criteria for a

LOCA are satisfied. A decrease in the initial safety injection boron

concentration from 20,000 ppm to 2400 ppm will not adversely effect

the Large-or Small-Break Loss-of-Coolant Accident analysis because

the evaluation models used in analyzing these accidents do not take

credit for the high concentration boric acid stored in the BASTs.

However, in order to maintain the long term post LOCA reactor core

sub-critical, a minimum RWST boron concentration of 2400 ppm is

required. To meet this requirement, the RWST boron concentration is

being raised to 2400 ppm. All criteria of 10 CFR 50.46 can be

achieved for both the Large or Small-Break LOCA with no BASTs and

2400 ppm boron in the RWST. Since all criteria of 10 CFR 50.46 are

satisfied, there is no adverse effect on the health and safety of

the public and there is not a significant reduction in the margin of

safety for these casualties.

Since both the core response and the containment response can be

limiting in the SLB event, both were considered in the boron

concentration reduction analysis. This analysis concludes that a

minimum RWST boron concentration of 1950 ppm is sufficient to

provide adequate protection for the SLB event, although a 2400 ppm

boron solution will be maintained to provide protection for the post

LOCA concerns. Since the containment pressure and temperature

remains within the acceptable containment design limits, and a

minimum DNBR of 1.45 can be maintained throughout the event, there

is not a significant reduction in the margin of safety for this

event and therefore there is no adverse effect on the health and

safety of the public.

These proposed changes involve the conversion of the TS to Word

Perfect format now being used at WPSC. Minor typographical errors

and format inconsistencies were corrected. These proposed changes

are administrative in nature; accordingly, these proposed changes do

not involve a significant hazards consideration.

Additionally, the proposed changes are similar to example

C.2.e.(i) in 51 FR 7751. Example C.2.e.(i) states that changes which

are purely administrative in nature; i.e., to achieve consistency

throughout the Technical Specifications, correct an error, or a

change in nomenclature, are not likely to involve a significant

hazard.

Significant Hazards Determination for Proposed Changes to Table

TS 4.1-1, ``Minimum Frequencies for Checks, Calibrations and Test of

Instrument Channels'' and Table TS 4.1-2 ``Minimum Frequencies for

Sampling Tests''

The proposed changes were reviewed in accordance with the

provisions of 10 CFR 50.92 to show no significant hazards exist. The

proposed changes will 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 the margin of safety.

The above listed surveillance requirements insure BAST

operability. The BASTs will no longer be relied upon as a source of

boron for safety injection, and will serve no safety related

function. Whether the BASTs are operable or not will have no effect

on plant safety. Therefore, elimination of the surveillance

requirements which insure BAST operability is possible without any

adverse effect on the health and safety of the public and presents

no significant hazards.

Significant Hazards Determination for Proposed Changes to

Technical Specification TS 3.3 and Section 4.5.

The proposed changes were reviewed in accordance with the

provisions of 10 CFR 50.92 to show no significant hazards exist. The

proposed changes will not:

1) Involve a significant increase in the probability or

consequences of an accident previously evaluated.

Neither the RWST, the boron solution contained within the RWST

nor valves SI-3, SI-4A/B are accident initiators. Therefore, a

change to these systems will not significantly increase the

probability of an accident previously evaluated. The effect of a

reduction in initial Safety Injection boron concentration on the

accident analysis was evaluated. The limiting accidents were the

Large-Break Loss-of-Coolant Accident (LOCA) and the Steam Line Break

(SLB) event. A decrease in the initial safety injection boron

concentration from 20,000 ppm to 2400 ppm will not adversely effect

the Large or Small-Break Loss-of-Coolant Accident analysis because

the evaluation models used in analyzing these accidents do not take

credit for the high concentration boric acid stored in the BASTs.

However, the evaluation models did take credit for boron in

maintaining the long term post LOCA reactor core sub-critical. An

analysis was performed which concluded that the BASTs could be

eliminated provided the minimum RWST boron concentration was

increased to 2400 ppm. The SLB event is the other design basis event

that could be affected by the proposed elimination of the high

concentration BASTs as a safety-related source for reactivity

control injection fluid. However, analyses have been performed which

conclude that a minimum RWST boron concentration of only 1950 ppm is

sufficient to provide adequate protection for the SLB event although

2400 ppm will be maintained to address post-LOCA subcriticality thus

providing further safety margin. The results of these analyses

indicate that the departure from nucleate boiling (DNB) design basis

continues to be met. (A minimum Departure from Nucleate Boiling

Ratio (DNBR) of 1.45 can be maintained throughout the event.)

Furthermore, maintaining the suction of the SI pumps to the RWST

with valves SI-4A or SI-4B open with power removed places the system

in a normal SI sequence and eliminates the requirement to switch

suction from the BASTs to the RWST. This eliminates a potential

failure mechanism and increases the overall reliability of the ECCS

system. Finally, the containment pressure and temperature remains

within the acceptable containment design limits.

Since these criteria have been satisfied, there will be no

adverse effect on the health and safety of the public and the

consequences of any accident previously evaluated have not

significantly increased.

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

from any accident previously evaluated.

This change to the Technical Specifications allows use of 2400

ppm boron for safety injection. SI pump suction would be directly

from the RWST. This eliminates [[Page 510]] the necessity of

shifting suction from the BASTs to the RWST, reducing the complexity

of the operation. Since the pumps remain connected to the RWST

throughout the injection phase, there is no possibility of a new or

different kind of accident from any accident previously evaluated.

Neither the reduction in initial boron concentration for safety

injection, nor the increase in the boron concentration in the RWST

would create the possibility of a new or different kind of accident

from any accident previously evaluated.

Lastly, the reactivity control function of the boron in the CVCS

and SI systems is not being changed. Therefore, the proposed changes

will not adversely affect the health and safety of the public or

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

any accident previously evaluated.

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

The change in concentration of boron injected into the primary

system for accident mitigation has been analyzed. These analyses

conclude that all applicable criteria for a LOCA are satisfied. A

change in safety injection boron concentration to 2400 ppm will not

adversely affect the Large or Small-Break LOCA analysis because the

evaluation model codes used in analyzing these accidents did not

take credit for boron. However, a minimum RWST boron concentration

of 2400 ppm is required to maintain long term post LOCA reactor core

sub-criticality. To meet this requirement, the RWST minimum boron

concentration is being raised to 2400 ppm. All criteria of 10 CFR

50.46 can be achieved for both the Large or Small-Break LOCA with

2400 ppm boron in the RWST. Since all criteria of 10 CFR 50.46 are

satisfied, there is no adverse effect on the health and safety of

the public and there is not a significant reduction in the margin of

safety for these casualties.

Since both the core response and the containment response can be

limiting in the SLB event, both were considered in the boron

concentration reduction analysis. Although a minimum RWST boron

concentration of 1950 ppm is sufficient to provide adequate

protection for the SLB event, a 2400 ppm boron solution will be

maintained to provide protection for the post large break LOCA

concerns. Since the containment pressure remains below the design

pressure, and a minimum DNBR of 1.45 can be maintained throughout

the event, there is not a significant reduction in the margin of

safety for this event.

These proposed changes involve the conversion of the TS to Word

Perfect format now being used at WPSC. Minor typographical errors

and format inconsistencies were corrected. These proposed changes

are administrative in nature; accordingly, these proposed changes do

not involve a significant hazards consideration.

Additionally, the proposed changes are similar to example

C.2.e.(i) in 51 FR 7751. Example C.2.e.(i) states that changes which

are purely administrative in nature; i.e., to achieve consistency

throughout the Technical Specifications, correct an error, or a

change in nomenclature, are not likely to involve a significant

hazard.

Significant Hazards Determination for Proposed Changes to

Technical Specification (TS) Section 4.5 ``Emergency Core Cooling

System and Containment Air Cooling System Tests.''

The proposed changes were reviewed in accordance with the

provisions of 10 CFR 50.92 to show no significant hazards exist. The

proposed changes will 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 the margin of safety.

The above listed surveillance requirements insure BAST

operability. The BASTs will no longer be relied upon as a source of

boron for safety injection, and will serve no safety related

function. Whether the BASTs are operable or not will have no effect

on plant safety. Therefore, elimination of the surveillance

requirements which insure BAST operability is possible without any

adverse effect on the health and safety of the public and presents

no significant hazards.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: University of Wisconsin

Library Learning Center, 2420 Nicolet Drive, Green Bay, Wisconsin

54301.

Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner,

P. O. Box 1497, Madison, Wisconsin 53701-1497.

NRC Project Director: Leif J. Norrholm

Previously Published Notices Of Consideration Of Issuance Of

Amendments ToFacility Operating Licenses, Proposed No Significant

Hazards Consideration Determination,And Opportunity For A Hearing

The following notices were previously published as separate

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

published as individual notices either because time did not allow the

Commission to wait for this biweekly notice or because the action

involved exigent circumstances. They are repeated here because the

biweekly notice lists all amendments issued or proposed to be issued

involving no significant hazards consideration.

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

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

of the original notice.

Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear

One,Unit No. 2, Pope County, Arkansas

Date of amendment request: November 29, 1994

Brief description of amendment request: The proposed amendment

would delete requirements to perform the full complement of steam

generator surveillances as outlined in the technical specifications

(TSs) when the steam generators are subjected to special inspections

that are in addition to inspections that are required by the TSs.

Date of individual notice in the Federal Register: December 5, 1994

(59 FR 62416)

Expiration date of individual notice: January 4, 1995

Local Public Document Room location: Tomlinson Library, Arkansas

Tech University, Russellville, Arkansas 72801

Notice Of Issuance Of Amendments To Facility Operating Licenses

During the period since publication of the last biweekly notice,

the Commission has issued the following amendments. The Commission has

determined for each of these amendments that the application complies

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

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

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

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

forth in the license amendment.

Notice of Consideration of Issuance of Amendment to Facility

Operating License, Proposed No Significant Hazards Consideration

Determination, and Opportunity for A Hearing in connection with these

actions was published in the Federal Register as indicated.

Unless otherwise indicated, the Commission has determined that

these amendments satisfy the criteria for categorical exclusion in

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

no environmental impact statement or environmental assessment need be

prepared for these amendments. If the Commission has prepared an

environmental assessment under the special circumstances provision in

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

it is so indicated.

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

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

related letter, Safety Evaluation and/or Environmental

[[Page 511]] Assessment as indicated. All of these items are available

for public inspection at the Commission's Public Document Room, the

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

public document rooms for the particular facilities involved.

Arizona Public Service Company, et al., Docket No. STN 50-528, Palo

Verde Nuclear Generating Station, Unit No. 1, Maricopa County,

Arizona

Date of application for amendment: November 22, 1994

Brief description of amendment: The amendment adds a note to

Technical Specification Table 3.7-2. The note allows continuous

operation of Unit 1 during Cycle 5 at 100-percent maximum steady state

power with one main steam safety valve inoperable per steam generator.

This note applies only during the current fuel cycle (Cycle 5) for Unit

1.

Date of issuance: December 19, 1994

Effective date: December 19, 1994

Amendment No.: 87

Facility Operating License No. NPF-41: The amendment revised the

Technical Specifications.

Public comments requested as to proposed no significant hazards

consideration: Yes (59 FR 61907, dated December 2, 1994). The 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 December 19, 1994, but stated 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 significant hazards

consideration is contained in a Safety Evaluation dated December 19,

1994.

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

Local Public Document Room location: Phoenix Public Library, 12

East McDowell Road, Phoenix, Arizona 85004

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

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

Date of application for amendment: October 7, 1994

Brief description of amendment: The amendment revises the

introduction to TS Section 6.9.3.3 to require the approved revision

number for the referenced analytical methods to be listed in the Core

Operating Limits Report. The methodology referenced in 6.9.3.3.b.f (XN-

NF-82-49(A)) has been updated to clarify that all supplements are

included. New methodologies ANF-89-151(A) and EMF-92-081(A) will be

added to TS Section 6.9.3.3.b.

Date of issuance: December 12, 1994

Effective date: December 12, 1994

Amendment No.: 154

Facility Operating License No. DPR-23. Amendment revises the

Technical Specifications.

Date of initial notice in Federal Register: November 9, 1994 (59 FR

55868)The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated December 12, 1994.No significant hazards

consideration comments received: No

Local Public Document Room location: Hartsville Memorial Library,

147 West College, Hartsville, South Carolina 29550

Consolidated Edison Company of New York, Docket No. 50-247, Indian

PointNuclear Generating Unit No. 2, Westchester County, New York

Date of application for amendment: February 18, 1994, as

supplemented by letters dated June 3, 1994, November 1, 1994, December

2, 1994, December 14, 1994 and December 16, 1994.

Brief description of amendment: The amendment revises surveillance

intervals for the Vapor Containment Sump Discharge Flow and Temperature

Channel, the Loss of Power Undervoltage and Degraded Voltage Relays,

and the Control Rod Protection System Trip to accommodate a 24-month

refueling cycle. In addition it changes the trip setpoint for the

Control Rod Protection System Trip. These revisions are being made in

accordance with the guidance provided by Generic Letter 91-04,

``Changes in Technical Specification Surveillance Intervals to

Accommodate a 24-Month Fuel Cycle.''

Date of issuance: December 20, 1994

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

30 days.

Amendment No.: 179

Facility Operating License No. DPR-26: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: April 28, 1994 (59 FR

22003) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated December 20, 1994.No significant

hazards consideration comments received: No

Local Public Document Room location: White Plains Public Library,

100 Martine Avenue, White Plains, New York 10610.

Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388

Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County,

Pennsylvania

Date of application for amendments: July 19, 1994

Brief description of amendments: These amendments change Technical

Specification 3.1.5 for each unit for the standby liquid control system

(SLCS) to remove the operability requirement for the SLCS while the

plant is in Operational Condition 5 (refueling) with any control rod

withdrawn, and to delete the 18-month system surveillance requirement

(Surveillance Requirement 4.1.5.d.3).

Date of issuance: December 20, 1994

Effective date: December 20, 1994

Amendment Nos.: 136 and 106

Facility Operating License Nos. NPF-14 and NPF-22. The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: August 17, 1994 (59 FR

42344)The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated December 20, 1994.No significant

hazards consideration comments received: No

Local Public Document Room location: Osterhout Free Library,

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

Pennsylvania 18701.

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: July 22, 1994

Brief description of amendments: The amendment removes the

surveillance frequency details regarding 10 CFR Part 50, Appendix J,

Types B and C testing from the Technical Specifications.

Date of issuance: December 19, 1994

]Effective date: December 19, 1994

Amendment Nos. 83 and 44

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

revised the Technical Specifications.

Date of initial notice in Federal Register: September 14, 1994 (59

FR 47180) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated December 19, 1994.No significant

hazards consideration comments received: No

Local Public Document Room location: Pottstown Public Library, 500

[[Page 512]] High Street, Pottstown, Pennsylvania 19464.

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

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

Date of application for amendment: October 7, 1994

Brief description of amendment: The amendment revises Technical

Specification (TS) 4.6E.4 and the associated Bases to establish that

the manual cycling of reactor coolant system safety/relief valves

(SRVs) during plant startups is to be accomplished within 12 hours

after steam pressure and flow are adequate to perform the testing. TS

4.6E.4 had previously required that this testing be performed within 12

hours of continuous power operation at a reactor steam dome pressure of

at least 940 psig. The amendment also makes several editorial changes

to clarify the intent of TSs involving SRV testing and performance

requirements.

Date of issuance: December 16, 1994

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

30 days.

Amendment No.: 219

Facility Operating License No. DPR-59: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: November 9, 1994 (59 FR

55889)The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated December 16, 1994.No significant hazards

consideration comments received: No

Local Public Document Room location: Reference and Documents

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

York 13126.

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

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

County, New Jersey Date of application for amendments: September 9,

1994

Brief description of amendments: The amendments revise the

Technical Specification surveillance requirements regarding visual

inspection of snubbers and are consistent with the guidance provided in

Generic Letter 90-09, ``Alternative Requirements for Snubber Visual

Inspection Intervals and Corrective Actions.''

Date of issuance: December 12, 1994

Effective date: December 12, 1994

Amendment Nos. 161 and 142

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

revised the Technical Specifications.

Date of initial notice in Federal Register: November 9, 1994 (59 FR

55889)The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated December 12, 1994.No significant

hazards consideration comments received: No

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

West Broadway, Salem, New Jersey 08079

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

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

County, New Jersey Date of application for amendments: March 28,

1994, as supplemented June 1, 1994, and August 24, 1994

Brief description of amendments: The amendments revise the

sustained degraded voltage relay trip setpoint and the allowable value

due to changes in the switchyard configuration.

Date of issuance: December 14, 1994

Effective date: December 14, 1994

Amendment Nos. 162 and 143

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

revised the Technical Specifications.

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

29633) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated December 14, 1994.No significant

hazards consideration comments received: No

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

West Broadway, Salem, New Jersey 08079

Tennessee Valley Authority, Docket Nos. 50-259, 50-260, and 50-296,

Browns Ferry Nuclear Plant, Units 1, 2, and 3, Limestone County,

Alabama

Date of application for amendments: October 7, 1993 (TS 313)

Brief description of amendments: The changes include the addition

of the high range primary containment radiation monitors and recorders

and the wide range gaseous effluent radiation recorder and monitor,

which were installed at the Browns Ferry facility in response to NUREG

0737 ``Clarification of TMI Action Plan Requirements'' and GL 83-36,

into the technical Specifications (TS) for Units 1 and 3. Similar

changes to the Unit 2 TS were issued previously (Amendment Nos. 125 and

171). The amendment also clarifies that the high range primary

containment radiation recorders and monitors are both part of the

instrument loop. The amendment contains administrative typographical

changes which provide consistency for the TS tables and footnotes for

Units 1 and 3.

Date of issuance: December 21, 1994

Effective Date: December 21, 1994

Amendment Nos.: 214, 230, 187

Facility Operating License Nos. DPR-33, DPR-52 and DPR-68:

Amendments revised the Technical Specifications.

Date of initial notice in Federal Register: December 22, 1993 (58

FR 67863)The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated December 21, 1994.No significant

hazards consideration comments received: None

Local Public Document Room location: Athens Public library, South

Street, Athens, Alabama 35611

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: November 11, 1994, as supplemented by

letter dated November 16, 1994.

Brief description of amendments: The proposed amendment would

modify Comanche Peak Steam Electric Station Technical Specification

Table 4.8-1, ``Diesel Generator Test Schedule,'' by excluding two valid

failures of the Unit 2 Train B diesel generator from contributing

towards an accelerated test schedule.

Date of issuance: December 9, 1994

Effective date: December 9, 1994

Amendment Nos.: Unit 1 - Amendment No. 33; Unit 2 - Amendment No.

19

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

revised the Technical Specifications.Public comments requested as to

proposed no significant hazards consideration: Yes (59 FR 69399, dated

November 23, 1994). The 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 December 23,

1994, but stated that, if the Commission makes a final no significant

hazards consideration determination, any such hearing would take place

after issuance of the amendments.

The Commission's related evaluation of the amendments, finding of

exigent circumstances, and final determination of no significant

hazards consideration is contained in a Safety Evaluation dated

December 9, 1994.

Local Public Document Room location: University of Texas at

Arlington library, Government [[Page 513]] Publications/Maps, 702

College, P.O. Box 19497, Arlington, Texas 76019.

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: March 29, 1994

Brief description of amendments: These amendments revise Point

Beach Nuclear Plant Technical Specification 15.3.2, ``Chemical and

Volume Control System,'' by eliminating the necessity for high

concentration boric acid and removing the operability requirements for

the associated heat tracing. The basis for Section 15.3.2 and

applicable surveillances in Table 15.4.1-2 are also revised to support

the above changes.

Date of issuance: December 12, 1994

Effective date: Date of issuance, to be implemented within 45 days.

Amendment Nos.: 158 & 162

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

revised the Technical Specifications.

Date of initial notice in Federal Register: July 20, 1994 (59 FR

37091) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated December 12, 1994.No significant

hazards consideration comments received: No.

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

Sixteenth Street, Two Rivers, Wisconsin 54241.

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 12, 1994

Brief description of amendments: These amendments revise Point

Beach Nuclear Plant Technical Specification (TS) 15.3.3, ``Emergency

Core Cooling System, Auxiliary Cooling Systems, Air Recirculation Fan

Coolers, and Containment Spray,'' by incorporating allowed outage times

similar to those contained in NUREG-1431, Revision 0, ``Westinghouse

Owner's Group Improved Standard Technical Specifications,'' and by

clarifying the operability requirements for the service water pumps.

The changes also clarify the completion times for placing a unit in hot

or cold shutdown, if a limiting condition for operation cannot be met.

Date of issuance: December 21, 1994

Effective date: Date of issuance, to be implemented within 45 days

Amendment Nos.: 159 & 163

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

revised the Technical Specifications.

Date of initial notice in Federal Register: October 24, 1994 (59 FR

53844)The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated December 21, 1994.No significant

hazards consideration comments received: No.

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

Sixteenth Street, Two Rivers, Wisconsin 54241.

Dated at Rockville, Maryland, this 27th day of December 1994.

For The Nuclear Regulatory Commission.

Elinor G. Adensam,

Acting Director, Division of Reactor Projects - III/IV, Office of

Nuclear Reactor Regulation.

[Doc. 95-5 Filed 1-3-95

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