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 August 19, 1996, through August 29, 1996. The last biweekly notice was published on August 28, 1996 (61 FR 44353).

Federal RegisterSep 11, 1996

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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

[[Page 47974]]

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 August 19, 1996, through August 29, 1996.

The last biweekly notice was published on August 28, 1996 (61 FR

44353).

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 Chief, Rules

Review and Directives Branch, Division of Freedom of Information and

Publications Services, Office of Administration, U.S. Nuclear

Regulatory Commission, Washington, DC 20555-0001, 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 October 11, 1996, the licensee may file a request for a hearing

with respect to issuance of the amendment to the subject facility

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

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

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

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

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

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

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

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

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

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

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

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

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

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

Secretary or the designated Atomic Safety and Licensing Board will

issue a notice of a hearing or an appropriate order.

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

shall set forth with particularity the interest of the petitioner in

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

the proceeding. The petition should specifically explain the reasons

why intervention should be permitted with particular reference to the

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

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

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

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

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

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

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

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

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

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

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

the specificity requirements described above.

Not later than 15 days prior to the first prehearing conference

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

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

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

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

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

brief explanation of the bases of the contention and a concise

statement of the alleged facts or expert opinion which support the

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

contention at the hearing. The petitioner must also provide references

to those specific sources and documents of which the petitioner is

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

facts or expert opinion. Petitioner must provide sufficient information

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

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

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

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

petitioner who fails to file such a supplement which satisfies these

requirements with respect to at least one contention will not be

permitted to participate as a party.

Those permitted to intervene become parties to the proceeding,

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

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

hearing, including the opportunity to present evidence and cross-

examine witnesses.

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

determination on the issue of no significant hazards consideration. The

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

If the final determination is that the amendment request involves

no significant hazards consideration, the Commission may issue the

amendment and make it immediately effective,

[[Page 47975]]

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 20555-0001, 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 20555-0001,

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.

Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318,

Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert

County, Maryland

Date of amendments request: August 1, 1996

Description of amendments request: The amendment will allow use of

blind flanges during MODES 1-4 in the Calvert Cliffs Units 1 and 2

Containment Purge Systems. These flanges will establish integrity in

Mode 5, prior to entering Mode 4, and maintain it in Modes 1-4,

functions presently served by the valve.

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. Would not involve a significant increase in the probability

or consequences of an accident previously evaluated.

The purpose of the Containment Purge System is to provide

ventilation for the containment while in a shutdown condition.

Valves, which are disabled in the shut position in Modes 1-4, may be

opened in Modes 5 and 6 to allow air flow, are provided in the

supply and exhaust piping, and are automatically shut on a

Containment Radiation Signal to prevent release of radioactive

material in the event of a fuel handling incident. Manual operation

is also provided. In Modes 1-4, the valves are kept shut to provide

containment integrity to withstand a presumed increase in

containment pressure in the event of a loss-of-coolant accident. The

proposed change will allow blind flanges to serve in place of the

purge valves in Modes 1-4 by blocking off the purge penetration on

both the supply and exhaust sides. The blind flanges will provide

the same level of containment integrity previously provided by the

purge valves. The revised Technical Specifications will continue to

verify containment building leakage is maintained within the

allowable limits by requiring the performance of a 10 CFR Part 50,

Appendix J, Type B, leakage test on the blind flanges. The outside

valve in each containment purge penetration will be removed and the

inside valves will be left in place. The remaining inside valves

will no longer by required to provide containment integrity in Modes

1-4. Only one of each pair of valves was credited for containment

closure (Modes 5 and 6); therefore, removing the outside valves and

the associated automatic closure signals is not a modification of

the required capability to close the penetration. The inside valves

will maintain their current safety function to close containment (if

needed) by closing either on a Containment Radiation Signal (Mode 6)

or manually (Modes 5 and 6). The Technical Specification

surveillances associated with the purge valves will be changed to

reflect the proposed modification to the plant. Since the blind

flanges will limit radiological releases in Modes 1-4, and the purge

valves will limit radiological releases in Modes 5 and 6, the

proposed change will not increase the consequences of an accident

previously evaluated.

The Containment Purge System is not an accident initiator but

acts to limit the consequences of accidents. The system will provide

containment isolation in Modes 1-4 as before, and the inside valves

will still be available to close in Modes 5 and 6. Therefore, the

proposed change does not increase the probability of an accident

previously evaluated.

As stated above, the proposed changes do not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

2. Would not create the possibility of a new or different type

of accident from any accident previously evaluated.

This requested change does not involve a significant alteration

of the operation of the plant, and no new accident initiation

mechanism is created by the modification. Four purge valves per unit

currently provide containment closure in Modes 5 and 6. The outside

valve in the supply and the exhaust lines will be removed to allow

for installation of a blind flange in each line. The remaining

supply and exhaust valves inside containment will continue to

provide containment closure. The function currently performed by the

four purge valves in Modes 1, 2, 3 and 4 will be performed by the

blind flanges. Other, similar, blind flanges have been in service in

the plant for a number of years, and have proven reliable. The

Technical Specification surveillances associated with the testing of

the purge valves and flanges will be changed to reflect the proposed

modification to the plant. Therefore, this change does not create

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

accident previously evaluated.

3. Would not involve a significant reduction in the margin of

safety.

The valves in the Containment Purge System currently provide

containment integrity during Modes 1, 2, 3 and 4, and containment

closure during Modes 5 and 6. The function currently performed by

the purge valves in Modes 1, 2, 3 and 4 will be performed by the

blind flanges. Because of their design and mounting method, the

blind flanges will perform the containment integrity function as

well as, or better than, the purge valves. In Modes 1-4, the double

o-rings in the blind flanges will provide single-failure protection

similar to the other existing Type B penetrations. The established

allowable containment building leakage rate will be maintained by

the implementation of a requirement to perform 10 CFR Part 50,

Appendix J, Type B, leakage rate on the installed blind flanges. The

outside valve in each purge containment penetration will be removed.

Single failure is not assumed in the fuel handling accident

analysis, therefore, removing the outside valves and their

Containment Radiation Signal channels is not a modification of the

required capability to close the penetration. The remaining inside

valves will continue to provide automatic and manual containment

closure in Mode 6 to mitigate the effects of a fuel handling

accident. The Technical Specification surveillances associated with

purge valve testing will be changed to reflect the proposed

modification to the plant. 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 50.92(c) are

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

amendments request

[[Page 47976]]

involves no significant hazards consideration.

Local Public Document Room location: Calvert County Library, Prince

Frederick, Maryland 20678.

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

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

NRC Project Director: Jocelyn A. Mitchell, Acting Director

Consolidated Edison Company of New York, Docket No. 50-247, Indian

Point Nuclear Generating Unit No. 2, Westchester County, New York

Date of application for amendment: August 7, 1996

Brief description of amendment: The amendment proposes revising the

Technical Specifications (TSs) to allow the use of 10 CFR Part 50,

Appendix J, Option B, Performance-Based Containment Leakage Rate

Testing. This performance-based Option B may be used as an alternative

to the requirements in Appendix J, ``Primary Reactor Containment

Leakage Testing for Water-Cooled Power Reactors,'' of 10 CFR Part 50.

To implement Option B to Appendix J, the amendment proposes modifying

TSs to eliminate reference to the prescriptive Appendix J requirements

and instead reference NRC Regulatory Guide 1.163, ``Performance-Based

Containment Leak-Test Program.'' The amendment also proposes an

editorial correction to the mathematical formula minimum testing

frequency in the basis for TS 4.1.

Basis for proposed no significant hazards consideration

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

its analysis of the issue of no significant hazards consideration,

which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

For Indian Point Unit No. 2, the integrated leak rate testing

[ILRT] as-found measured leakage rate acceptance criteria is changed

from 0.75 La to 1.0 La. This change is consistent with the revised

10 CFR 50 Appendix J, NEI 94-01, ``Industry Guidelines for

Implementing Performance-Based Option of 10 CFR Part 50, Appendix

J.'' In addition, an as-found leakage rate acceptance criteria of

1.0 LA for Type A tests is consistent with the design basis and

accident analysis assumptions. The as-left acceptance criteria

remains unchanged at 0.75 La in accordance with the NEI guidance.

Therefore, prior to entering an operating mode where containment

integrity is required the as-left leakage rate will not exceed 0.75

La. The combined leakage rate for containment isolation valves

listed in Technical Specification Table 4.4-1 subject to gas or

nitrogen pressurization testing, air lock testing, and portions of

the sensitive leakage rate test which pertain to containment

penetrations and double-gasketed seals shall be less than 0.6 La.

The extensive operations and testing experience derived from

industry show that risk to the general population is generally

insensitive to changes in the allowable leakage rate. It has been

determined that the allowable containment leakage can be increased

by one to two orders of magnitude without significantly impacting

the estimates of population dose in the event of an accident.

Furthermore, the Indian Point Unit No. 2 ILRT test history provides

substantial justification for the proposed changes.

Test results demonstrate that IP2 [Indian Point 2] has a low

leakage containment and that the proposed changes would not

jeopardize the ability of the containment to maintain the leakage

rate at or below the required limits. The proposed change to

Technical Specification 4.1 Basis represent a minor editorial

correction to the mathematical formula for minimum testing frequency

which does not change the formula. Therefore, the probability and

the consequence of a design basis accident are not being increased

by the proposed changes.

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

or different kind of accident from any accident previously

evaluated.

Plant systems and components will not be operated in a different

manner as a result of the proposed Technical Specification change.

The proposed change permits a performance-based approach to

determining the leakage-rate test frequency for the containment and

containment penetrations (Type A, B, and C tests). There are no

plant modifications, or changes in methods of operation. Therefore,

the changes in testing intervals for the containment and containment

penetrations have no affect on the probability of occurrence of a

LOCA [loss-of-coolant-accident]. The Limiting Conditions for

Operation are not being changed. Changing the as-found leakage-rate

acceptance criterion to 1.0 La does not increase the probability or

consequences of an accident. Changing the test interval for the

containment and containment penetrations does not create any new

accident precursors or methods of operation. The proposed change to

Technical Specification 4.1 Basis represent a minor editorial

correction to the mathematical formula for minimum testing frequency

which does not change the formula. Therefore, the possibility for an

accident of a different type than was previously evaluated in the

safety analysis report is not created by the proposed Technical

Specification.

3. The proposed change does not involve a significant reduction

in a margin of safety.

While the proposed changes do increase the probability for

malfunction of equipment important to safety due to the longer

intervals between leakage tests, it has been estimated that the

longer test intervals will have an insignificant increase in the

overall accident risk to the public. This increase has been reviewed

and found to be acceptable by the NRC as documented in NUREG-1493

and the recent rulemaking to 10 CFR 50 Appendix J. We also agree

that this increase in accident risk is insignificant. Changing the

as-found acceptance criterion to 1.0 La does not increase the

consequences of an accident, since the accident analysis assume a

leakage rate of La for design basis accidents. The as-left Type A

test acceptance criterion remains at less than 0.75 La. Given that

the Indian Point Unit No. 2 ILRT test history show no failures

during plant life, the proposed changes should not lead to a

significant probability of creating new leakage paths or increased

leakage rates. The proposed change to Technical Specification 4.1

Basis represent a minor editorial correction to the mathematical

formula for minimum testing frequency which does not change the

formula. Therefore, the accident analysis assumptions for design

basis accidents are unaffected and the margin of safety is not

decreased by the proposed Technical Specification change.

Public Document Room location: White Plains Public Library, 100

Martine Avenue, White Plains, New York 10610.

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

Buren County, Michigan

Date of amendment request: January 18, 1996

Description of amendment request: The proposed amendment would

delete the requirement to perform inservice inspections of the primary

coolant pump (PCP) flywheels.

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 change to the Technical

Specifications would not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed change to the Technical Specifications would delete

the requirement to perform non-destructive examination of the upper

flywheel on the PCPs. The fracture mechanics analyses conducted to

support the change show that a preexisting crack sized just below

detection level will not grow to the flaw size necessary to result

in flywheel failure within the life of the plant. This analysis

conservatively assumes minimum material properties, maximum flywheel

accident speed, location of the flaw in the highest stress area and

a number of startup/shutdown cycles eight times greater than

expected. Since an existing flaw in the flywheel will not grow to

the allowable flaw size under normal operating conditions or to the

critical flaw size under LOCA [loss-of-coolant accident] conditions

over the life of the plant, elimination of inservice inspection for

such cracks during the plant's life will not involve a significant

increase in the

[[Page 47977]]

probability of an accident previously considered.

The proposed changes do not increase the amount of radioactive

material available for release or modify any systems used for

mitigation of such releases during accident conditions. Therefore,

operation of the facility in accordance with the proposed change to

the Technical Specifications would not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

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

from any previously evaluated.

The proposed change to the Technical Specifications would not

change the design, configuration, or method of operation of the

plant and therefore, operation of the facility in accordance with

the proposed change to the Technical Specifications would not create

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

previously evaluated.

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

The proposed change to the Technical Specifications would not

result in a significant reduction in the margin of safety.

Significant conservatisms have been used for calculating the

allowable flaw size, critical flaw size and crack growth rate in the

PCP flywheels. These include minimum material properties, maximum

flywheel accident speed, location of the postulated flaw in highest

stress area and a number of startup/shutdown cycles eight times

greater than expected. Since an existing flaw in the flywheel will

not grow to the maximum allowable flaw size under normal operating

conditions or to the critical flaw size under LOCA conditions over

the life of the plant, elimination of inservice inspections for such

cracks during the plant's life will not involve a significant

reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: 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 Hannon

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

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

Date of amendment request: August 8, 1996

Description of amendment request: The proposed amendments would

change the Technical Specifications (TS) of each unit to reference

updated or recently approved methodologies used to calculate cycle-

specific limits contained in the Core Operating Limits Report (COLR).

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

(1) The proposed changes will not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The proposed changes are administrative in nature, and do not

affect any system, procedure, or manipulation of any equipment which

could affect the probability or consequences of any accident.

(2) The proposed changes will not create the possibility of a

new or different kind of accident from any accident previously

evaluated.

The proposed changes are administrative in nature, and cannot

introduce any new failure mode or transient which could create any

accident.

(3) The proposed changes will not involve a significant

reduction in a margin of safety.

The proposed changes are administrative in nature, and will not

affect any operating parameters or limits which could result in a

reduction in a margin of safety.

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

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

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

proposed amendments involve no significant hazards consideration.

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

Black Street, Rock Hill, South Carolina 29730

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

South Church Street, Charlotte, North Carolina 28242

NRC Project Director: Herbert N. Berkow

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

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

Date of amendment request: June 21, 1996

Description of amendment request: The proposed amendments would

revise the term ``lifting loads'' used in Technical Specification

3.9.6b.2, Manipulator Crane, to ``lifting force.'' This revision will

clarify that the static loads associated with the lifting tool, drive

rod and control rod weights are not included in the lifting force

limit.

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:

Will the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

No. The proposed change is administrative in nature and does not

represent any changes to the refueling process in the field. It more

accurately describes the components for which the LCO's [Limiting

Condition for Operation] protection is intended as well as giving a

more accurate description of the auxiliary hoist's minimum capacity.

It also broadens the domain of activities for which protective

measures are taken by including drag load testing into monitored

activities. At CNS [Catawba Nuclear Station], the auxiliary hoists

and the manipulator cranes are rated at [greater than or equal to]

3000 pounds and are surveillance tested to greater than 1000 pounds.

This brackets the limit force lifting value change from 600 to 1000

pounds in the amendment proposal.

Will the change create the possibility of a new or different

kind of accident from any accident previously evaluated?

No. This proposed administrative change reflects no changes in

the refueling processes, or any systems, structures or components

connected with the refueling process.

Will the change involve a significant reduction in a margin of

safety?

No. The proposed administrative change has no impact on

refueling processes, systems, structures or components, and does not

result in any significant reduction in a margin of safety. The

subject change only clarifies the original intent of the

specification and more accurately describes the involved components,

component capacities and the domain of activities for which measures

are taken to protect the reactor internals.

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: Atkins Library, University of

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

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

South Church Street, Charlotte, North Carolina 28242

NRC Project Director: Herbert N. Berkow

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

Nuclear Generating Station, Ocean County, New Jersey

Date of amendment request: August 23, 1996 (TSCR 245)

Description of amendment request: The amendment request proposes

new pressure-temperature (P-T) limits up to

[[Page 47978]]

22, 27, and 32 effective full power years (EFPY). The new sets of P-T

curves would be used beyond 17 EFPY in the future as the corresponding

EFPY of operation is completed.

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:

We have determined that this change request with respect to P-T

limits involves no significant hazards considerations in that

operation of the Oyster Creek Plant in accordance with the proposed

amendment, will not:

1. Involve a significant increase in the probability of an

accident because the new limits account for the increase in RT

NDT, including statistical uncertainty, due to neutron

irradiation of the reactor vessel as well as establishing initial RT

NDT on the basis of current Code requirements, also including

statistical uncertainty, in accordance with Reg. Guide 1.99, Rev. 2.

The new P-T curves will assure that brittle fracture of the reactor

vessel is prevented.

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

from any accident previously evaluated. These new limits are the

result of the calculation methodology in Reg. Guide 1.99, Rev. 2

[Radiation Embrittlement of Reactor Vessel Materials], as required

by Generic Letter 88-11 [NRC Position on Radiation Embrittlement of

Reactor Materials and its Impact on Plant Operations]. Primary

system configuration and function remain unchanged.

3. Involve a significant reduction in margin of safety because

the bases for the margin of safety remain the same as current

limits, i.e., ASME [American Society of Mechanical Engineers], Sect.

XI, App. G for available fracture toughness and applied stress

intensity, Reg. Guide 1.99, Rev. 2 for calculating applied stress

intensity, Reg. Guide 1.99, Rev. 2 for calculating adjusted RT

NDT and 10 CFR 50, App. G, for criticality conditions.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Ocean County Library,

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

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

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

NRC Project Director: John F. Stolz

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

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

Illinois

Date of amendment request: August 15, 1996

Description of amendment request: The proposed amendment would

modify the Clinton Power Station Technical Specifications to

incorporate the revised Safety Limit Minimum Critical Power Ratio

(SLMCPR) as calculated by General Electric (GE) for Cycle 7 operation.

The need to change the SLMCPR resulted from the 10 CFR Part 21

condition reported by GE in their letter to the NRC dated May 24, 1996.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration which is presented below:

(1) This change does not involve a significant increase in the

probability or consequences of any accident previously evaluated. In

lieu of utilizing a potentially nonconservative generic value, this

change revises the SLMCPR to be appropriately conservative as it has

been specifically calculated on a plant- and cycle-specific basis.

Although the SLMCPR does not apply (i.e., is not assumed or required

to be met) during any analyzed accident, the MCPR fuel cladding

Safety Limit ensures that during normal operation and during

anticipated operational occurrences (AOOs), at least 99.9% of the

fuel rods in the core do not experience transition boiling. The

revised value for the SLMCPR is determined using the same

methodology as the previous SLMCPR with the exception that it

utilizes plant specific conditions to determine the safety limit.

The revised SLMCPR, therefore, accounts for actual expected power

distributions in the Clinton Power Station (CPS) core as well as

CPS-specific uncertainties. This provides a more conservative SLMCPR

than the generic value used previously.

The proposed change does not affect any of the parameters or

conditions that contribute to initiation of any accidents previously

evaluated. In addition, the proposed change does not affect the

ability of any plant systems or equipment to operate as assumed in

the safety analyses. The revised SLMCPR will continue to ensure that

the fuel cladding integrity is not lost as a result of over-heating

during normal plant operation or any AOO. As a result, the proposed

change will not result in a significant increase in the consequences

of any accident previously evaluated.

(2) The proposed change does not involve any new modes or

operation, any changes to setpoints, or any plant modifications.

Further, the incorporation of a revised MCPR safety limit, which has

been determined to be acceptable for CPS Cycle 7 operation, does not

result in the creation of any new failure modes or potential

precursors to an accident. Therefore, the proposed change does not

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

any accident previously evaluated.

(3) The proposed SLMCPR has been evaluated to ensure that during

normal operation and during AOOs, at least 99.9% of the fuel rods in

the core do not experience transition boiling. As noted above, the

revised SLMCPR has been determined using the same methodology as

used previously with the exception of using CPS Cycle 7 specific

core and fuel design data. This change ensures that the margin of

safety for fuel cladding integrity is maintained by providing a CPS

specific MCPR safety limit as opposed to utilizing a potentially

less conservative generic limit. Therefore, the implementation of

the proposed change to the SLMCPR 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: Vespasian Warner Public

Library, 120 West Johnson Street, Clinton, Illinois 61727

Attorney for licensee: Leah Manning Stetzner, Vice President,

General Counsel, and Corporate Secretary, 500 South 27th Street,

Decatur, Illinois 62525

NRC Project Director: Gail H. Marcus

Southern California Edison Company, et al., Docket Nos. 50-361 and

50-362, San Onofre Nuclear Generating Station, Unit Nos. 2 and 3,

San Diego County, California

Date of amendment requests: May 8, 1996

Description of amendment requests: The licensee proposes to revise

improved Technical Specifications (TS) 3.9.4 and 3.9.5 to facilitate

testing of low pressure safety injection system components and permit

additional flexibility in scheduling maintenance on the shutdown

cooling system.

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.

Limiting Conditions for Operation (LCO) in Technical

Specifications (TSs) 3.9.4 and 3.9.5 define the operability

requirements for the Shutdown Cooling (SDC) system during refueling

operations (Mode 6) while the water level above the top of the

reactor vessel

[[Page 47979]]

flange is at least 23 feet and less than 23 feet, respectively. The

objective of these TSs is to ensure that 1) sufficient cooling is

available to remove decay heat, 2) the water in the reactor vessel

is maintained below 140 deg.F, and 3) sufficient coolant circulation

is maintained in the reactor core to minimize boron stratification

leading to a boron dilution incident.

The proposed TS changes affect the current limits imposed while

ensuring adherence to the bases of the TS. No plant modifications

are being made. The reactor cavity water level limitations and SDC

system required operating times are being changed based on plant

specific calculations and the objectives of the TSs are being

maintained.

1) Reduce the water level where two loops of SDC are required

from 23 feet to 20 feet above the reactor vessel flange,

Prior to the approval of Unit 2 Amendment No. 127 and Unit 3

Amendment No. 116, Technical Specification Bases Section 3/4.9.8 has

stated that ``With the reactor vessel head removed and 23 feet of

water above the reactor vessel flange, a large heat sink is

available for core cooling, thus in the event of a failure of the

operating shutdown cooling loop, adequate time is provided to

initiate emergency procedures to cool the core.''

In the Bases for the New Standard Technical Specifications,

``NUREG 1432, Revision 0, dated September 30, 1992, Section B 3.9.4

it is stated that; ``The 23 ft level was selected because it

corresponds to the 23 ft requirement established for fuel movement

in LCO 3.9.6, ``Refueling Water Level.''

Southern California Edison (Edison) calculations show that there

is an insignificant difference in the time to boil due to the 3-foot

change in required water level. Therefore, adequate water is still

available to mitigate the consequences of losing SDC.

2) Increase the time a required loop of the SDC system may be

removed from service from up to 1 hour per 8-hour period to up to 2

hours per 8-hour period, provided the upper guide structure has been

removed from the reactor vessel,

The proposed TS changes the time the SDC loop may be removed

from operation from up to 1 hour per 8-hour period to up to 2 hours

per 8-hour period, and allows removal of the SDC loop from operation

for testing of the Low Pressure Safety Injection (LPSI) system

components as well as for core alterations in the vicinity of the

hot legs. The proposed TS change also imposes certain restrictions

to ensure operating the SDC system in accordance with this proposed

TS change is of no safety significance. These [r]estrictions are

discussed separately below.

Specifically stating that the upper guide structure will be

removed assures that natural heat transfer is not impeded.

When securing the only operating loop of the SDC system the

maximum Reactor Coolant System (RCS) temperature is maintained [less

than or equal to] 140 deg.F. The initial conditions and heatup rate

are selected such that the RCS temperature remains [less than or

equal to] 140 deg.F during the test. Therefore, there is ample

margin to boiling. Typical initial temperatures are less than

100 deg.F.

The water being injected by the LPSI system test is cool water

from the Refueling Water Storage Tank (RWST) and will increase the

reactor cavity water level by several inches, providing more cool

water to the heat sink. The two hours is sufficient time to align

the system to test, perform the test, and restore the loop of SDC to

operation prior to exceeding 140 deg.F.

No operations are permitted that would cause a reduction of the

RCS boron concentration. This minimizes the probability of an

inadvertent boron dilution event. The use of adequately borated

water for injection into the RCS during the test provides assurance

that the test itself cannot lead to a boron dilution event. When the

SDC system is operating, the minimum SDC flow rate of 2200 gpm

imposed by Surveillance Requirements SR 3.9.4.1 and SR 3.9.5.1 is

sufficient to ensure complete mixing of the boron within the RCS.

Securing SDC flow is only allowed when the reactor cavity water

level is maintained greater than or equal to 20 feet above the

reactor vessel flange. This level ensures an adequate heat sink to

perform the LPSI pump suction header check valve test.

3) Allow for running 1 loop of shutdown cooling with additional

requirements when the water level is less than 20 feet but greater

than or equal to 12 feet above the reactor vessel flange,

4) Add an action to be taken when operating 1 loop of SDC with

less than 20 feet of water above the reactor vessel flange when the

specified requirements are not met,

In the event of a loss of SDC the time to boil is reduced from

approximately 4.0 hours when the water level is 23 feet above the

reactor vessel flange to approximately 2.3 hours at 12 feet,

assuming the reactor has only been shutdown for 6 days. However,

this is ample time to close containment (less than 1 hour) and to

restore SDC or initiate alternative cooling (e.g., add water to the

cavity (approximately 1 hour)). The reactor pressure vessel flange

is approximately 11' above the top of the fuel. Therefore, the water

level will be a minimum of 23' above the fuel, which still maintains

a large volume of water to provide a heat sink.

Requiring the reactor to be shutdown for at least 6 days to have

only one loop of SDC operable when the reactor cavity level is

between 20 feet and 12 feet above the reactor vessel flange ensures

that the time to boil is greater than twice the time it would take

to establish containment closure and to commence reactor cavity fill

with the required standby equipment.

One loop of SDC operating with a containment spray pump allows

for the high capacity LPSI pump to be the main standby pump capable

of filling the reactor cavity to at least 20 feet above the reactor

pressure vessel flange in the event SDC is lost. The high pressure

safety injection pump will also be maintained OPERABLE to increase

the water level if needed. In support of this contingency the RWST

will be required to contain the volume of water needed to raised

[raise] the level to 20 feet above the reactor pressure vessel

flange. As discussed above, the reactor cavity can be filled at a

rate of approximately 4.0 inches per minute with the LPSI pump.

If operating one loop of the SDC system with less than 20 feet

of water above the reactor vessel flange and any of the required

conditions are not met, requiring immediate action to establish

greater than or equal to 20 feet of water above the reactor vessel

flange ensures no time is wasted trying to restore the required

condition not met. By taking action to restore the level to 20 feet

above the reactor vessel flange the plant will be placed in TS

3.9.4, which only requires one loop of SDC to be operable.

Additionally, the core will not heat up while the water level in the

reactor cavity is being raised with cool water from the RWST. This

will provide additional time to either restore the one loop of SDC

or take other actions to provide core cooling as required by TS

3.9.4.

A Probabilistic Risk Assessment (PRA), with a) one loop of the

SDC system operable with the reactor cavity water level greater than

or equal to 12 feet above the reactor vessel flange, and b) one loop

of the SDC system operable with the reactor cavity water level

greater than or equal to 20 feet above the reactor vessel flange,

showed that the operations in accordance with the proposed TS would

not significantly increase the probabilities of inventory boiling

and core damage.

5) Item 6 adds wording to the notes in LCOs 3.9.4 and 3.9.5 that

was unintentionally deleted by the Unit 2 Amendment No. 127 and Unit

3 Amendment No. 116.

This is an editorial change.

Therefore, proposed changes 1 through 5 do not involve a

significant increase in the probability or consequences of an

accident.

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

different kind of accident from any accident previously evaluated.

1) Reduce the water level where two loops of SDC are required

from 23 feet to 20 feet above the reactor vessel flange,

2) Increase the time a required loop of the SDC system may be

removed from service from up to 1 hour per 8-hour period to up to 2

hours per 8-hour period, provided the upper guide structure has been

removed from the reactor vessel,

3) Allow for running 1 loop of shutdown cooling with additional

requirements when the water level is less than 20 feet but greater

than or equal to 12 feet above the reactor vessel flange,

4) Add an action to be taken when operating 1 loop of SDC with less

than 20 feet of water above the reactor vessel flange when the

specified requirements are not met,

The Limiting Conditions for Operation (LCO) in Technical

Specifications (TSs) 3.9.4 and 3.9.5 define the operability

requirements for the SDC system during refueling operations (Mode 6)

while the water level above the top of the reactor vessel flange is

at least 23 feet and less than 23 feet, respectively. The objective

of the proposed TS changes is to ensure that the intent of the Bases

is maintained. [i.e., 1) sufficient cooling is available to remove

decay heat, 2) water in the reactor vessel is maintained below

140 deg.F, and 3) sufficient coolant

[[Page 47980]]

circulation is maintained in the reactor core to minimize boron

stratification leading to a boron dilution incident.]

The proposed TS changes affect the current limits imposed while

ensuring adherence to the bases of the TS. No plant modifications

are being made. The reactor cavity water level limitations and SDC

system required operating times are being changed based on plant

specific calculations, and the objective of the TSs are being

maintained. The added requirements and action statement facilitate

safe operation.

5) Item 6 adds wording to the notes in LCOs 3.9.4 and 3.9.5 that

was unintentionally deleted by the Unit 2 Amendment No. 127 and Unit 3

Amendment No. 116.

This is an editorial change.

Therefore, the operation of the facility in accordance with

proposed changes 1 through 5 does not create the possibility of a

new or different kind of accident from any accident previously

evaluated.

3. The proposed change will not involve a significant reduction

in a margin of safety.

Limiting Conditions for Operation (LCO) in TSs 3.9.4 and 3.9.5

define the operability requirements for the SDC system during

refueling operations (Mode 6) while the water level above the top of

the reactor vessel flange is at least 23 feet and less than 23 feet,

respectively. The objectives of these TSs are to ensure that 1)

sufficient cooling is available to remove decay heat, 2) the water

in the reactor vessel is maintained below 140 deg.F, and 3)

sufficient coolant circulation is maintained in the reactor core to

minimize boron stratification leading to a boron dilution incident.

1) Reduce the water level where two loops of SDC are required

from 23 feet to 20 feet above the reactor vessel flange,

Prior to the approval of Unit 2 Amendment No. 127 and Unit 3

Amendment No. 116, Technical Specification Bases Section 3/4.9.8 has

stated that ``With the reactor vessel head removed and 23 feet of

water above the reactor vessel flange, a large heat sink is

available for core cooling, thus in the event of a failure of the

operating shutdown cooling loop, adequate time is provided to

initiate emergency procedures to cool the core.''

In the Bases for the New Standard Technical Specifications,

NUREG 1432, Revision 0, dated September 30, 1992, Section B 3.9.4 it

is stated that ``The 23 ft level was selected because it corresponds

to the 23 ft requirement established for fuel movement in LCO 3.9.6,

``Refueling Water Level.''

Edison calculations show that there is a minimal difference in

the time to boil due to the 3-foot change in required water level.

Therefore, the margin of safety has not been significantly reduced.

2) Increase the time a required loop of the SDC system may be

removed from service from up to 1 hour per 8-hour period to up to 2

hours per 8-hour period, provided the upper guide structure has been

removed from the reactor vessel,

The proposed TS changes the time the SDC loop may be removed

from operation from up to 1 hour per 8-hour period to up to 2 hours

per 8-hour period, and allows removal of the SDC loop from operation

for testing of the LPSI system components as well as for core

alterations in the vicinity of the hot legs. The proposed TS change

also imposes certain restrictions to ensure operating the SDC system

in accordance with this proposed TS change is of no safety

significance. These restrictions are discussed separately below.

Specifically stating that the upper guide structure will be

removed assures that natural heat transfer is not impeded.

When securing the only operating loop of the SDC system, the

maximum RCS temperature is maintained [less than or equal to]

140 deg.F. The initial conditions and heatup rate are selected such

that RCS temperature remains [less than or equal to] 140 deg.F

during the test. Therefore, there is ample margin to boiling.

Typical initial temperatures are less than 100 deg.F.

The water being injected by the LPSI system test is cool borated

water from the RWST and will increase the level of the reactor

cavity by several inches. The two hours is sufficient time to align

the system to test, perform the test, and restore the loop of SDC to

operation prior to exceeding 140 deg.F.

No operations are permitted that would cause a reduction of the

RCS boron concentration. This minimizes the probability of an

inadvertent boron dilution event. The use of adequately borated

water for injection into the RCS during the test provides assurance

that the test itself cannot lead to a boron dilution event. When the

SDC system is operating, the minimum SDC flow rate of 2200 gpm is

sufficient to ensure complete mixing of the boron within the RCS.

Securing SDC flow is only allowed when the reactor cavity water

level is maintained greater than or equal to 20 feet above the

reactor vessel flange. This level ensures an adequate heat sink to

perform the LPSI pump suction header check valve test.

The added requirements and the nature of the test provide

assurances that the water temperature will be maintained less than

140 deg.F and that boron stratification is prevented.

3) Allow for running 1 loop of shutdown cooling with additional

requirements when the water level is less than 20 feet but greater

than or equal to 12 feet above the reactor vessel flange,

4) Add an action to be taken when operating 1 loop of SDC with less

than 20 feet of water above the reactor vessel flange when the

specified requirements are not met,

In the event of a loss of SDC, the time to boil is reduced from

approximately 4.0 hours when the water level is 23 feet above the

reactor vessel flange to approximately 2.3 hours at 12 feet, when

the reactor has only been shutdown for 6 days. However, this is

ample time to close containment (less than 1 hour), and to restore

SDC or initiate alternative cooling (e.g., add water to the cavity

(approximately 1 hour)).

Requiring the reactor to be shutdown for at least 6 days to have

only one loop of SDC operable when the reactor cavity level is

between 20 feet and 12 feet above the reactor vessel flange ensures

that the time to boil is greater than twice the time it would take

us to establish containment closure and to commence reactor cavity

fill with the required standby equipment.