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

UNITED STATES NUCLEAR REGULATORY COMMISSION

Biweekly Notice

Applications and Amendments to Facility Operating Licenses

Involving No Significant Hazards Considerations

I. Background

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

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

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

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

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

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

the Commission the authority to issue and make immediately effective

any amendment to an operating license upon a determination by the

Commission that such amendment involves no significant hazards

consideration, notwithstanding the pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all notices of amendments issued, or

proposed to be issued from June 8, 1996, through June 21, 1996. The

last biweekly notice was published on June 19, 1996 (61 FR 31171).

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 August 2, 1996, the licensee may file a request for a hearing

with respect to issuance of the amendment to the subject facility

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

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

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

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

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

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

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

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

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

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

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

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

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

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

Secretary or the designated Atomic Safety and Licensing Board will

issue a notice of a hearing or an appropriate order.

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

shall set forth with particularity the interest of the petitioner in

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

the proceeding. The petition should specifically explain the reasons

why intervention should be permitted with particular reference to the

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

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

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

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

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

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

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

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

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

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

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

the specificity requirements described above.

[[Page 34885]]

Not later than 15 days prior to the first prehearing conference

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

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

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

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

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

brief explanation of the bases of the contention and a concise

statement of the alleged facts or expert opinion which support the

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

contention at the hearing. The petitioner must also provide references

to those specific sources and documents of which the petitioner is

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

facts or expert opinion. Petitioner must provide sufficient information

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

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

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

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

petitioner who fails to file such a supplement which satisfies these

requirements with respect to at least one contention will not be

permitted to participate as a party.

Those permitted to intervene become parties to the proceeding,

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

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

hearing, including the opportunity to present evidence and cross-

examine witnesses.

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

determination on the issue of no significant hazards consideration. The

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

If the final determination is that the amendment request involves

no significant hazards consideration, the Commission may issue the

amendment and make it immediately effective, notwithstanding the

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

of the amendment.

If the final determination is that the amendment request involves a

significant hazards consideration, any hearing held would take place

before the issuance of any amendment.

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

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

Commission, Washington, DC 20555-0001, 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.

Boston Edison Company, Docket No. 50-293, Pilgrim Nuclear Power

Station, Plymouth County, Massachusetts

Date of amendment request: May 1, 1996

Description of amendment request: The proposed amendment would

modify Table 3.1.1, ``Reactor Protection System (SCRAM) Instrumentation

Requirement,'' Table 3.2.C.1, ``Instrumentation that Initiates Rod

Blocks,'' and Technical Specification 3/4.4, ``Standby Liquid

Control.''

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

Note 7 to Table 3.1.1 and Note 6 to Table 3.2.C.1

The changes to Note 7 to Table 3.1.1 and the addition of Note 6

to Table 3.2.C.1 are proposed to clarify their requirements, the

appropriate action to take, and their relationship to plant modes.

This revised scram and rod block applicability is acceptable because

control rods withdrawn from a core cell containing no fuel

assemblies have a negligible impact on the reactivity of the core,

and, therefore, these features are not required to be operable (i.e.

provide the capability to scram). Provided all rods otherwise remain

inserted, the RPS [Reactor Protection System] functions serve no

purpose and are not required. In this condition, the required

shutdown margin (Specification 3.3.A.1) and the required one-rod-out

interlock (Specification 3.10.A) ensure that no event requiring the

RPS or Rod Block will occur.

The Actions of Table 3.1.1 for inoperable equipment were

previously revised in Amendment 147 to be consistent with

the improved STS [Standard Technical Specifications]. Action (A)

requires fully inserting all insertable control rods in core cells

containing one or more fuel assemblies. Since Specification 3.10.A

requires all control rods to be fully inserted during fuel movement,

the proposed applicable conditions cannot be entered while moving

fuel. In addition, Specification 3.10.D used for controlling

multiple control rod removal, requires all control rods in a 3X3

array centered on the CRDs [Control Rod Drive] being removed to be

fully inserted and electrically disarmed and all other control rods

fully inserted. The only possible action is control rod withdrawal,

which is addressed by Action A.

Hence operating Pilgrim in accordance with the proposed changes

will not involve a significant increase in the probability or

consequences of an accident previously evaluated.

Section 3/4.4

The proposed change involves reformatting, renumbering, and

rewording of the existing Technical Specifications and Bases along

with other changes to the Technical Specifications discussed above.

The reformatting, renumbering, and rewording along with the other

changes listed involves no technical changes to existing Technical

Specifications, and does not impact initiators of analyzed events.

It also does not impact the assumed mitigation of accidents or

transient events. Therefore, the change does not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

The proposed change relocates requirements to other sections of

the Technical Specifications, to plant procedures, or to the

Technical Specifications BASES. The procedure change and BASES

change processes require any changes that reflect plant design as

described in the FSAR [Final Safety Analysis Report] be evaluated in

accordance with 10 CFR 50.59. Since any changes will be evaluated

per 10 CFR 50.59, no increase (significant or insignificant) in the

probability or consequences of an accident previously evaluated will

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

increase in the probability or consequences of an accident

previously evaluated.

[[Page 34886]]

The proposed change provides more stringent requirements than

previously existed in the Technical Specifications. The more

stringent requirements will not result in operation that will

increase the probability of initiating an analyzed event. If

anything the new requirements may decrease the probability or

consequences of an analyzed event by incorporating the more

restrictive changes discussed above. The change will not alter

assumptions relative to mitigation of an accident or transient

event. The more restrictive requirements will not alter the

operation of process variables, structures, systems, or components

as described in the safety analyses. Therefore, the change will not

involve a significant increase in the probability or consequences of

an accident previously evaluated.

The proposed change deletes the requirements for Standby Liquid

Control (SLC) System operability during Hot Shutdown, Cold Shutdown,

and Refueling. The SLC System is not assumed in the initiation of

any previously evaluated events and therefore the proposed change

will not increase the probability or consequence of a previously

analyzed accident. The SLC System is not assumed to operate in the

mitigation of any previously analyzed accidents which are assumed to

occur during Hot Shutdown, Cold Shutdown or Refueling. This change

will not result in operation that will increase the probability of

initiating an analyzed event. This change will not alter assumptions

relative to mitigation of an accident or alter the operation of

process variables, structures, systems, or components as described

in the safety analyses. Therefore, this change will not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

The proposed change adds an action for both SLC subsystems

inoperable that delays the requirement to initiate plant shutdown

immediately and allows time to recover at least one subsystem before

subjecting the plant to a potentially unnecessary transient.

Allowing a short period of time to recover one subsystem is

acceptable because of the large number of independent control rods

available to shut down the reactor and the diversity of means

available to cause control rod insertion. This change will not alter

assumptions relative to mitigation of an accident or alter the

operation of process variables, structures, systems, or components

as described in the safety analyses. Therefore, this change will not

involve a significant increase in the probability or consequences of

an accident previously evaluated.

The proposed change deletes requirements for demonstrating

operability of the redundant subsystems which eliminates excessive

and unnecessary testing of safety significant equipment. This is

consistent with guidance 10.1 of Generic Letter 93-05, ``Line-Item

Technical Specifications Improvements to Reduce Surveillance

Requirement for Testing During Power Operations''. The change does

not affect the ability of the SLC system to perform on demand, and

by actually lowering the number of demands to demonstrate

operability, reduces the probability of equipment failure. Since the

change will not alter assumptions relative to mitigation of an

accident or alter the operation of process variables, structures,

systems, or components as described in the safety analyses, the

change will not involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed change replaces the requirement to verify B-10

enrichment concentration by test anytime boron is added to the

solution and each refueling outage with verifying the enrichment

prior to addition. Since enrichment of the solution in the tank

cannot change by any other means but chemical addition, ensuring

that only properly enriched material is available for addition is

adequate to maintain enrichment at the required level. This change

will not alter assumptions relative to mitigation of an accident or

alter the operation of process variables, structures, systems, or

components as described in the safety analyses. Therefore, this

change will not involve a significant increase in the probability or

consequences of an accident previously evaluated.

2. The operation of Pilgrim Station in accordance with the

proposed amendment will not create the possibility of a new or

different kind of accident from any accident previously evaluated.

Note 7 to Table 3.1.1 and Note 6 to Table 3.2.C.1

The changes to Note 7 to Table 3.1.1, and the addition of Note 6

to Table 3.2.C.1 are proposed to clarify their requirements, the

appropriate action to take, and their relationship to plant modes.

This revised scram and rod block applicability is acceptable because

control rods withdrawn from a core cell containing no fuel

assemblies have a negligible impact on the reactivity of the core,

and, therefore, are not required to be operable. Provided all rods

otherwise remain inserted, the RPS functions serve no purpose and

are not required. In this condition, the required shutdown margin

(Specification 3.3.A.1) and the required one-rod-out interlock

(Specification 3.10.A) ensure that no event requiring the RPS or Rod

Block will occur.

The Actions of Table 3.1.1 for inoperable equipment were

previously revised in Amendment 147 to be consistent with

the improved STS. Action (A) requires fully inserting all insertable

control rods in core cells containing one or more fuel assemblies.

Since Specification 3.1O.A requires all control rods to be fully

inserted during fuel movement, the proposed applicable conditions

cannot be entered while moving fuel. In addition, Specification

3.10.D, used for controlling multiple control rod removal, requires

all control rods in a 3X3 array centered on the CRDs being removed

to be fully inserted and electrically disarmed and all other control

rods fully inserted. The only possible action is control rod

withdrawal, which is addressed by Action A. Hence, operating Pilgrim

in accordance with the proposed changes will not create the

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

previously evaluated.

Section 3/4.4

The proposed change involves reformatting, renumbering, and

rewording of the existing Technical Specifications and Bases along

with other changes to the Technical Specifications discussed above.

The reformatting, renumbering, and rewording along with the other

changes listed involves no technical changes to existing Technical

Specifications. These changes are administrative and do not impact

the assumed mitigation of accidents or transient events. Therefore,

these changes do not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The proposed change relocates requirements to other Technical

Specification sections, to plant procedures, or to the Technical

Specification BASES. Relocating requirements will not alter the

plant configuration (no new or different type of equipment will be

installed) or changes in methods governing normal plant operation.

Relocating requirements will not impose different requirements and

adequate control of information will be maintained. Relocating

requirements will not alter assumptions made in the safety analysis

and licensing basis. Therefore, these changes will not create the

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

previously evaluated.

The proposed changes make some existing requirements more

restrictive and add additional requirements to the Technical

Specifications but will not alter the plant configuration (no new or

different type of equipment will be installed) or change methods

governing normal plant operation. These changes do impose different

requirements, however, they are consistent with assumptions made in

the safety analyses. Therefore, these changes will not create the

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

previously evaluated.

The proposed change relaxes the modes of applicability for the

SLC. Relaxing the applicability will not involve a physical

alteration of the plant (no new or different type of equipment will

be installed) or changes in methods governing normal plant

operation. Therefore, this change will not create the possibility of

a new or different kind of accident from any accident previously

evaluated.

3. The operation of Pilgrim Station in accordance with the

proposed amendment will not involve a significant reduction in a

margin of safety.

Note 7 to Table 3.1.1 and Note 6 to Table 3.2.C.1

This revised scram and rod block applicability is acceptable

because control rods withdrawn from a core cell containing no fuel

assemblies have a negligible impact on the reactivity of the core,

and, therefore, are not required to be operable (provide a scram).

Provided all rods otherwise remain inserted, the RPS functions serve

no purpose and are not required. In this condition, the required

shutdown margin (Specification 3.3.A.1) and the required one-rod-out

interlock (Specification 3.10.A) ensure that no event requiring the

RPS or Rod Block will occur.

The Actions of Table 3.1.1 for inoperable equipment were

previously revised in

[[Page 34887]]

Amendment 147 to be consistent with the improved STS.

Action (A) requires fully inserting all insertable control rods in

core cells containing one or more fuel assemblies. Since

Specification 3.10.A requires all control rods to be fully inserted

during fuel movement, the proposed applicable conditions cannot be

entered while moving fuel. In addition, Specification 3.10.D, used

for controlling multiple control rod removal, requires all control

rods in a 3X3 array centered on the CRDs being removed to be fully

inserted and electrically disarmed and all other control rods fully

inserted. The only possible action is control rod withdrawal, which

is adequately addressed by Action A.

Therefore, operating Pilgrim in accordance with the proposed

changes will not involve a significant reduction in a margin of

safety.

Section 3/4.4

The administrative changes involve no technical changes. These

proposed changes will not reduce a margin of safety because there is

no impact on any safety analysis assumptions. Also, because the

change is administrative in nature, no question of safety is

involved. Therefore, these changes do not involve a significant

reduction in a margin of safety. The change relocates requirements

to other Technical Specification sections, to plant procedures, or

to the Technical Specification BASES. These changes will not reduce

a margin of safety since there is no impact on any safety analysis

assumptions. In addition, the requirements to be transposed are the

same as the existing Technical Specifications. Since any changes to

plant procedures and Technical Specification BASES are required to

be evaluated per 10 CFR 50.59, no reduction (significant or

insignificant) in a margin of safety will be allowed. Therefore,

these changes will not involve a significant reduction in a margin

of safety.

The addition of new requirements and making existing ones more

restrictive either increases or does not affect the margin of

safety. These changes do not impact any safety analysis assumptions.

As such, no question of safety is involved. Therefore, these changes

will not involve a significant reduction in a margin of safety.

The proposed change would remove a backup (in the Hot Shutdown,

Cold Shutdown, and Refueling Modes) to the available systems for

reactivity control; however, this backup is not considered in the

margin of safety when determining the required reactivity for

shutdown and refueling events. This change will have no impact on

any safety analysis assumptions. As such, no question of safety is

involved. Therefore, this change does not involve a significant

reduction in a margin of safety.

The SLC system is not assumed to function in any DBA or

transient and is not the primary success path of a safety sequence

analysis. It is a backup to the CRD scram function, therefore,

allowing a short period of time to recover one subsystem will have

no impact on any safety analysis assumptions. As such, no question

of safety is involved. Therefore, this change does not involve a

significant reduction in a margin of safety.

The change does not alter the requirements for enrichment/

concentration of the boron solution necessary to satisfy 10 CFR

50.62. Since enrichment of the solution in the tank cannot change by

any other means but chemical addition, this change does not involve

a significant reduction in a margin of safety.

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

this review, it appears that the three standards of 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: Plymouth Public Library, 11

North Street, Plymouth, Massachusetts 02360

Attorney for licensee: W. S. Stowe, Esquire, Boston Edison Company,

800 Boylston Street, 36th Floor, Boston, Massachusetts 02199

NRC Project Director: Jocelyn A. Mitchell, Acting Director

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 15, 1995

Description of amendments request: The proposed amendments would

revise the Technical Specifications (TS) to alter the wording of TS

4.8.2.5.a in accordance with the guidance of Generic Letter (GL) 91-09,

``Modification of Surveillance Interval For The Electrical Protection

Assemblies In Power Supplies For The Reactor Protection 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 amendments do not involve a significant increase

in the probability or consequences of an accident previously

evaluated because the proposed change does not alter the design,

function, or operation of the EPAs [Electrical Protective

Assemblies]. The proposed amendments modify the surveillance

requirement for an electrical protective device on the Reactor

Protection System [RPS]. The RPS-EPA units are designed to protect

RPS equipment from abnormal operating voltage or frequency. The

proposed change will preclude the need to test the RPS-EPA units

during power operation. This will eliminate the potential for

reactor scrams and Group isolations during performance of the

surveillance, thus, preventing unwarranted challenges to safety

systems. The proposed change does not affect any accident precursor

or initiator. Therefore, the probability of an accident is not

affected by the proposed change. The proposed amendments do not

affect the operability of the RPS-EPA units. The proposed change

does not affect the ability of the Reactor Protection System to

maintain the integrity of the fuel cladding, protect the reactor

coolant pressure boundary, or limit the amount of energy released to

primary containment. Therefore, the consequences of an accident is

not affected by the proposed change.

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

new or different kind of accident from any accident previously

evaluated. As stated above, these proposed amendments do not alter

the design, functions, or operation of the EPAs. The RPS relay trip

logic remains protected from power supplies operating with abnormal

voltage or frequency. Additionally, the redundancy of this

protection is not changed.

Thus, the proposed amendments do not create the possibility of a

new or different kind of accident.

3. The proposed amendments do not involve a significant

reduction in a margin of safety because the benefit to safety by

reducing the frequency of testing during power operation and

attendant possible challenges to safety systems more than offsets

any risk to safety from relaxing the surveillance requirement to

test the EPAs during power operation. The testing of each EPA

channel involves a dead-bus transfer and the momentary interruption

of power results in a half scram and half isolation. Generic Letter

91-09 notes that many plants have encountered problems with the

reset of the half trip resulting in inadvertent scrams and group

isolations that challenge safety systems during power operation.

Eliminating EPA testing at power operation increases the margin of

safety by eliminating the potential for trips due to testing that

challenge safety systems. An insignificant reduction in the margin

of safety is introduced by increasing the test interval up to a

maximum of a refuel cycle which will produce a small increase in

risk that an inoperable EPA would not be detected. The elimination

of potential challenges to safety systems provides a safety benefit

that offsets the increased risks of component failure.

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: William D. Johnson, Vice President and

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

Raleigh, North Carolina 27602

NRC Project Director: Eugene V. Imbro

[[Page 34888]]

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: June 6, 1996

Description of amendment request: The proposed change would revise

technical specifications (TS) Section 4.2.3 to allow the licensee to

defer the ultrasonic inspection of the reactor coolant pump flywheel

for one operating cycle.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The safety function of the Reactor Coolant Pump (RCP) flywheel

is to provide a coastdown period during which the RCPs would

continue to provide reactor coolant flow to the core after a loss of

power to the RCPs. The maximum loading on the RCP motor flywheel

results from overspeed following a large break Loss of Coolant

Accident (LOCA). The estimated maximum obtainable speed in the event

of a Reactor Coolant System (RCS) piping break was established

conservatively, and the proposed one-time change does not affect

that analysis.

The RCP flywheels have been carefully designed and manufactured

from high quality steel. Twenty-two inspections have been performed

at HBRSEP, Unit No. 2 over the past 25 years and no indications have

been discovered that would affect the integrity of the flywheel. The

Westinghouse Owners Group (WOG) has performed an extensive study

documented in WCAP-14535, ``Topical Report on Reactor Coolant Pump

Flywheel Inspection Elimination,'' that includes an evaluation of

industry experience, a stress and fracture evaluation, and a risk

assessment, and has concluded that RCP flywheel inspections may be

safely eliminated.

Reduced coastdown times due to a single failed flywheel would

not place the plant in an unanalyzed condition since a locked rotor

(i.e., an instantaneous coastdown) is analyzed in the Updated Final

Safety Analysis Report (UFSAR). The proposed change also does not

increase the amount of radioactive material available for release or

modify any systems used for mitigation of releases during an

accident. Therefore, the proposed change does not involve an

increase in the probability of consequences of an accident

previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

The proposed change will not change the design, configuration,

or method of operation of the plant. Therefore, the proposed change

will not create the possibility of a new kind of accident from any

previously evaluated.

3. The proposed change does not involve a significant reduction

in the margin of safety.

The RCP flywheels have been carefully designed and manufactured

from high quality steel. Twenty-two inspections have been performed

at HBRSEP, Unit No. 2 over the past 25 years and no indications have

been discovered that would affect the integrity of the flywheel. The

Westinghouse Owners Group (WOG) has performed an extensive study

documented in WCAP-14535, ``Topical Report on Reactor Coolant Pump

Flywheel Inspection Elimination,'' that includes an evaluation of

industry experience, a stress and fracture evaluation, and a risk

assessment, and has concluded that RCP flywheel inspections may be

safety eliminated. The proposed change would only result in a one-

time deferral of the scheduled inspection for one operating cycle.

In consideration of the historical integrity of the HBRSEP, Unit No.

2 RCP flywheels, the industry experience, the results of the WOG

study, and the deferral of the risk of RCP flywheel damage during

disassembly and inspection, we conclude that a one operating cycle

deferral of the scheduled RCP flywheel inspection will not result in

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: Hartsville Memorial Library,

147 West College Avenue, Hartsville, South Carolina 29550

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

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

Raleigh, North Carolina 27602

NRC Project Director: Eugene V. Imbro

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

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

North Carolina

Date of amendment request: May 31, 1996

Description of amendment request: The proposed amendment would

change the plant Technical Specifications (TS) Table 3.3-7, Seismic

Monitoring Instrumentation, and TS Table 4.3-4, Seismic Monitoring

Instrumentation Surveillance Requirements, to correct the location

described for one of the three Triaxial Peak Accelerograph Recorders.

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 or consequences of an accident

previously evaluated.

These recorders are passive components which serve only a

recording function. They can neither initiate an accident nor serve

to mitigate accident consequences. The proposed change serves only

to correct the location, commensurate with design documents, for one

of the three recorders described in the Technical Specifications.

Accordingly, this change is administrative in nature. Therefore,

there would be no increase in the probability or consequences of an

accident previously evaluated.

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

new or different kind of accident from any accident previously

evaluated.

The proposed correction is an administrative change to correct

the location of a recorder currently described in the Technical

Specifications. No physical alterations to plant equipment are being

made, and there will be no changes that alter how any safety-related

system performs its function. 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 amendment does not involve a significant

reduction in the margin of safety.

Technical Specification Bases 3/4.3.3.3 specify the acceptance

level for seismic instrumentation as ``consistency'' with the

recommendations of Regulatory Guide 1.12. Since the regulatory guide

states only that one recorder should be provided at a ``selected

location on the reactor piping,'' it is not material whether it is

installed on Loop 1 versus Loop 2. Therefore, the proposed change

does not affect a margin of safety as defined in the Bases to the

Technical Specifications.

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: Cameron Village Regional

Library, 1930 Clark Avenue, Raleigh, North Carolina 27605

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

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

Raleigh, North Carolina 27602

NRC Project Director: Eugene V. Imbro

[[Page 34889]]

Florida Power and Light Company, Docket Nos. 50-250 and 50-251,

Turkey Point Plant Units 3 and 4, Dade County, Florida

Dates of amendment request: December 18, 1995, May 3 and June 11,

1996

Description of amendment request: The licensee proposed to change

the Turkey Point Units 3 and 4 Technical Specifications (TS) to uprate

the core thermal output of Turkey Point Units 3 and 4 from 2200 MWt to

2300 MWt. The proposed TS changes were divided into eight groups. The

submittal included a ``No Significant Hazards'' evaluation for each of

the eight groups. The groupings are as follows:

TS changes associated with the uprated power level, the revised

core safety limits, revised DNB [departure from nucleate boiling]

parameters, Engineered Safety Features Actuation System (ESFAS) and

reactor trip setpoint changes, and Reactor Coolant Pump (RCP) Breaker

Position Trip, were evaluated together. The safety of these proposed

changes were verified by the accident analyses that were completed in

support of the uprated power.

TS changes associated with reducing the SI [safety injection] pump

discharge head requirement and increasing usable volume requirements

for the Demineralized Water Storage Tank (DWST) and the Condensate

Storage Tank (CST) were addressed together.

TS changes associated with pressurizer and main steam safety valve

(MSSV) setpoint tolerance increases were assessed together.

TS changes associated with operation at reduced power with

inoperable MSSVs were assessed separately.

TS changes associated with the service period for heatup and

cooldown pressure-temperature limit curves were assessed together.

The Surveillance Requirement change for the emergency containment

cooling [ECC] unit operability was handled separately since this was a

design change that required extensive evaluations.

TS change associated with the methyl iodide removal efficiency in

the Control Room Emergency Ventilation System was assessed separately.

All LOCA [loss-of-coolant accident] related changes dealing with

the peaking factor increase, COLR [core operating limit report]

changes, Evaluation Model references, and relocation of peaking factors

from the TS and subsequent inclusion in the COLR were included in one

``No Significant Hazards'' evaluation. All of the items are closely

related since the LOCA analysis is performed to ensure peaking factor

acceptability.

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.

LICENSE CONDITION, RATED THERMAL POWER, CORE SAFETY LIMITS,

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS, ESFAS

INSTRUMENTATION TRIP SETPOINTS, DNB PARAMETERS AND RCP BREAKER

POSITION TRIP

(1) Operation of the facility in accordance with the proposed

amendments would not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The proposed changes do not involve an increase in the

probability or consequences of an accident previously evaluated

because operation with these revised values will not cause any

design or analysis acceptance criteria to be exceeded. The

structural and functional integrity of all plant systems are

unaffected. The overtemperature Delta T and overpower Delta T

reactor trip functions as well as ESFAS functions are part of the

accident mitigation response and are not accident initiators. All

proposed changes have been assessed and no design and analysis

acceptance criteria have been exceeded. Therefore the probability of

occurrence previously evaluated is not affected.

The proposed changes do not affect the integrity of the fission

product barriers utilized for mitigation of dose consequences as a

result of an accident. Dose consequences were reviewed and

reanalyzed (as needed) and found acceptable. Therefore, the

probability or consequences of an accident previously evaluated are

not significantly increased.

(2) Operation of the facility in accordance with the proposed

amendments would not create the possibility of a new or different

kind of accident from any accident previously evaluated.

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

different kind of accident from any accident previously evaluated

because their effects do not affect accident initiation sequences.

All new operating configurations have been evaluated and no new

limiting single failures have been identified. In addition, no new

failure modes have been identified. Therefore, it is concluded that

no new or different kind of accident from any accident previously

evaluated has been created as a result of these revisions.

(3) Operation of the facility in accordance with the proposed

amendments would not involve a significant reduction in a margin of

safety.

The proposed changes do not involve a reduction in a margin of

safety because the margin of safety associated with these parameters

as verified by the results of the accident analyses, are within

acceptable limits. All transients impacted have been analyzed and

have met the applicable accident analyses acceptance criteria (e.g.,

DNBR [departure from nucleate boiling ratio], RCS [reactor coolant

system] pressure, secondary side pressure, etc.). The margin of

safety required for each affected safety analysis is maintained. The

adequacy of the revised Technical Specifications values has been

confirmed such that there is no reduction in the margin of safety.

Therefore, the proposed changes do not involve a significant

reduction in the margin of safety.

AVAILABLE VOLUME CHANGE FOR CONDENSATE STORAGE TANK (CST) AND

DEMINERALIZED WATER STORAGE TANK (DWST), AND REDUCED SAFETY

INJECTION (SI) PUMP DISCHARGE HEAD REQUIREMENT.

(1) Operation of the facility in accordance with the proposed

amendments would not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The revised tank volumes and SI head requirements have been

evaluated with respect to system performance and analysis impacts.

All accident analysis acceptance criteria continue to be met. The

design function of all affected systems have been reviewed and all

system design criteria continue to be met. The structural and

functional integrity of the affected systems are unaffected. These

changes are not initiators for any accident and therefore the

probability of occurrence of an accident previously evaluated has

not increased.

The proposed changes do not affect the integrity of the fission

product barriers for mitigation of dose consequences. All dose

consequences remain well within the 10 CFR 100 limits. Therefore

there is no increase in the probability or consequences of an

accident previously evaluated.

(2) Operation of the facility in accordance with the proposed

amendments would not create the possibility of a new or different

kind of accident from any accident previously evaluated.

The revised tank volumes and SI head requirements do not create

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

accident previously evaluated because these modifications do not

affect accident initiation sequences. No new operating configuration

is being imposed by the adjustments that would create a new failure

scenario. In addition, no new failure modes or limiting single

failures have been identified. Therefore, it is concluded that no

new or different kind of accident from any accident previously

evaluated have been created as a result of these revisions.

(3) Operation of the facility in accordance with the proposed

amendments would not involve a significant reduction in a margin of

safety.

The proposed changes do not involve a reduction in a margin of

safety because the margin of safety associated with these

parameters, as verified by the results of the accident analyses and

system evaluations, are within acceptance limits. The margin of

safety required for each affected safety analysis is maintained.

Therefore, the proposed changes do not involve a significant

reduction in the margin of safety.

PRESSURIZER AND MAIN STEAM SAFETY VALVE SETPOINT TOLERANCES

[[Page 34890]]

(1) Operation of the facility in accordance with the proposed

amendments would not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The revised tolerances for main steam safety valves and

pressurizer safety valves do not involve an increase in the

probability or consequences of an accident previously evaluated

because operation with these revised values will not cause any

design or analytical acceptance criteria, such as those applicable

to primary and secondary side pressures to be exceeded. The

structural and functional integrity of the valves are unaffected by

this proposed change. The tolerance changes do not initiate or cause

initiation of any transient. Therefore, the probability of

occurrence previously evaluated is not affected.

The changes do not affect the integrity of the fission product

barriers utilized for dose consequence mitigation. Therefore, the

probability or consequences of an accident previously evaluated is

not increased.

(2) Operation of the facility in accordance with the proposed

amendments would not create the possibility of a new or different

kind of accident from any accident previously evaluated.

The revised valve tolerances do not create the possibility of a

new or different kind of accident from any accident previously

evaluated because the tolerances do not affect accident initiation

sequences. No new operating configuration is being imposed by the

tolerances that would create a new failure scenario. In addition, no

new failure modes or limiting single failures have been identified.

Therefore, it is concluded that no new or different kind of accident

from any accident previously evaluated have been created as a result

of these revisions.

(3) Operation of the facility in accordance with the proposed

amendments would not involve a significant reduction in a margin of

safety.

The changes to valve tolerances do not involve a reduction in a

margin of safety because the margin of safety associated with the

MSSVs and the pressurizer safety valves, as verified by the results

of the accident analyses and valve evaluations, are within

acceptable limits. Transients impacted by this change have been

analyzed and have met the applicable accident analyses acceptance

criteria, such as those applicable to primary and secondary side

pressure. The margin of safety required for each affected safety

analysis is maintained. This conclusion is not changed by the valve

tolerances for the main steam safety valves and the pressurizer

safety valves. Therefore, the changes do not involve a significant

reduction in the margin of safety.

OPERATION AT REDUCED POWER WITH INOPERABLE MAIN STEAM SAFETY

VALVES

(1) Operation of the facility in accordance with the proposed

amendments would not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The proposed changes do not involve a significant increase in

the probability or consequences of an accident previously evaluated.

The proposed maximum allowable power level values will ensure that

the secondary side steam pressure will not exceed 110 percent of the

design pressure following a Loss of Load/Turbine Trip event, when

one or more main steam safety valves (MSSVs) are declared

inoperable. The proposed change will not impact the classification

of the Loss of Load/Turbine Trip event as a Condition II probability

event (faults of moderate frequency) per ANSI - N18.2, 1973.

Accordingly, since the proposed maximum allowable power level will

maintain the capability of the MSSVs to perform their pressure

relief function associated with a Loss of Load/Turbine Trip event,

there will be no effect on the probability or consequences of an

accident previously evaluated.

(2) Operation of the facility in accordance with the proposed

amendments would not create the possibility of a new or different

kind of accident from any accident previously evaluated.

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

different kind of accident from any accident previously evaluated.

The proposed changes do not involve any change to the configuration

of any plant equipment, and no new failure modes have been defined

for any plant system or component. The proposed maximum allowable

power level as specified in TS Table 3.7-1 will improve the

capability of the MSSVs to perform their pressure relief function to

ensure the secondary side steam pressure does not exceed 110 percent

of design pressure following a Loss of Load/Turbine Trip event.

Therefore, since the function of the MSSVs is improved by the

proposed changes, the possibility of a new or different kind of

accident from any accident previously evaluated is not created.

(3) Operation of the facility in accordance with the proposed

amendments would not involve a significant reduction in a margin of

safety.

The proposed changes to the Technical Specifications do not

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

methodology used to calculate the maximum allowable power level is

conservative and bounding since it is based on a number of

inoperable MSSVs per loop; i.e., if only one MSSV in one loop is out

of service, the required action to reduce power to the maximum

allowable power level would be the same as if one MSSV in each loop

were out of service. Another conservatism with the algorithm

methodology is with the assumed minimum total steam flow rate

capability of the operable MSSVs. The assumption is that if one or

more MSSVs are inoperable per loop, the inoperable MSSVs are the

largest capacity MSSVs, regardless of which capacity MSSVs are

actually inoperable.

Therefore, since the maximum allowable power level calculated

for the proposed changes using the algorithm methodology are more

conservative and ensure that 110 percent of secondary side steam

pressure is not exceeded following a Loss of Load/Turbine Trip

event, this proposed license amendment will not involve a

significant reduction in a margin of safety.

SERVICE PERIOD FOR HEATUP AND COOLDOWN PRESSURE-TEMPERATURE

LIMIT CURVES

(1) Operation of the facility in accordance with the proposed

amendments would not involve a significant increase in the

probability or consequences of an accident previously evaluated.

Calculation of the service period for the heatup and cooldown

curves does not involve an increase in the probability or

consequences of an accident previously evaluated because the

calculations were completed to verify the adequacy of the existing

curves and to determine an appropriate service period. The use of

approved methods and the acceptable results have shown that no

design or analysis criteria are changed. The structural and

functional integrity of the reactor vessel has been verified.

No fission product barriers or inputs to dose analyses are

adversely affected by these calculations and reverification of the

existing heatup/cooldown curves. Therefore, the probability or

consequences of an accident previously evaluated are not increased.

(2) Operation of the facility in accordance with the proposed

amendments would not create the possibility of a new or different

kind of accident from any accident previously evaluated.

The revised service period does not create the possibility of a

new or different kind of accident from any accident previously

evaluated because the recalculation of an acceptable service period

does not affect accident initiation sequences. No new operating

configuration is being imposed by the calculations that would create

a new failure scenario. In addition, no new failure modes or

limiting single failures have been identified. Therefore, the types

of accidents defined in the UFSAR continue to represent the credible

spectrum of events to be analyzed which determine safe plant

operation. Therefore, it is concluded that no new or different kind

of accident from any accident previously evaluated have been created

as a result of these revisions.

(3) Operation of the facility in accordance with the proposed

license amendments would not involve a significant reduction in a

margin of safety.

Calculations were performed to determine the service period

appropriate for the existing curves. The changes to service period

do not involve a reduction in a margin of safety because the margin

of safety associated with the heatup/cooldown curves, as verified by

the results of the analyses, are unchanged. Therefore, the proposed

change to the service period does not involve a significant

reduction in the margin of safety.

MODIFICATION TO SURVEILLANCE REQUIREMENT FOR EMERGENCY

CONTAINMENT COOLING SYSTEM

(1) Operation of the facility in accordance with the proposed

amendments would not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The purpose of the ECC units is to help mitigate the

consequences of an accident (i.e., to help maintain the containment

pressure and temperature within their design

[[Page 34891]]

values following a design basis accident). The ECC units do not

operate during normal operation of the plant. Failure of the ECC

units would not initiate a plant transient or accident. Therefore,

the proposed change involving the ECC units would not affect the

probability of occurrence of an accident previously evaluated.

Evaluations demonstrate that, with two ECC units operating

during a LOCA or MSLB [main steamline break], the containment

pressure and temperature will be maintained within their design

values. These evaluations also demonstrate that, with two ECC units

operating during a LOCA or MSLB, the temperature of the CCWS

[component cooling water system] will be maintained within its

design temperature. Therefore, the proposed change involving the ECC

units would not affect the consequences of an accident previously

evaluated.

(2) Operation of the facility in accordance with the proposed

amendments would not create the possibility of a new or different

kind of accident from any accident previously evaluated.

The purpose of the ECC units is to mitigate design basis

accidents, and failure of the ECC units would not cause a plant

transient or accident. Furthermore, a single failure of an ECC unit

during a LOCA or MSLB would not lead to a new or different kind of

accident. Although the revised Technical Specifications require two

ECC units to start automatically on a LOCA signal, they would also

require that all three ECC units be operable. On a single failure of

an operating ECC unit, there would be sufficient time to start the

standby ECC unit to accomplish the design function of the ECC

system. Therefore, the proposed amendment would not create the

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

previously evaluated.

(3) Operation of the facility in accordance with the proposed

amendments would not involve a significant reduction in a margin of

safety.

The proposed change in the actuation logic of the ECC units

would not cause either the containment pressure and temperature or

the CCWS temperature to exceed their design values. While the energy

released into containment and subsequently transferred to the CCWS

will increase as a result of the thermal uprate, this increase is

insignificant and will not result in either the containment or CCWS

exceeding a design limit. Therefore, the proposed change would not

affect the margin of safety.

CONTROL ROOM EMERGENCY VENTILATION SYSTEM

(1) Operation of the facility in accordance with the proposed

amendments would not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The proposed change does not affect the integrity of the fission

product barriers utilized for mitigation of dose consequences as a

result of an accident. Only the iodide removal efficiency of the

control room emergency ventilation system is increased, and this

change is in the conservative direction.

To assure consistency between testing efficiency and analysis

assumptions for post-accident control room doses, the methyl iodide

removal efficiency required to be demonstrated by laboratory test,

is being increased from 90% to 99%. This increase in testing

efficiency is consistent with the recommendations set by the NRC

staff in Regulatory Guide 1.52 to support analysis efficiencies for

elemental iodine and methyl iodide removal of 95%, respectively.

Testing performed to verify methyl iodide removal efficiency will be

performed under conditions representative of the control room

environment.

Since this change in removal efficiency is in the conservative

direction, plant safety will not be adversely impacted.

(2) Operation of the facility in accordance with the proposed

amendments would not create the possibility of a new or different

kind of accident from any accident previously evaluated.

The proposed change to the control room emergency ventilation

system iodide removal efficiency does not create the possibility of

a new or different kind of accident from any accident previously

evaluated because operation of the control room emergency

ventilation system is not identified in any accident initiation

sequence. The system is provided to minimize operator exposure to

airborne radioactivity released as a result of an accident. The new

operating configuration has been evaluated and no new limiting

single failures have been identified as a result of the proposed

modification. Therefore, it is concluded that no new or different

kind of accidents from any accident previously evaluated have been

created as a result of these revisions.

(3) Operation of the facility in accordance with the proposed

amendments would not involve a significant reduction in a margin of

safety.

The proposed changes do not involve a reduction in the margin of

safety because the margin of safety associated with this change is

in the conservative direction. Thus, plant safety will not be

adversely impacted and the margin of safety required for the

affected safety analysis is maintained. The adequacy of the revised

Technical Specification values to maintain the plant in a safe

operating condition has been confirmed, since the testing will be

done to a more conservative criteria (i.e., 99% efficiency).

Therefore, the proposed changes do not involve a significant

reduction in the margin of safety.

RELOCATION OF FQ(Z) [HEAT FLUX HOT CHANNEL FACTOR] AND F

Delta H [NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR] LIMITS FROM

TECHNICAL SPECIFICATIONS TO CORE OPERATING LIMITS REPORT AND

EDITORIAL CORRECTIONS

(1) Operation of the facility in accordance with the proposed

amendments would not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The relocation of the values for FQ and F Delta H from the

Technical Specifications to the Core Operating Limits Report is

administrative in nature and has no impact on the probability or

consequences of any Design Bases Event (DBE) occurrence which was

previously evaluated. The determination of the FQ and F Delta H

limits will be performed using methodology approved by the NRC and

poses no significant increase in the probability or consequences of

any accident previously evaluated.

The changes being proposed as editorial in nature do not affect

assumptions contained in the safety analyses, the physical design

and/or operation of the plant, nor do they affect Technical

Specifications that preserve safety analysis assumptions. Therefore,

these proposed changes do not affect the probability or consequences

of accidents previously analyzed.

(2) Operation of the facility in accordance with the proposed

amendments would not create the possibility of a new or different

kind of accident from any accident previously evaluated.

The relocation of the FQ and F Delta H limits from the

Technical Specifications to the Core Operating Limits Report is

administrative in nature and has no impact, nor does it contribute

in any way to the possibility of a new or different kind of accident

from any accident previously evaluated.

The determination of the FQ and F Delta H limits will be

performed using NRC-approved methodology and are submitted to the

NRC as a revision to the COLR to allow the NRC staff to trend

peaking factors. The Technical Specifications will continue to

require operation within the required core operating limits and

appropriate actions will be taken if the FQ and F Delta H

limits are exceeded. Therefore, the proposed amendments does not in

any way create the possibility of a new or different kind of

accident from any accident previously evaluated.

The editorial changes proposed are administrative in nature and

do not affect assumptions contained in plant safety analyses, the

physical design and/or operation of the facility, nor do they affect

Technical Specifications that preserve safety analysis assumptions.

Therefore, these changes do not create the possibility of a new or

different kind of accident.

(3) Operation of the facility in accordance with the proposed

amendments would not involve a significant reduction in a margin of

safety.

The relocation of the FQ and F Delta H limits from the

Technical Specifications to the Core Operating Limits Report is

administrative in nature and has no impact on the margin of safety.

The determination of the FQ and F Delta H limits will be

performed using methodology approved by the NRC and does not

constitute a significant reduction in the margin of safety.

The supporting Technical Specification values are defined by the

accident analyses which are performed to conservatively bound the

operating conditions defined by the Technical Specifications.

Performance of analysis and evaluation have confirmed that the

operating envelope defined by the Technical Specifications continues

to be bounded by the analytical basis, which in no case exceeds the

acceptance limits. Therefore, the margin of safety provided in the

analyses in accordance with the acceptance limits is maintained and

not significantly reduced.

[[Page 34892]]

The changes being proposed as editorial in nature do not relate

to or modify the safety margins defined in, and maintained by the

Technical Specifications. Therefore, the proposed changes which

correct administrative errors and clarify existing Technical

Specification requirements do not involve any reduction in a margin

of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Florida International

University, University Park, Miami, Florida 33199

Attorney for licensee: J. R. Newman, Esquire, Morgan, Lewis &

Bockius, 1800 M Street, NW., Washington, DC 20036

NRC Project Director: Frederick J. Hebdon

Florida Power and Light Company, Docket Nos. 50-250 and 50-251,

Turkey Point Plant Units 3 and 4, Dade County, Florida

Dates of amendment request: April 19, 1996, May 10, 1996, and May

28, 1996

Description of amendment request: The licensee proposed to change

the Turkey Point Units 3 and 4 Technical Specifications (TS) to address

frequency extension for actions required on a periodic basis, delete

the separate notification requirement for an inoperable startup

transformer, and allow the operating RHR loop to be removed from

operation during refueling operations under certain conditions.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below.

(1) Operation of the facility in accordance with the proposed

amendments would not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The proposed amendments do not involve a significant increase in

the probability or consequences of an accident previously evaluated

because the proposed amendments are purely administrative in nature.

These amendments will not involve a significant increase in the

probability or consequences of an accident previously evaluated

because they do not affect assumptions contained in plant safety

analyses, the physical design and/or operation of the plant, nor do

they affect Technical Specifications that preserve safety analysis

assumptions. Therefore, the proposed changes do not affect the

probability or consequences of accidents previously analyzed.

(2) Operation of the facility in accordance with the proposed

amendments would not create the possibility of a new or different

kind of accident from any accident previously evaluated.

The use of the modified specifications can not create the

possibility of a new or different kind of accident from any

previously evaluated since the proposed amendments will not change

the physical plant or the modes of plant operation defined in the

facility operating license. No new failure mode is introduced due to

the administrative changes and clarifications, since the proposed

changes do not involve the addition or modification of equipment nor

do they alter the design or operation of affected plant systems,

structures, or components.

(3) Operation of the facility in accordance with the proposed

amendments would not involve a significant reduction in a margin of

safety.

The operating limits and functional capabilities of the affected

systems, structures, and components are unchanged by the proposed

amendments. The modified specifications which correct administrative

errors and clarify existing Technical Specification requirements do

not significantly reduce any of the margins 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.

Local Public Document Room location: Florida International

University, University Park, Miami, Florida 33199

Attorney for licensee: J. R. Newman, Esquire, Morgan, Lewis &

Bockius, 1800 M Street, NW., Washington, DC 20036

NRC Project Director: Frederick J. Hebdon

Gulf States Utilities Company, Cajun Electric Power Cooperative,

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

Station, Unit 1, West Feliciana Parish, Louisiana

Date of amendment request: May 30, 1996

Description of amendment request: The proposed amendment would

revise the technical specifications surveillance requirement (SR)

3.8.3.4 to specify a 5-start pressure for the air receivers associated

with the Division III, High Pressure Core Spray emergency diesel

generator.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Does the change involve a significant increase in the

probability or consequence of an accident previously evaluated?

The purpose of the proposed Technical Specification change is to

establish consistency between the basis for the air start pressure

required for the Division I and II diesels and the value required

for the Division III diesel. The value of 160 psig currently

specified in SR 3.8.3.4 is representative of a 5-start value for the

Division I and II diesels, however, this value is not representative

of a 5-start for the Division III diesel. While the 160 psig value

does serve to satisfy the requirements of 10 CFR 50.36 with regard

to maintaining the lowest functional level required for the Division

III diesel to perform its design safety function, the current value

does not serve to maintain the design margin utilized when sizing

the air receivers for the purpose of satisfying the Standard Review

Plan guidance contained in section 9.5.6 (NUREG-0800 Revision 2).

The proposed value fully complies with the guidance provided in

NUREG-0800 and is more conservative than the value currently

included in the Technical Specifications. The proposed value is well

within the capability of the air system's design and will not

subject the air system to excessive pressures or undue cycling of

the system's compressors. The proposed change has no effect on the

probability of an accident as diesel generators have no bearing on

the initiation of any analyzed event. In addition, the capability of

the Division III diesel to perform its design basis function (i.e.,

starting, accelerating to rated speed and voltage, and connecting to

its respective bus within 13 seconds) is not affected by this

change. The ability of the diesel to support the mitigation of

analyzed accidents is not affected and hence the consequences of any

analyzed event are not affected. Therefore, the proposed change does

not increase the probability or the consequences of previously

analyzed accidents.

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

kind of accident from any accident previously evaluated?

The proposed change does not introduce any new failure modes.

All of the affected components remain within their applicable design

limits. In addition, the environmental qualification of any plant

equipment is not adversely affected by the proposed change. Since

the performance of this system is not adversely affected by this

change and the design margins of this system are not challenged in a

manner differently than previously analyzed, the proposed change

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

accident from any accident previously evaluated.

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

of safety?

The proposed change raises the required starting air pressure

for the Division III above that currently required by the Technical

Specifications to establish consistency between the basis of the

Division III value with the value used for the Division I and II

diesels. Issuance of the proposed change will establish a 5 start

air receiver pressure for each of the three safety-related diesels

at

[[Page 34893]]

River Bend. While the proposed value is slightly less than the 5

start value discussed in River Bend's SER, the proposed value is

supported by the River Bend site-specific test data and does not

adversely affect existing analyses or system performance. Therefore,

the proposed change does not 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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Government Documents

Department, Louisiana State University, Baton Rouge, LA 70803

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

1400 L Street, NW., Washington, DC 20005

NRC Project Director: William D. Beckner

Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear

Station, Nemaha County, Nebraska

Date of amendment request: June 6, 1996, as supplemented by letters

dated June 7 and 9, 1996

Description of amendment request: The proposed amendment would

revise the technical specification Limited Safety System Setting for

the MINIMUM CRITICAL POWER RATIO (MCPR) for dual recirculation loop

operation and for single recirculation loop 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. Does the proposed license amendment involve a significant

increase in the probability or consequences of an accident

previously evaluated?

The purpose of the Safety Limit Minimum Critical Power Ratio

(SLMCPR) is to provide statistical confidence that less than 0.1% of

the fuel rods in a core would experience transition boiling during

the most limiting analyzed Anticipated Operational Occurrence

(transient). While transition boiling in a BWR does not in and of

itself signal the onset of fuel cladding failure, this criterion has

been selected as a conservative and convenient parameter for the

evaluation of fuel designs. Therefore, while this safety limit does

not provide any control over either the probability or consequences

of any accident previously evaluated, it does ensure that evaluated

transients remain within NRC-approved criteria. Revision of the

SLMCPR will establish in the CNS Technical Specifications a valid

limit, based on the NRC approved GESTAR II methodology using cycle-

specific inputs. This change will result in the input of more

restrictive core operating limits into the plant process computer,

ensuring that CNS will be operated within the constraints of the new

SLMCPR limits of 1.07 for dual recirculation loop operation, and

1.08 for single recirculation loop operation. No plant hardware

modifications are associated with this change. Therefore, since this

proposed change will not change the physical configuration of the

plant, nor result in operational changes which invalidate

assumptions used in any CNS accident analysis, this change does not

involve an increase in the probability or consequences of any

accident previously evaluated.

2. Does the proposed License Amendment create the possibility of

a new or different kind of accident from any accident previously

evaluated?

This change revises the SLMCPR values in the CNS Technical

Specifications in accordance with a cycle specific analysis

performed for the remainder of the current cycle. The SLMCPR ensures

that less than 0.1% of the fuel rods in a core would experience

transition boiling during the most limiting Anticipated Operational

Occurrence. Increasing the SLMCPR from 1.06 to 1.07 for dual

recirculation loop operation and from 1.07 to 1.08 for single

recirculation loop operation will ensure that the specified

statistical confidence will be met for all analyzed transients. This

change does not involve any plant hardware changes. The only

operational changes will be the institution of appropriate thermal

restrictions on reactor core operation in accordance with the SLMCPR

changes. Therefore, this proposed change will not create the

possibility of a new or different kind of accident from any

previously evaluated.

3. Does the proposed change create a significant reduction in

the margin of safety?

This change will establish in the CNS Technical Specifications,

SLMCPR values that ensure the margin of safety to the NRC approved

Anticipated Operational Occurrence evaluation acceptance criteria

will be met. Increasing the SLMCPR institutes more restrictive

thermal limitations on core operation. The change of the SLMCPR from

1.06 to 1.07 for dual recirculation loop operation, and from 1.07 to

1.08 for single loop operation will ensure that the acceptance

criteria for evaluated transients will continue to be met, and that

the appropriate limit is reflected in the CNS Technical

Specifications. Therefore, this proposed change does not create 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: Auburn Memorial Library, 1810

Courthouse Avenue, Auburn, NE 68305

Attorney for licensee: Mr. John R. McPhail, Nebraska Public Power

District, Post Office Box 499, Columbus, NE 68602-0499

NRC Project Director: William D. Beckner

Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile

Point Nuclear Station, Unit 2, Oswego County, New York

Date of amendment request: May 15, 1996

Description of amendment request: The proposed amendment would

revise Technical Specification (TS) 3/4.3.2, ``Isolation Actuation

Instrumentation,'' to establish a range of allowable and trip setpoints

for high temperature (varying as a function of ambient temperature) in

the Main Steam Line Tunnel Lead Enclosure Area. Specifically, a new TS

Figure 3.3.2-1 would be added to provide a curve of allowable

temperature values and a curve of trip temperature setpoints, both

plotted over a range of ambient temperatures. The new Figure would be

referenced by Table 3.3.2-2 at item 1.d.3 (High Temperature Main Steam

Line Tunnel Lead Enclosure Trip Function) by a new footnote stating:

The trip setpoint and allowable value for a channel may be

established based on Figure 3.3.2-1, if:

a. The actual ambient temperature readings for all operable

channels in the Lead Enclosure Area are equal to or greater than the

ambient temperature used as the basis for the setpoint, and

b. The absence of steam leaks in the Main Steam Line Tunnel Lead

Enclosure Area is verified by visual inspection prior to increasing

a channel setpoint, and

c. A surveillance is implemented in accordance with Note (d) of

Table 4.3.2.1-1.

Similarly, TS Surveillance Table 4.3.2.1-1 would be supplemented at

item 1.d.3 (High Temperature Main Steam Line Tunnel Lead Enclosure)

with a new footnote stating:

(d) In addition to the normal shift channel check, if a channel

setpoint has been established using Figure 3.3.2-1, then once per

shift, the actual ambient temperature reading for all operable

channels in the Lead Enclosure Area shall be verified to be equal to

or greater than the ambient temperature used as the basis for the

setpoint.

Basis for proposed no significant hazards consideration

determination: The main steam tunnel high temperature isolation

actuation instrumentation is part of the Leak Detection System (LDS).

It is used to detect leakage early at 25 gallons per minute (gpm) and

initiate signals to automatically close the Main Steam Isolation Valves

before a pipe break could occur. The existing temperature setpoints for

the tunnel lead enclosure are based upon transient analyses for steam

leaks in the steam tunnel utilizing

[[Page 34894]]

winter temperatures as an initial condition. The licensee finds that a

change is needed because actual temperatures in the tunnel, especially

during the summer, are approaching the setpoints when steam leakage is

not occurring. Under the present conditions, a minor disturbance in the

turbine building ventilation system could cause an unwarranted

isolation actuation at full power with resulting Main Steam Isolation

Valve closure and reactor scram.

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 operation of NMP2 [Nine Mile Point Unit 2] in accordance

with the proposed amendment will not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The LDS instrumentation in the main steam line tunnel isolates

the Main Steam Isolation Valves upon sensing a steam leak of 25 gpm.

For an elevated ambient temperature in the Lead Enclosure area, a

setpoint established using the proposed Figure 3.3.2-1 ensures that

the Main Steam Isolation Valves continue to receive an isolation

signal upon sensing a steam leak of 25 gpm. Verifying the absence of

any steam leak in the area prior to raising any temperature

instrument setpoint ensures that the ability to sense a 25 gpm leak

is not compromised by an increased ambient temperature resulting

from a smaller steam leak. The periodic surveillance to verify the

actual ambient temperature ensures the continued validity of the

ambient temperature used for the setpoint basis, and provides

sufficient advance indication to take appropriate compensatory

action. Accordingly, this change will not involve a significant

increase in the consequences of any accident previously evaluated.

Furthermore, the LDS function provides a mitigation action for a

postulated main steam line pipe leak which could lead to a pipe

break. This function does not affect any accident precursors, and

the proposed change does not affect the function of the LDS system.

Accordingly, this change will not involve a significant increase in

the probability of any accident previously evaluated.

2. The operation of NMP2 in accordance with the proposed

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

of accident from any previously evaluated.

The qualification of safety-related equipment in the main steam

lead enclosure is evaluated using actual temperatures and component

qualified life is adjusted accordingly. The temperature elements are

the only safety-related equipment affected by this change,

therefore, the instrumentation response to previously evaluated

accidents will not be adversely affected. This change will not

affect the performance of safety related structures. Accordingly,

the design capabilities of those structures, systems and components

affected by the proposed change are not challenged in a manner not

previously evaluated so as to create the possibility of a new or

different kind of accident from any previously evaluated.

3. The operation of NMP2 in accordance with the proposed

amendment will not involve a significant reduction in a margin of

safety.

The proposed change provides a range of setpoints and allowable

values for the Main Steam Line Tunnel Lead Enclosure temperatures.

The calculation of the allowable values and trip setpoints was

performed using the same methodologies as previously employed. For

an elevated ambient temperature in the Lead Enclosure area, a

setpoint established using the proposed Figure 3.3.2-1 ensures that

the Main Steam Isolation Valves receive an isolation signal upon

sensing a steam leak of 25 gpm, resulting in a main steam line

isolation prior to a pipe break. Therefore, the proposed change

provides the same level of protection against a main steam line

break as the existing setpoint values. The proposed setpoints will

provide increased scram avoidance, and thereby reduce unnecessary

challenges to the plant shutdown systems. Accordingly, the proposed

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

safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Reference and Documents

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

York 13126

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

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

NRC Project Director: Jocelyn A. Mitchell, Acting Director

PECO Energy Company, Public Service Electric and Gas Company,

Delmarva Power and Light Company, and Atlantic City Electric

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

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

Date of application for amendments: March 25, 1996

Description of amendment request: These amendments revise the

safety limit minimum critcal power ratios (SLMCPRs) to support use of

GE-13 fuel at Peach Bottom Atomic Power Station.

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 TS [technical specification] changes do not

involve a significant increase in the probability or consequences of

an accident previously evaluated.

The derivation of the revised GE13 SLMCPRs for incorporation

into the TS, and its use to determine cycle-specific thermal limits,

have been performed using USNRC [U.S. Nuclear Regulatory

Commission]-approved methods within the existing fuel licensing

criteria as discussed in NEDE-32198P, ``GE13 Compliance With

Amendment 22 of NEDE-24011-P-A (GESTAR II),'' and cannot increase

the probability or severity of an accident.

The basis of the SLMCPRs calculation is to ensure that greater

than 99.9% of all fuel rods in the core avoid boiling transition if

the limit is not violated. The new SLMCPRs preserve the existing

margin to transition boiling and fuel damage in the event of a

postulated accident. The fuel licensing acceptance criteria for the

SLMCPRs calculation apply to the GE13 fuel in the same manner that

they have applied to previous fuel designs. The probability of fuel

damage is not increased. Therefore, the proposed TS changes do not

involve an increase in the probability or consequences of an

accident previously evaluated.

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

new or different kind of accident from any accident previously

evaluated.

The SLMCPR for the GE13 fuel design is a Technical Specification

numerical value, designed to ensure that transition boiling does not

occur in 99.9% of all fuel rods in the core during the limiting

postulated accident. It cannot create the possibility of any new

type of accident. The new SLMCPRs are calculated using USNRC-

approved methods and have the same calculational basis as the SLMCPR

for other GE fuel designs previously used at PBAPS, Units 2 and 3.

Therefore, the proposed TS changes do not create the possibility of

a new or different kind of accident, from any accident previously

evaluated.

3) The proposed TS changes do not involve a significant

reduction in a margin of safety.

The margin of safety as defined in the TS Bases will remain the

same. The new SLMCPRs are calculated using USNRC-approved methods

which are in accordance with the current fuel licensing criteria.

The SLMCPRs for the GE13 fuel remain high enough to ensure that

greater than 99.9% of all fuel rods in the core will avoid boiling

transition if the limit is not violated, thereby preserving the fuel

cladding integrity. Therefore, the proposed TS changes do not

involve a reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Government Publications

[[Page 34895]]

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

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

Pennsylvania 17105

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

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

Pennsylvania 19101

NRC Project Director: John F. Stolz

PECO Energy Company, Public Service Electric and Gas Company,

Delmarva Power and Light Company, and Atlantic City Electric

Company, Docket No. 50-277, Peach Bottom Atomic Power Station, Unit

No. 2, York County, Pennsylvania

Date of application for amendment: June 13, 1996

Description of amendment request: The proposed amendment to the

Technical Specifications (TS) will permit a one time performance of

Surveillance Requirement 3.3.1.1.12, for the Average Power Range

Monitor Flow Biased High Scram function, with a delayed entry into its

associated TS Conditions and Required Actions for up to 6 hours

provided core flow is maintained at or above 82 percent. This change

would be in effect until the end of refueling outage 2R11, currently

scheduled for early October 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:

i) The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The APRM system provides monitoring and accident mitigation

functions to limit peak flux in the core during startup and run

modes. This proposed TS change for delaying entry into Conditions

and Required Actions associated with SR 3.3.1.1.12 for the APRM flow

bias function will have no impact on the APRM system or any system

that interfaces with it. No pressure boundary interfaces or process

control parameters will be challenged.

This change does not affect the operation of any equipment.

Delaying entry into Conditions and Required Actions associated with

SR 3.3.1.1.12 does not affect either the initiator of any accident

previously evaluated or any equipment required to mitigate the

consequences of an accident, or the isotopic inventory in the fuel.

Thus, the change does not increase either the probability or the

consequences of accidents previously evaluated.

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

or different kind of accident from any previously evaluated.

Because there is no direct pressure boundary interface or

process control function associated with the APRM system or its

interfacing electronics, the possibility of a new or different type

of accident than any previously evaluated will not be created.

Although the flow bias instrument loop does employ flow transmitters

to measure recirculation drive flow, delaying entry into Conditions

and Required Actions associated with SR 3.3.1.1.12 will have no

impact on their pressure boundary function. Also, failure of the

sensing line associated with these transmitters has already been

accounted for in the initial plant design by including excess flow

check valves for sensing line break isolation.

The proposed change does not introduce a new mode of plant

operation and does not involve the installation of any new equipment

or modifications to the plant. Therefore, it does not create the

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

previously evaluated.

iii)The proposed change does not involve a significant reduction

in a margin of safety.

The APRM flow biased high scram function is not specifically

credited in the safety analysis. However, it is intended to provide

an additional margin of protection from transient induced fuel

damage during operation where recirculation flow is reduced to below

the minimum required for rated power operation.

The margin of safety associated with this change refers to the

margin inherent in the accident analyses that takes credit for the

clamped high flux scram only (i.e., margin between scramming at 120%

peak flux and the peak flux necessary for fuel damage). The current

reactor operating state (end of cycle coast down extended core flow)

dictates that only the 120% flux trip be enforced. This trip remains

functional during the APRM flow biased high scram calibration.

Currently, the Conditions and Required Actions associated with

SR 3.3.1.1.12 permit a one hour delay prior to entry because it

minimizes risk while allowing time for restoration or tripping of

channels by operations personnel. Because the APRM flow biased

function is not enforced during end of cycle, coast down, extended

core flow conditions, extending entry in associated Conditions and

Required Actions from one to six hours has no impact on the margin

associated with the clamped high flux scram. In the event core flow

drops below 82%, the flow point below which APRM setpoints

automatically become flow biased, the associated Conditions and

Required Actions will be entered.

Therefore, extending entry into associated Conditions and

Required Actions associated with SR 3.3.1.1.12, provided core flow

remains at or above 82%, from one to six hours does not reduce any

margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Government Publications

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

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

PA 17105

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

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

PA 19101

NRC Project Director: John F. Stolz

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

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

Date of amendment request: May 16, 1996

Description of amendment request: The proposed amendment to the

James A. FitzPatrick Technical Specifications (TSs) proposes to delete

the requirement for the Plant Operating Review Committee (PORC) to

review the fire protection program and implementing procedures. This

proposal will reduce the administrative burden on the committee while

making PORC's responsibilities more consistent with the other

responsibilities described in Section 6.1.5.6 of the TS.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

Operation of the FitzPatrick plant in accordance with the

proposed amendment would not involve a significant hazards

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

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed changes delete the Plant Operating Review Committee

(PORC) review of the fire protection program and implementing

procedures, and deleted fire protection inspection and audit

requirements that are redundant to those performed under the

cognizance of the Safety Review Committee (SRC). The changes do not

introduce any new modes of plant operation, make any physical

changes, or alter any operational setpoints. Therefore, the changes

do not degrade the performance of any safety system assumed to

function in the accident analysis. Consequently, there is no effect

on the probability or consequences of an accident.

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

from those previously evaluated.

No physical changes to the plant or changes to equipment

operating procedures are proposed. The changes are administrative

and will not have any direct affect on equipment important to

safety. Therefore the changes cannot create the possibility of a new

or different kind of accident.

[[Page 34896]]

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

Adequacy of the fire protection program and implementing

procedures is assured by the fire protection license condition, the

procedure review and approval process implemented by Amendment 222,

the provisions of 10 CFR 50.59, and inspections and audits performed

under the cognizance of the SRC. Consequently, deleting PORC's

responsibility for review of the fire protection program and

implementing procedures, and deleting the inspection and audit

requirements contained in Specification 6.14.A and 6.14.B will not

degrade the fire protection program. Therefore, the proposed changes

do not involve a significant reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Reference and Documents

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

York 13126

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

York, New York 10019

NRC Project Director: Jocelyn A. Mitchell, Acting Director

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

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

Date of amendment request: May 30, 1996

Description of amendment request: The proposed amendment would

revise Minimum Critical Power Ratio Safety Limit and associated basis.

The changes are required to support introduction of General Electric

Company supplied, GE12, 10x10 fuel into the Cycle 13 reactor core.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

Operation of the FitzPatrick plant in accordance with the

proposed Amendment would not involve a significant hazards

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

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated because:

A change in the SLMCPR [Safety Limit Minimum Critical Power

Ratio] does not affect initiation of any accident. Operation in

accordance with the revised SLMCPR ensures the consequences of

previously analyzed accidents are not changed.

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

from any accident previously evaluated because:

The SLMCPR establishes a performance limit for the fuel.

Therefore changing the limit will not initiate any accident.

3. Involve a significant margin of safety because:

The analyses performed to determine the revised SLMCPR assure

maintenance of the same margin of safety as presently exists for the

prevention of onset of transition boiling.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Reference and Documents

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

York 13126

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

York, New York 10019

NRC Project Director: Jocelyn A. Mitchell, Acting Director

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

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

Date of amendment request: May 30, 1996

Description of amendment request: The proposed amendment would

revise Anticipated Transient Without Scram (ATWS) Recirculation Pump

Trip Reactor Pressure - High setpoint when either zero or one Safety

Relief Valves are out-of service.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

Operation of the FitzPatrick plant in accordance with the

proposed Amendment would not involve a significant hazards

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

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated because:

A change in the ATWS Recirculation Pump Trip Reactor Pressure -

High setpoint does not affect initiation of any accident. Operation

in accordance with the revised setpoints ensures the consequences of

previously analyzed accidents are not changed.

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

from any accident previously evaluated because:

RPV [reactor pressure vessel] pressure following an ATWS with

MSIV [main steam isolation valve] closure event (worst case

transient for RPV pressurization) remains within acceptable limits

with the revised setpoint. Therefore changing the setpoint will not

lead to a new type of accident.

3. Involve a significant reduction in a margin of safety

because:

The analyses performed to determine the revised ATWS

Recirculation Pump Trip Reactor Pressure - High setpoint assure

maintenance of the same margin of safety as presently exists for

limiting RPV pressure following an ATWS with MSIV closure (limiting

transient).

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Reference and Documents

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

York 13126

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

York, New York 10019

NRC Project Director: Jocelyn A. Mitchell, Acting Director

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

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

Date of amendment request: May 30, 1996

Description of amendment request: The proposed amendment would

eliminate selected response time testing requirements. The affected

Technical Specifications (TS) are TS 4.1.A, ``Surveillance

Requirements, Reactor Protection System,'' and TS 4.2.A, ``Surveillance

Requirements, Instrumentation, Primary Containment Isolation

Functions.''

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

Operation of the FitzPatrick plant in accordance with the

proposed Amendment would not involve a significant hazards

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

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated because:

The purpose of the proposed TS change is to eliminate response

time testing requirements for selected sensors in the RPS [reactor

protection system] and Primary Containment Isolation System. The

BWROG [Boiling Water Reactor Owners Group] has completed an

evaluation which demonstrates that response time testing is

redundant to the other TS required testing. These other tests

[[Page 34897]]

in conjunction with actions taken in response to NRC Bulletin 90-01,

``Loss of Fill-Oil in Transmitters

Manufactured by Rosemount,'' and Supplement 1 to Bulletin 90-01,

are sufficient to identify failure modes or degradation in

instrument response time and ensure operation of the associated

systems within acceptable limits. Furthermore, failure modes

detected by response time testing are detectable by other TS

required testing. This evaluation was documented in Reference 1 [See

application dated May 30, 1996]. NYPA [New York Power Authority] has

confirmed the applicability of this evaluation to the FitzPatrick

Plant. In addition, NYPA will complete the actions identified in the

NRC staff's safety evaluation of NEDO-32291-A.

Because of the continued application of other existing TS

required tests such as channel calibrations, channel checks, channel

functional tests, and logic system functional tests, the response

time of these systems will be maintained within the acceptance

limits assumed in plant safety analyses and required for successful

mitigation of an initiating event. The proposed changes do not

affect the capability of the associated systems to perform their

intended function within their required response time, nor do the

proposed changes themselves affect the operation of any equipment.

Therefore, the proposed change does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

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

from those previously evaluated because:

The proposed changes do not affect the ability of the systems to

perform their intended function within the acceptance limits assumed

in plant safety analyses and required for successful mitigation of

an initiating event. No new failure modes are introduced by the

changes. Therefore, the proposed changes do not create the

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

previously evaluated.

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

The current TS required response time test limits are based on

the maximum allowable values assumed in the plant safety analyses.

These analyses conservatively establish the margin of safety. As

described above, the proposed changes do not affect the capability

of the associated systems to perform their intended function within

the allowed response time used as the basis for the plant safety

analysis. Plant and system response to an initiating event will

remain in compliance within the assumptions of the safety analyses,

and therefore the margin of safety is not affected.

Further, although not explicitly evaluated, the proposed changes

will provide an improvement to plant safety and operation by

reducing the time safety systems are unavailable, reducing safety

systems actuations, reducing plant shutdown risk, limiting radiation

exposure to plant personnel, and eliminating the diversion of key

personnel to conduct unnecessary testing. Therefore, the overall

effect of the changes should increase the margin the 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.

Local Public Document Room location: Reference and Documents

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

York 13126

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

York, New York 10019

NRC Project Director: Jocelyn A. Mitchell, Acting Director

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

Creek Generating Station, Salem County, New Jersey

Date of amendment request: March 6, 1996, as supplemented by letter

dated May 30, 1996

Description of amendment request: The proposed change to Hope Creek

Technical Specification (TS) 3.8.1, ``A.C. Sources - Operating'', would

decrease the minimum fuel oil storage capacity of the Emergency Diesel

Generator Fuel Oil Storage Tanks, from 48,800 to 44,800 gallons. In

addition, footnote ** is deleted from TS 3.8.1.1.b.2. The proposed

change would also add an Action Statement to address remedial action

when a fuel oil transfer pump becomes inoperable.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

consequences of an accident previously evaluated.

TANK LEVEL

Amendment 59 provides an allowance for transferring fuel oil

from a pair of storage tanks associated with an inoperable

[Emergency Diesel Generator] EDG to another pair of storage tanks in

order to demonstrate compliance with PSE&G's commitment to

Regulatory Guide 1.137. The proposed change is consistent with that

transfer strategy and extends this allowance to include using fuel

oil in operable EDG storage tanks in order to reduce the amount of

stored fuel oil. Transfer from operable EDG storage tanks is,

actually, less complex than transferring from an inoperable EDG

storage tank since power to the transfer pumps would be available.

The low level alarm setpoint is the only physical change to be

made. No change is being made to the EDGs, to the fuel oil storage

tanks, or to the fuel oil transfer system and since EDG fuel oil

supply is associated with mitigating the consequences of an

accident, there is no change in the probability of any accident

analyzed in the [Updated Final Safety Analysis Report] UFSAR.

Since the proposed change still ensures the minimum fuel oil

storage capacity meets the existing licensing basis and since off-

site replacement oil is expected to be available within 60 hours

there is no change in the consequences of an accident previously

evaluated.

TRANSFER PUMP ACTION STATEMENT

Since no change is being made to the EDGs, to the fuel oil

storage tanks or to the fuel oil transfer system, and since EDG fuel

oil supply is associated with mitigating the consequences of an

accident, there is no change in the probability of any accident

analyzed in the UFSAR.

The proposed change provides compensatory action in the event a

single fuel oil transfer pump is inoperable without having to

immediately declare the EDG inoperable. The change ensures the

affected EDG remains fully capable of functioning as assumed in the

safety analyses, therefore, there is no significant impact on the

consequences of an accident previously evaluated.

Therefore, the proposed changes will not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

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

accident from any previously evaluated.

TANK LEVEL AND TRANSFER PUMP ACTION STATEMENT

The proposed changes will result in a setpoint change to the low

level alarm. No other physical changes to the EDGs, to the fuel oil

storage tanks, or to the fuel oil transfer system will result from

the proposed changes. Operation including the proposed changes will

not impair the diesel generators from performing as provided in the

design basis. In addition, EDG fuel oil supply is associated with

mitigating accident consequences, not accident prevention.

Therefore, the proposed change will not create the possibility of a

new or different kind of accident from any previously evaluated.

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

TANK LEVEL

The margin of safety is provided by the on-site storage of an

adequate supply of diesel fuel oil to ensure uninterrupted EDG

operation for seven days. Although the proposed change may result in

a reduction of stored fuel oil, the new minimum continues to provide

for an on-site seven day supply of diesel fuel oil.

TRANSFER PUMP ACTION STATEMENT

The margin of safety is provided by the ability of the fuel oil

transfer pumps to supply an adequate flow of the stored fuel to each

EDG day tank. The proposed change continues to provide 100% capacity

to the EDG day tank for a minimum of three days with no operator

action. With the proposed action, adequate transfer capability is

[[Page 34898]]

provided for a minimum of seven days fuel oil supply at which time

refilling of the tanks would provide an indefinite supply. With both

transfer pumps on a single EDG inoperable, the remaining three EDGs

would provide adequate power for safe shutdown. Transfer of fuel oil

from the storage tanks with inoperable transfer pumps can still be

effected using temporary hoses.

Since the proposed changes do not involve the addition of plant

equipment, are consistent with the intent of the existing Technical

Specifications, are consistent with allowances for fuel oil

transfers approved in Amendment 59, meets the intent of Regulatory

Guide 1.137, and are consistent with the design basis of the diesel

generators and the accident analysis, no action proposed by this

request will occur that will 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: Pennsville Public Library, 190

S. Broadway, Pennsville, New Jersey 08070

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

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

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: May 10, 1996

Description of amendment request: The proposed amendments would

change Technical Specification Sections, 1.0, 2.0, 3/4 1.0, 3/4 2.0,

5.0 and 6.0. These changes support the Margin Recovery Program (MRP)

and support increased steam generator tube plugging, improved fuel

reliability, reduced fuel costs, longer fuel cycles, reduced spent fuel

storage, and enhanced reactor safety. These changes incorporate the

results of the revised safety analyses (margin recovery) and the

establishment of a Core Operating Limits Report.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The accidents potentially affected by the parameters and

assumptions associated with the MRP have been evaluated/ analyzed

and all design standards and applicable safety criteria are met. The

consideration of these changes does not result in a situation where

the design, material, or construction standards that were applicable

prior to the change have been altered. Therefore, the changes

occurring with the MRP will not result in any additional challenges

to plant equipment that could increase the probability of any

previously evaluated accident.

The changes associated with the MRP do not affect plant systems

such that their function in the control of radiological consequences

is adversely affected. The safety evaluation documents that the

design standards and applicable safety criteria limits continue to

be met and therefore fission barrier integrity is not challenged.

The MRP changes have been shown not to adversely affect the response

of the plant to postulated accident scenarios. In all cases, the

calculated doses are within the regulatory criteria and therefore do

not constitute an increase in consequences. These changes will,

therefore, not affect the mitigation of the radiological

consequences of any accident described in the Updated Final Safety

Analysis Report (UFSAR).

Based on the above, it is concluded that the probability or

consequences of an accident previously evaluated is not

significantly 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.

The possibility for a new or difference[t] type of accident from

any accident previously evaluated is not created since the changes

associated with the MRP do not result in a change to the design

basis of any plant component or system. The evaluation of the

effects of the MRP changes shows that all design standards and

applicable safety criteria limits are met. These changes therefore

do not cause the initiation of a new accident nor create any new

failure mechanisms. Component integrity is not challenged. The

changes do not result in any event previously deemed incredible

being made credible. The MRP changes will not result in more adverse

conditions and will not result in any increase in the challenges to

safety systems.

Therefore, the consideration of the MRP as described in the

safety evaluation does not create the possibility of a new or

different type of accident from any accident previously evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The margin of safety is maintained by assuring compliance with

acceptance limits reviewed and approved by the NRC. Since all of the

appropriate acceptance criteria for the various analyses and

evaluations have been met, by definition there has not been a

reduction in any margin of safety.

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

Salem Unit 1 and 2 Technical Specifications has not been

significantly reduced.

Based on the above, PSE&G has determined that the proposed

changes do 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: 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

South Carolina Electric & Gas Company (SCE&G), South Carolina

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

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

Date of amendment request: April 16, 1996

Description of amendment request: The proposed amendment would

revise the Virgil C. Summer Nuclear Station, Unit 1 (VCSNS), Technical

Specifications (TS) to implement the amended regulation to 10 CFR Part

50, Appendix J, Option B (new rule), to provide a performance-based

option for leakage-rate testing of containment. The proposed amendment

will revise the VCSNS TS 3/4.6 ``Containment Systems,'' TS Bases 3/4.6,

and TS 6.8 ``Administrative Controls - Programs and Procedures,'' to

adopt the implementation requirements of 10 CFR Part 50, Appendix J,

Option B. The proposed amendment utilizes the guidelines (guidelines)

provided in ``Option B'' of Regulatory Guide (RG) 1.163 ``Performance-

Based Containment Leak-Test Program, September 1995,'' and NEI 94-01,

``Industry Guideline for Implementing Performance-Based Option of 10

CFR 50, Appendix J, July 26, 1995.'' The licensee has stated that the

proposed amendment is within these prescribed guidelines and does not

propose any deviations to the established methods which would impact

already approved analyses/justifications and established review

process.

The proposed change will remove the prescriptive TS requirements

for the performance of containment leakage testing and allow leakage

testing to be conducted as determined appropriate through the

performance-based or risk-based alternatives described in the VCSNS

Containment Leakage Rate Testing Program developed in accordance with

RG 1.163 and NEI 94-01. Since the requirements of Appendix J to 10 CFR

Part 50 will continue to

[[Page 34899]]

apply, the type of testing will not change. The proposed request does

not modify any plant equipment or systems.

The requirements of Appendix J will continue to govern the type of

test, testing methodology, and acceptance criteria for Type A, B, and C

testing. The performance-based testing of Option B eliminates or

modifies prescriptive regulatory requirements for which the burden is

marginal to safety for which the reviews and analyses have been

presented in NUREG-1493, ``Performance-Based Containment Leak-Test

Program, Final Report, September 1995.''

Earlier leakage testing performed at VCSNS has demonstrated low

overall containment leakage and supports the implementation of Option

B.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The probability or consequences of an accident previously

evaluated is not significantly increased.

There is no increase in the probability of an accident since

there is no work that would affect containment integrity. The

testing of containment isolation valves (CIVs) and other containment

penetration sealing devices is not postulated as an accident

precursor or initiating event.

Type A testing is capable of determining the total leakage from

both local leakage paths and gross containment leakage paths. Our

Type B and C testing has consistently provided accurate leakage

rates for valves and penetrations.

Administrative controls govern maintenance and testing such that

there is very low probability that unacceptable maintenance or

alignments can occur. Prior to and following maintenance on CIVs and

penetrations, a local leak rate test (LLRT) is required to be

performed. As a result, Type A testing is not required to accurately

quantify the leakage through containment penetrations.

Any specific exemptions to the requirements of Appendix J will

require approval by the NRC before implementation.

Therefore, this proposed change does not involve a significant

increase in the possibility or consequences of an accident

previously evaluated.

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

different type than any previously evaluated is not created.

The proposed request does not involve any physical changes to

the plant, affect the operation of the plant, or change testing

methods or acceptance criteria. The history of containment testing

verifies that containment integrity has been maintained.

The frequency changes allowed by implementation of Option B will

not significantly decrease the level of confidence in the ability of

the reactor building to limit offsite doses to allowable values. No

accident or malfunction can be the result of the allowed changes to

test schedule or frequency.

Since the proposed request will not directly impact equipment,

procedures or operations, the changes will not create the

possibility of any new or different kind of accident from any

previously evaluated.

3. The margin of safety has not been significantly reduced.

The reason for performing containment leakage rate testing is to

assure that the leakage paths are identified, and that any accident

release will be restricted to those paths assumed in the safety

analysis. The purpose for the schedule is to assure that containment

integrity is verified on a periodic basis.

Implementation of Option B to provide flexibility in the

scheduled requirements does not mean that containment integrity will

be compromised. The historical leakage rate test results for VCSNS

and for the nuclear industry support extension of testing

frequencies and demonstrate that structural integrity has been

maintained.

Therefore, the margin of safety has not been significantly

reduced.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Fairfield County Library, 300

Washington Street, Winnsboro, SC 29180

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

Gas Company, Post Office Box 764, Columbia, SC 29218

NRC Project Director: Eugene V. Imbro

Southern Nuclear Operating Company, Inc., Docket No. 50-364, Joseph

M. Farley Nuclear Plant, Unit 2, Houston County, Alabama

Date of amendment request: April 22, 1996

Description of amendment request: The amendment would revise the

Technical Specifications to implement the L* Tubesheet Region Plugging

Criterion, which would allow a steam generator tube to remain in

service with bands of axial degradation in the tubesheet region

provided sufficient non-degraded tubing remains to satisfy regulatory

guidance concerning structural and leakage integrity.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Operation of the Farley Nuclear Plant Unit 2 steam generators

in accordance with the proposed license amendment does not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

The supporting technical evaluations of the subject criteria

demonstrate that the presence of the tubesheet enhances the tube

integrity in the region of the hardroll by precluding tube

deformation beyond its initial expanded outside diameter. The

resistance to both tube rupture and tube collapse is strengthened by

the presence of the tubesheet in that region. The result of the

hardroll of the tube into the tubesheet is an interference fit

between the tube and the tubesheet. Tube rupture [cannot] occur

because the contact between the tube and tubesheet does not permit

sufficient movement of tube material. In a similar manner, the

tubesheet does not permit sufficient movement of tube material to

permit buckling collapse of the tube during postulated LOCA [loss-

of-coolant accident] loadings.

The type of degradation for which the L* criterion has been

developed (cracking with an axial or near axial orientation) has

been found not to significantly reduce the axial strength of a tube.

An evaluation including analysis and testing has been done to

determine the strength reduction for axial loads with simulated

axial and near axial cracks. This evaluation provides the basis for

the acceptance criteria for tube degradation subject to the L*

criterion.

The SRE [sound roll expansion] L* length is sufficient to

preclude significant leakage from tube degradation located below the

L* length. The existing Technical Specification leak rate

requirements and accident analysis assumptions remain unchanged in

the unlikely event that significant leakage from this region does

occur. Any leakage from the tube within the tubesheet at any

elevation in the tubesheet is fully bounded by the existing steam

generator tube rupture analysis included in the Farley Nuclear Plant

Final Safety Analysis Report. A conservative leakage allowance for

each L* tube is provided to determine the impact of L* criterion

upon offsite doses in the event of a postulated double ended

guillotine break of the main steam line outside of containment, but

upstream of the main steam line isolation valves. Since Farley Unit

2 has implemented the Interim Plugging Criteria (IPC) for ODSCC at

the tube support plates, projected steam line break (SLB) leakage at

the end of the next successive operating cycle must be evaluated.

Per Generic Letter 95-05, plants implementing the IPC can utilize

SLB leakage limits higher than the originally assumed 1.0 gpm

primary to secondary leakage value provided an analysis of offsite

doses consistent with the Standard Review Plan methodology is

performed. This analysis performed for the Farley Unit 2 plant

indicates that primary to secondary leakage of 11.2 gpm in the

faulted loop (0.1 gpm in the intact loops) will result in offsite

doses at the site boundary of less than 10% of the 10 CFR [Part] 100

guidelines. The total projected SLB leakage from all leakage sources

must remain below this value. Per attachment 4 addressing the L*

methodology,

[[Page 34900]]

the number of tube ends to which L* criterion can be applied is

limited to 600 per steam generator. Using a bounding SLB leakage

allowance per L* tube, the SLB leakage component from 600 L* tube

ends will be less than 0.33 gpm in the faulted loop. The proposed

alternate plugging criterion does not adversely impact any other

previously evaluated design basis accident. As the current Unit 2

IPC SLB leakage has been calculated to be less than 2 gpm in the

faulted loop, [an] SLB leakage margin of over 9 gpm is provided for

this cycle.

As noted above, tube rupture and pullout is not expected for

tubes using the L* criterion. In addition to the L* length, a

minimum length of SRE below the identified degradation must be

established. The aggregate L* distance of SRE provides the

structural integrity to prevent tube pullout. Conservatively, it is

assumed that the degraded band length does not provide any support

in resisting tube pullout.

Therefore SNC [Southern Nuclear Operating Company, Inc.]

concludes that Operation of the Farley Nuclear Plant Unit 2 steam

generators in accordance with the proposed license amendment does

not involve 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 kind of accident from any accident

previously evaluated.

Implementation of the proposed L* criterion does not introduce

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

criterion does not provide a mechanism to result in an accident

initiated outside of the region of the tubesheet expansion. The

structural integrity of L* tubes will be maintained during all plant

conditions. Any hypothetical accident as a result of any tube

degradation in the expanded portion of the tube would be bounded by

the existing tube rupture accident analysis. If it is postulated

that a circumferential separation of an L* tube were to occur below

the PLRL [pullout load reaction length], tube structural and leakage

integrity will be maintained during all plant conditions.

Verification of the L* distance of non-degraded tube roll expansion

prevents the postulated separated tube from lifting out of the

tubesheet during all plant conditions. Verification of the L*

criterion prevents tube displacement of any magnitude, and

therefore, postulated axial cracks existing a minimum of 0.5 inch

from either the bottom of the roll transition or top of tubesheet,

whichever is lower, from migrating out of the tubesheet.

Therefore, SNC concludes that the proposed license 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 a margin of safety.

The use of the L* criterion has been concluded to maintain the

integrity of the tube bundle commensurate with the requirements of

draft Regulatory Guide 1.121 under normal and postulated accident

conditions. The safety factors used in the verification of the

strength of the degraded tube are consistent with the safety factors

in the ASME [American Society of Mechanical Engineers] Boiler and

Pressure Vessel Code used in steam generator design. The L* length

has been verified by testing to be greater than the length of roll

expansion required to preclude significant leakage during normal and

postulated accident conditions. The leak testing acceptance criteria

are based on the primary to secondary leakage limit in the Technical

Specifications and the leakage assumptions used in the FSAR accident

analyses. The L* distance provides for structural integrity during

all plant conditions.

Implementation of the L* criterion will decrease the number of

tubes which must be taken out of service with tube plugs or repaired

with sleeves. Both plugs and sleeves reduce the RCS [reactor coolant

system] flow margin, thus implementation of the L* criterion will

maintain the margin of flow that would otherwise be reduced in the

event of increased plugging or sleeving.

Therefore, SNC, concludes based on the above, it is concluded

that the proposed change does not result in a significant reduction

in a loss of margin with respect to plant safety as defined in the

Final Safety Analysis Report [FSAR] or the bases of the FNP [Farley

Nuclear Plant] technical specifications.

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: Herbert N. Berkow

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

Callaway County, Missouri

Date of application request: May 29, 1996

Description of amendment request: The application requests staff

review and approval of a modification to the facility, as described in

the safety analysis report, that involves an unreviewed safety

question. The modification will reduce the single failure trip

potential for the main feedwater control and bypass valves.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The Callaway safety analysis assumes the MFC&BVs [main feedwater

control and bypass valves] close during certain events in order to

terminate fluid inventory addition to faulted steam generators and

thereby preclude the diversion of auxiliary feedwater to the main

feedwater system. This feature is necessary because each feedwater

line at Callaway is equipped with only one MFIV [main feedwater

isolation valve]. It should be noted that the safety analysis simply

requires the valves to close and does not prescribe a mechanism for

accomplishing that action.

The following are accidents that credit feedwater isolation or

AFW [auxiliary feedwater] addition. There is no impact by the

proposed modification on the consequences of each accident.

Feedwater System Malfunctions That Result In An Increase

In Feedwater Flow

Inadvertent Opening Of A Steam Generator Relief or

Safety Valve

Steam System Piping Failure

Loss of Nonemergency AC Power to the Station Auxiliaries

Loss of Normal Feedwater Flow

Feedwater System Pipe Break

Decrease in Reactor Coolant Inventory

The modification will not change the radiological consequences

of FSAR [final safety analysis report] Chapter 15 accidents because

the feedwater isolation function (and NSSS [nuclear steam supply

system] break response) has not changed. Therefore, there will be no

increase in the consequences of an accident evaluated previously in

the FSAR.

An analysis was performed to quantify the impact of the proposed

modification on the probability of MFCV [main feedwater control

valve] failure (closure) during normal plant operation. Comparison

of this failure probability for the existing design (1.20E-1 per

year) versus the proposed design (6.99E-2 per year) indicates that

the percentage reduction in the system failure probability at power

is 41.75%. Thus, the proposed design results in a reduction in the

probability of inadvertent MFCV failures at power and hence, a

reduction in the probability of a reactor trip and subsequent

challenges to other safety systems.

While this modification reduces the probability of a reactor

trip, it slightly increases the unavailability of the feedwater

isolation function. This is because the original design required

actuation of only one FWIS [feedwater isolation system] train to

close the MFC&BVs, whereas the new design requires actuation of both

trains. The impact of the modification on the probability of

incurring a feedwater isolation failure was therefore quantified,

utilizing PRA [probabilistic risk assessment] techniques. Fault

trees were developed for both the new and existing designs. Failure

probabilities for each event were then obtained from the IPE

[individual plant examination] and utilized to calculate failure

probabilities for the feedwater isolation safety function. This

calculation considered hardware failures

[[Page 34901]]

only, i.e., failure of an MFIV to close after receiving an actuation

signal. The failure probability of feedwater isolation, based on the

proposed design, was determined to be 6.1E-5 per demand (1 event

every 16,400 demands). The existing design was found to have a

failure probability of 2.8E-5 per demand (1 event every 35,700

demands). Therefore, this modification will not significantly

increase the probability or consequences of an accident evaluated

previously in the FSAR.

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

or different kind of accident from any accident previously

evaluated.

The modification maintains the present de-energize-to-actuate

configuration of the MFC&BV trip solenoid valves.

Thus, the proposed modification does not create the possibility

of an accident of a different type than any previously evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

Credit is taken in the accident analyses for the MFIVs to close

on demand for feedwater isolation. Because of this, the MFIVs have

been incorporated into the Callaway Technical Specifications. Action

Statements and surveillance requirements have been developed to

assure the availability of the valves when needed.

The MFC&BVs are not addressed by any of the Callaway Technical

Specifications or their bases. Therefore, this modification will not

involve a significant reduction in the margin of safety as defined

in the basis for any technical specification.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Callaway County Public

Library, 710 Court Street, Fulton, Missouri 65251

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

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

NRC Project Director: William H. Bateman

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

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

Creeks, Manitowoc County, Wisconsin

Date of amendment request: May 29, 1996

Description of amendment request: The proposed amendment would

revise Technical Specification (TS) Section 15.4.4, ``Containment

Tests,'' to incorporate the provisions of 10 CFR Part 50, Appendix J,

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

Reactors,'' Option B. Revisions would also be made to TS Sections 15.1,

``Definitions,'' 15.3.6, ``Containment System,'' and 15.6,

``Administrative Controls,'' to support the proposed changes to Section

15.4.4.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration which is presented below:

1. Operation of this facility under the proposed Technical

Specifications will not create a significant increase in the

probability or consequences of an accident previously evaluated.

The proposed change does not involve a change to structures,

systems, or components which would affect the probability or

consequences of an accident previously evaluated in the PBNP [Point

Beach Nuclear Plant] Final Safety Analyses Report (FSAR).

Furthermore, containment leakage rate testing is not an initiator of

any accident. The proposed change simply provides a mechanism within

the Technical Specifications for implementing a performance-based

method of determining the frequency for leakage rate testing which

has been approved by the NRC. The proposed change does not affect

reactor operations or accident analysis and has no significant

radiological consequences. Therefore, this change will not create a

significant increase in the probability or consequences of an

accident previously evaluated.

2. Operation of this facility under the proposed Technical

Specifications change will not create the possibility of a new or

different kind of accident from any accident previously evaluated.

The proposed change does not involve a change to the plant

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

affect any of the parameters or conditions that contribute to

initiation of any accidents. This change involves a potential

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

of defining the test frequency, the methods for performing the

actual tests are not changed. No new accident modes are created by

extending the testing intervals. No safety-related equipment or

safety functions are altered as a result of this change. Extending

the test frequency has no influence on, nor does it contribute to,

the possibility of a new or different kind of accident or

malfunction from those previously analyzed. Therefore, the proposed

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

accident from any accident previously evaluated.

3. Operation of this facility under the proposed Technical

Specifications change will not create a significant reduction in a

margin of safety.

The proposed change potentially affects only the frequency of

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

frequency, the methods for performing the actual tests are not

changed. The proposed change is based on NRC accepted provisions and

maintains necessary levels of system and component reliability

affecting containment integrity. Evaluation of the performance-based

approach to leakage rate testing, as documented in NUREG-1493,

concludes that the impact on public health and safety due to revised

testing intervals is negligible. Furthermore, the proposed change

will not reduce the availability of systems associated with

containment integrity when they are required to mitigate accident

conditions. Therefore, the proposed change will not create 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: Joseph P. Mann Library, 1516

Sixteenth Street, Two Rivers, Wisconsin 54241

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

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

NRC Project Director: Gail H. Marcus

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

Nuclear Power Plant, Kewaunee County, Wisconsin

Date of amendment request: June 4, 1996

Description of amendment request: The proposed amendment would

revise the Kewaunee Nuclear Power Plant Technical Specifications (TS)

by reducing the surveillance test frequencies for the radiation

monitoring system (Table TS 4.1-1) and the control rods (Table TS 4.1-

3) in accordance with the guidance of Generic Letter 93-05, ``Line-Item

Technical Specifications Improvements to Reduce Surveillance

Requirements for Testing During Power Operation,'' dated September 27,

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:

Table TS 4.1-1, ``Minimum Frequencies for Checks, Calibrations

and Test of Instrument Channels,'' Item 19

The proposed changes were reviewed in accordance with the

provisions of 10 CFR 50.92 to determine that no significant hazards

exist. The proposed changes will not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The radiation monitors are not accident initiators; therefore,

they cannot increase the probability of an accident occurring. The

reliability of the radiation monitors is not expected to decrease

due to the decreased surveillance frequency; therefore, this change

does not increase the consequences of an accident.

[[Page 34902]]

The addition of comment (a) to the Check, Calibrate, and Test

columns is merely a clarification of the existing information in the

table and does not change the intent of the Technical

Specifications.

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

from any accident previously evaluated.

The proposed change revises only the testing frequency and does

not revise the test method or operational performance of the

radiation monitors. The radiation monitors are not accident

initiators; therefore, they cannot create a new or different kind of

accident.

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

Quarterly testing of the radiation monitoring system channels

will continue to verify operability of the monitors. Decreasing the

test surveillance frequency is not expected to decrease the

reliability of the radiation monitors. This change is acceptable in

accordance with Generic Letter 93-05 and NUREG-1366, ``Improvements

to Technical Specifications Surveillance Requirements.''

Table TS 4.1-3, ``Minimum Frequencies for Equipment Tests,''

Item 1

The proposed change in test frequency for control rod exercising

was reviewed in accordance with the provisions of 10 CFR 50.92 to

determine that no significant hazards exist. It has been determined

that the proposed change will not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed change revises only the testing frequency for

control rod exercising. The control rod exercise surveillance

procedure will continue to be conducted, on a quarterly basis, to

ensure that the equipment remains operable. The reduced frequency of

control rod exercising reduces the probability of an inadvertent

reactor trip occurring during testing due to a dropped control rod.

Surveillance procedure SP 49-075 is conducted to verify rod

movement. In accordance with NUREG-1366, the frequency of a stuck

control rod occurring is very low. This condition is most often

discovered during reactor startup or during low power physics

testing. The reduction in control rod exercising is, therefore,

considered acceptable and is not expected to affect the probability

of a stuck control rod occurring.

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

from any accident previously evaluated.

The proposed change revises only the testing frequency and does

not revise the test method or the design of the control rod system.

Therefore, a new or different kind of accident will not be created

by this change.

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

Quarterly control rod exercising will continue to verify

movement of the control rods. No adverse consequences are expected

to occur due to decreasing the test frequency. This change is

acceptable in accordance with Generic Letter 93-05 and NUREG-1366.

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,

Cofrin Library, 2420 Nicolet Drive, Green Bay, Wisconsin 54311-7001

Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner,

P. O. Box 1497, Madison, Wisconsin 53701-1497

NRC Project Director: Gail H. Marcus

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

Nuclear Power Plant, Kewaunee County, Wisconsin

Date of amendment request: June 10, 1996

Description of amendment request: The proposed amendment would

revise Technical Specification 4.2.b, ``Steam Generator Tubes,'' and

its associated basis, by allowing the use of Westinghouse laser-welded

sleeves to repair defective steam generator tubes. A description of the

sleeving repair process and supporting technical justification are

contained in WCAP-13088, Revision 3, ``Westinghouse Series 44 and 51

Steam Generator Generic Sleeving Report.'' WCAP-13088, and a non-

proprietary version (WCAP-13089), were submitted to the Nuclear

Regulatory Commission on April 13, 1995, to support a similar TS

amendment request for the DC Cook Nuclear Power Plant, Unit 1.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Operation of the KNPP [Kewaunee Nuclear Power Plant] in

accordance with the proposed license amendment does not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

The LWS [laser-welded sleeve] configuration has been designed

and analyzed in accordance with the requirements of the ASME

[American Society of Mechanical Engineers] Code. Fatigue and stress

analyses of the sleeved tube assemblies produced acceptable results;

i.e., the applied stresses and fatigue usage for the sleeve and weld

are bounded by the limits established in the ASME Code. ASME Code

minimum material property values are used for the structural and

plugging limit analysis. Ultrasonic inspection is used to verify

that minimum weld fusion zone thicknesses are produced. Mechanical

testing of 7/8'' tubesheet sleeves installed in roll expanded tubes

has shown that the individual joint structural strength of Alloy 690

LWSs provides margin to acceptance limits. These acceptance limits

bound the most limiting loadings (3 times normal operating pressure

differential) recommended by RG [Regulatory Guide] 1.121. Therefore,

each individual joint provides for structural integrity exceeding RG

recommendations. A hypothetical loss of integrity of one of the

joints will not result in a loss of structural integrity for the

sleeve. Leakage testing for 3/4'' and 7/8'' full length tubesheet

sleeves has demonstrated that unacceptable levels of primary-to-

secondary leakage are not expected during all plant conditions for

non-welded tubesheet sleeve lower joints. The welded joint produces

a hermetic seal, and therefore will not leak under any plant

conditions. Laser welded sleeves will not contribute to the current

SLB [steam-line break] primary-to-secondary leakage limit of 34 gpm

in the faulted loop. The 34 gpm leakage limit was calculated in

accordance with the standard review plan methodology to support

implementation of the voltage-based repair criteria for tube support

plate intersections.

The sleeve minimum acceptable wall thickness (used for

developing the depth based plugging limit for the sleeve) is

determined using the guidance of RG 1.121 and the pressure stress

equation of Section III of the ASME Code. With respect to the design

of the sleeve for KNPP, the limiting requirement of the RG which

applies to part throughwall degradation is that the minimum

acceptable wall must maintain a factor of safety consistent with the

analysis conditions as defined by the ASME Code. A bounding set of

design and transient loading input conditions was used for the

minimum wall thickness evaluation in the generic evaluation.

Evaluation of the minimum acceptable wall thickness for normal,

upset and postulated accident condition loading per the ASME Code

indicates the limiting condition is established for the normal

operating conditions, and the minimum acceptable wall thickness for

this case bounds the upset and faulted condition values.

According to RG recommendations, an allowance for non-

destructive evaluation (NDE) uncertainty and operational growth of

existing tube wall degradation indications within the sleeve must be

accounted for when determining the sleeve plugging limit. A

conservative tube wall degradation growth rate per cycle and an NDE

uncertainty has been assumed for determining the sleeve TS plugging

limit. The sleeve wall degradation extent determined by NDE, which

would require plugging sleeved tubes, is developed using the

guidance of RG 1.121 and is defined in WCAP-13088 [non-proprietary

WCAP-13089] to be 25% throughwall (plugging limit = 100% -

structural limit + NDE uncertainty + growth) for KNPP.

The hypothetical consequences of failure of the sleeve joint

would be bounded by the current SG [steam generator] tube rupture

analysis included in the KNPP Updated Safety Analysis Report. Due to

the slight reduction in diameter caused by the sleeve wall

thickness, primary coolant release rates would be slightly less than

assumed for the SG tube rupture analysis (depending on break

location), and therefore, would result

[[Page 34903]]

in lower total primary fluid mass release to the secondary system.

The proposed TS change to use Alloy 690 LWSs does not adversely

impact any other previously evaluated design basis accidents or the

results of LOCA [loss of coolant accident] and non-LOCA accident

analyses for the current TS minimum reactor coolant system flow

rate. The results of the analyses and testing, as well as plant

operating experience, demonstrates that the sleeve assembly is an

acceptable means of maintaining tube integrity. Plugging limit

criteria are established using the guidance of RG 1.121.

Furthermore, per RG 1.83 recommendations, the sleeved tube will be

monitored through periodic inspections with present NDE techniques.

These measures demonstrate that installation of sleeves spanning

degraded areas of the tube will restore the tube to a condition

consistent with its original design basis.

Corrosion testing of free span LWS joint has indicated that the

corrosion resistance (relative to roll transitions) can be increased

by greater than a factor of ten with the application of a PWHT [post

weld heat treatment] step. Estimations of joint susceptibility based

on expected far field stresses after heat treatment using the

expected original tube-to-tubesheet hydraulic expansion transition

residual stresses and actual time to crack in these transitions at

KNPP indicate that LWS joint lifetime should exceed the current

plant license. Consistent with other license amendments addressing

LWS, all free span laser welds will receive a PWHT; therefore, rapid

corrosion degradation of the free span joint is not expected.

Recently performed corrosion testing of LWS joints in locked tube

conditions indicates that with PWHT the stress corrosion cracking

resistance and initiation potential in the parent tube weld region

is greatly enhanced. Similar test results and conclusions would be

expected for KNPP. The Model 51 SG tube span between the top of the

tubesheet and the first support plate is such that even lower PWHT

residual stresses would be expected. Also, the weld placement within

the hydraulically expanded area and sleeve installation sequence

have been optimized to provide for some level of heat treatment at

the upper transition above the weld and lower far field residual

stress levels. While no parent tube degradation has been detected at

this elevation, or any other elevation in a laser welded sleeve

assembly, the relocation of the weld serves to provide further

resistance to PWSCC [primary water stress corrosion cracking] at

this elevation. The suggested target PWHT temperature has also been

optimized in that this temperature provides for adequate PWHT while

maintaining the parent tube far field stresses.

Approximately 19,500 LWSs have been installed in the U.S. Of

this number, ove

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