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

NUCLEAR REGULATORY COMMISSION

Biweekly Notice; Applications and Amendments to Facility

Operating Licenses Involving No Significant Hazards Considerations

I. Background

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

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

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

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

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

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

the Commission the authority to issue and make immediately effective

any amendment to an operating license upon a determination by the

Commission that such amendment involves no significant hazards

consideration, notwithstanding the pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all notices of amendments issued, or

proposed to be issued from July 20, 1998, through July 31, 1998. The

last biweekly notice was published on July 29, 1998 (63 FR 40551).

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 and

Directives Branch, Division of Administration 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 September 11, 1998, 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

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petition; and the Secretary or the designated Atomic Safety and

Licensing Board will issue a notice of a hearing or an appropriate

order.

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

shall set forth with particularity the interest of the petitioner in

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

the proceeding. The petition should specifically explain the reasons

why intervention should be permitted with particular reference to the

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

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

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

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

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

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

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

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

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

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

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

the specificity requirements described above.

Not later than 15 days prior to the first prehearing conference

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

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

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

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

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

brief explanation of the bases of the contention and a concise

statement of the alleged facts or expert opinion which support the

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

contention at the hearing. The petitioner must also provide references

to those specific sources and documents of which the petitioner is

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

facts or expert opinion. Petitioner must provide sufficient information

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

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

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

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

petitioner who fails to file such a supplement which satisfies these

requirements with respect to at least one contention will not be

permitted to participate as a party.

Those permitted to intervene become parties to the proceeding,

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

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

hearing, including the opportunity to present evidence and cross-

examine witnesses.

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

determination on the issue of no significant hazards consideration. The

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

If the final determination is that the amendment request involves

no significant hazards consideration, the Commission may issue the

amendment and make it immediately effective, notwithstanding the

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

of the amendment.

If the final determination is that the amendment request involves a

significant hazards consideration, any hearing held would take place

before the issuance of any amendment.

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

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

Commission, Washington, DC 20555-0001, Attention: Rulemakings and

Adjudications Staff, or may be delivered to the Commission's Public

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

by the above date. A copy of the petition should also be sent to the

Office of the General Counsel, U.S. Nuclear Regulatory Commission,

Washington, DC 20555-0001, and to the attorney for the licensee.

Nontimely filings of petitions for leave to intervene, amended

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

not be entertained absent a determination by the Commission, the

presiding officer or the Atomic Safety and Licensing Board that the

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

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

For further details with respect to this action, see the

application for amendment which is available for public inspection at

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

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

the particular facility involved.

Baltimore Gas and Electric Company, Docket No. 50-318, Calvert Cliffs

Nuclear Power Plant, Unit No. 2, Calvert County, Maryland

Date of amendment request: July 20, 1998.

Description of amendment request: Baltimore Gas and Electric

Company (BGE) request a modification involving replacing the service

water (SRW) heat exchangers with new plate and frame heat exchangers

having increased thermal performance capability. A similar license

amendment dated February 8, 1998, was granted to Operating License No.

DPR-53--Calvert Cliffs Nuclear Power Plant, Unit 1.

The planned modification for Unit 2 is virtually identical to the

one just completed for Unit 1 during the spring 1998 refueling outage.

The only exception is the addition of an extra manual valve in the Unit

2 system to isolate the bypass line for maintenance. This additional

manual valve is needed due to the change in location of the tie-in to

the main header. (The Unit 1 bypass line ties into the main header

downstream of a control valve; therefore, it did not need a separate

isolation valve for maintenance.)

The saltwater and SRW piping configuration will be modified as

necessary to allow proper fit-up to the new components. A flow control

scheme to throttle saltwater flow to the heat exchangers and the

associated bypass lines will be added. Saltwater strainers with an

automatic flushing arrangement will be added upstream of each heat

exchanger. The majority of the physical work associated with this

modification is restricted to the SRW pump room.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Would not involve a significant increase in the probability

or consequences of an accident previously evaluated.

None of the systems associated with the proposed modification

are accident initiators. The SW and SRW Systems are used to mitigate

the effects of accidents analyzed in the UFSAR [Updated Final Safety

Analysis Report]. The SW and SRW Systems provide cooling to safety-

related equipment following an accident. They support accident

mitigation functions; therefore, the proposed modification does not

increase the probability of an accident previously evaluated.

The proposed modification will increase the heat removal

capacity of the SRW System. The design provided under this activity

ensures that the safety features provided by the SW and SRW are

maintained, and in some instances enhanced; i.e., the availability

of important-to-safety equipment required to mitigate the

radiological consequences of an accident described in the UFSAR is

enhanced by the

[[Page 43202]]

flexibility and increased thermal margin provided with this design.

The redundant cooling capacity of the SW and SRW Systems have

not been altered. Furthermore, the proposed activity will not

change, degrade, or prevent actions described or assumed in any

accident described in the UFSAR. The proposed activity will not

alter any assumptions previously made in evaluating the radiological

consequences of any accident described in the UFSAR. Therefore, the

consequences of an accident previously evaluated in the UFSAR have

not increased.

Therefore, the proposed modification does not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

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

of accident from any accident previously evaluated.

The proposed activity involves modifying the SW and SRW System

components necessary to support the installation of new SRW heat

exchangers. None of the systems associated with this modification

are identified as accident initiators in the UFSAR. The SW and SRW

Systems are used to mitigate the effects of accidents analyzed in

the UFSAR. None of the functions required of the SRW or SW System

have been changed by this modification. This activity does not

modify any system, structure, or component such that it could become

accident initiator, as opposed to its current role as an accident

mitigator.

Therefore, the proposed change does not create the possibility

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

evaluated.

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

safety.

The Safety design basis for the SW and SRW System is the

availability of sufficient cooling capacity to ensure continued

operation of equipment during normal and accident conditions. The

redundant cooling capacity of these systems, assuming a single

failure, is consistent with assumptions used in the accident

analysis.

The design, procurement, installation, and testing of the

equipment associated with the proposed modification are consistent

with the applicable codes and standards governing the original

systems, structures, and components. The design of instruments and

associated cabling ensures that physical and electrical separation

of the two subsystems is maintained. Common-mode failure is not

introduced by the activity. The equipment is qualified for the

service conditions stipulated for that environment. New cable and

raceways for this design will be installed in accordance with

seismic design requirements. The additional electrical load has been

reviewed to ensure the load limits for the vital 1E buses are not

exceeded. The circuits and components related to the control valves

control loops are safety-related, are similar to those used for the

other safety-related flow control functions. The proposed

modification will not have any adverse effects on the safety-related

functions of the SW and SRW Systems.

For the above reasons, the existing licensing bases have not

been altered by the proposed modification. This activity will not

reduce the margin of safety as it exists now. In fact, the margin of

safety has been increased by this activity due to the increase in

the thermal capacity of the dual train design (i.e., two heat

exchangers per train versus one heat exchanger per train of the

original design) and the increased availability of safety-related

components.

Therefore, this proposed modification does not significantly

reduce the margin of safety.

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

this review, it appears that the three standards of 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: Calvert County Library, Prince

Frederick, Maryland 20678.

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

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

NRC Project Director: S. Singh Bajwa, Director.

Duquesne Light Company, et al., Docket Nos. 50-334 and 50-412, Beaver

Valley Power Station, Unit Nos. 1 and 2, Shippingport, Pennsylvania

Date of amendment request: July 13, 1998.

Description of amendment request: The proposed amendments would

revise the Beaver Valley Power Station, Unit Nos. 1 and 2 (BVPS-1 and

BVPS-2) Updated Final Safety Analysis Report (UFSAR) descriptions of

the Intake Structure main entrance and interconnecting cubicle doors.

The current UFSAR descriptions state that the cubicle access doors are

open to permit excess water from a major pipe rupture to flow out of

the cubicles thereby avoiding internal flooding. The proposed changes

would address a new failure mode of safety-related equipment that had

not been previously considered for BVPS-1. The proposed changes would

state that the cubicle interconnecting flood protection doors are

normally closed with their inflatable seals depressurized and that the

associated security/fire doors are normally closed. The proposed door

closure arrangement is intended to protect the safety-related equipment

in the interconnecting cubicles from the consequences of potential

internal flooding.

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?

The proposed change revises the text of the UFSAR for Unit 1 and

Unit 2 to describe how protection is provided against potential

internal floods in the cubicles that house the Unit 1 River Water

and Unit 2 Service Water Pumps. The previous description concluded

that the Unit 1 River Water pumps were protected because open

cubicle access doors will permit excess water to flow out of the

cubicles. The practice that has changed, and is described in the

proposed revisions to the Unit 1 and Unit 2 UFSARs, will provide

protection of the Unit 1 River Water Pumps and the Unit 2 Service

Water Pumps so that no flooding event can adversely affect more than

one Unit 1 or Unit 2 pump. Therefore, it can be concluded that the

proposed changes do not involve any increase in the probability or

consequences of an accident previously evaluated.

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

kind of accident from any accident previously evaluated?

The effect of flooding the pump cubicles was considered in BVPS-

1 to have no adverse effect because open cubicle access doors would

permit excess water to flow out of the cubicles, and pipe cracks in

moderate energy piping was not part of the design basis. Revising

the door arrangement described in the BVPS-1 UFSAR such that the

security/fire doors are normally closed, requires that the effects

of flooding be considered. Engineering analysis shows that a

moderate energy pipe crack, (i.e., the BVPS-2 design basis internal

flood), produces a leak rate of 1162 gpm, which results in a maximum

water level of 0.82 feet, with the security/fire doors closed. The

water level in the adjacent cubicle would reach a level at 0.37

feet. This is below the level which would cause failures of the MCCs

[Motor Control Centers] in the pump cubicles.

The maximum leak rate from a failure of a Unit 1 rubber

expansion joint in a pump cubicle would result in water rising to a

level which would cause the MCCs to be flooded and fail; therefore,

maintaining the flood door between the adjacent cubicles closed

limits the impact to a single train.

Failure of a single train of River Water is analyzed in the

USAR; therefore, this change would not introduce a new or different

type of accident.

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

of safety?

The proposed change in the Unit 1 and Unit 2 UFSARs describes

how protection is provided for the Unit 1 River Water, and the Unit

2 Service Water pumps. Protection of the Unit 1 River Water Pumps

and the Unit 2 Service Water pumps is provided so that no flooding

event can adversely affect more than one Unit 1 or Unit 2 pump.

Therefore, it can be concluded that the proposed changes 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 10 CFR 50.92(c) are

[[Page 43203]]

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

amendment request involves no significant hazards consideration.

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

663 Franklin Avenue, Aliquippa, PA 15001.

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

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

NRC Project Director: Robert A. Capra.

Duquesne Light Company, et al., Docket Nos. 50-334 and 50-412, Beaver

Valley Power Station, Unit Nos. 1 and 2, Shippingport, Pennsylvania

Date of amendment request: July 9, 1998.

Description of amendment request: The proposed amendment would

revise Technical Specification (TS) 3/4.7.1.1 and associated Bases for

both units. TS 3.7.1.1 currently provides requirements for reducing the

power range high neutron flux trip setpoint when one or more main steam

safety valves are inoperable. The current basis for determining the

amount of trip setpoint reduction has been determined to be non-

conservative. The proposed amendment would specify maximum allowable

reactor power level based on the number of operable main steam safety

valves rather than requiring a reduction in reactor trip setpoint. This

change would be consistent with the NRC staff's guidance provided in

the NRC's improved Standard Technical Specifications for Westinghouse

plants (NUREG-1431, Revision 1). The maximum allowable reactor power

level with inoperable safety valves would be calculated based on the

recommendations of Westinghouse Nuclear Safety Advisory Letter (NSAL)

94-01. The proposed change to the Unit 1 TS 3.7.1.1 would also delete

reference to 2 loop operation since 2 loop operation is not a licensed

condition for either unit.

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?

The proposed change will generally incorporate the Improved

Standard Technical Specification (ISTS) main steam safety valve

(MSSV) requirements of NUREG-1431 into Specification 3.7.1.1 and

associated Bases. The Unit 1 specification currently includes

reference to 2 loop operating requirements in Action ``b'' and Table

3.7-2. Reference to 2 loop operation is being deleted since it is

not addressed in the ISTS and is not a licensed condition for these

plants. The limiting condition for operation has been modified to

incorporate the ISTS wording and requires MSSV operability in

accordance with Tables 3.7-1 and 3.7-2. Table 3.7-1 lists the

maximum allowable power level as a function of the number of

operable MSSVs per steam generator and continues to require a

minimum of 2 operable MSSVs per steam generator for continued plant

operation. Table 3.7-2 specifies the MSSV lift setting and tolerance

for each MSSV. The valve lift setting remains unchanged along with

the current tolerance of +1 percent -3 percent. The Applicability

statement has not been changed since it is consistent with the ISTS

requirements.

Proposed Action ``a'' applies with one or more inoperable MSSVs

and requires that within 4 hours power must be reduced in accordance

with the value specified in Table 3.7-1; otherwise, shut down. This

action satisfies the same goal as the current action by restricting

thermal power so that the energy transfer to the most limiting steam

generator is not greater than the available relief capacity for that

steam generator. Proposed Action ``b'' incorporates additional

conservatism by specifically requiring at least 2 operable MSSVs per

steam generator. This ensures that a minimum overpressure protection

is available during all applicable modes of operation. Proposed

Action ``c'' provides an exception to Specification 3.0.4 which does

not allow entry into a mode where the Limiting Condition for

Operation (LCO) is not met and actions require a shutdown. This

exception is not addressed in the ISTS requirements; however, an

exception to Specification 3.0.4 allows entry into a mode where the

LCO applies in conformance with the action statements.

Proposed Surveillance Requirement 4.7.1.1 requires verification

of the lift setpoint for each MSSV listed in Table 3.7-2 in

accordance with the Inservice Test Program. Note (1) is applied to

Surveillance Requirement 4.7.1.1 to provide clarification of the

testing requirements, such that this testing is required only in

Modes 1 and 2 so that the plant can enter Modes 2 and 3 where this

specification applies without first performing the test. A note (2)

has been applied to the lift setting in Table 3.7-2 that requires a

setting corresponding to the ambient conditions of the valve at the

nominal operating temperature and pressure. The ISTS does not

include this note but it has been included for consistency with the

current note and provides a clear reminder to test personnel of the

required test conditions.

The safety valve Bases have been revised to generally

incorporate the ISTS Bases which significantly improve the content

and understanding of the MSSV requirements. These changes are

consistent with the UFSAR [Updated Final Safety Analysis Report]

design description and analysis assumptions where the MSSVs provide

the required overpressure protection. The proposed changes are

consistent with the regulations and provide additional assurance

that the secondary side pressure remains within the bounds of the

safety analyses; therefore, the proposed changes will not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

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

kind of accident from any accident previously evaluated?

The proposed changes generally incorporate the ISTS MSSV

requirements to ensure adequate secondary side overpressure

protection is available and properly maintained. The revised

Limiting Condition for Operation (LCO) limits plant power level

based on the number of operable MSSVs as stated in Table 3.7-1 and

provides the valve lift settings and tolerances as shown in Table

3.7-2. The actions require a reduction in power when the number of

valves is less than the full complement for each steam generator and

also require at least 2 operable MSSVs per steam generator. When

these requirements cannot be met a plant shutdown is required. An

action also provides an exception to Specification 3.0.4 and is

consistent with the exception currently provided. These actions are

more conservative than the current requirements and provide

additional assurance that Specification 3.7.1.1 will continue to

govern the MSSV limitations in a manner consistent with the accident

analyses assumptions. The revised surveillance requirement provides

clearly understandable testing requirements to ensure the MSSVs are

adequately monitored and will perform in accordance with the

accident analysis assumptions. The proposed change does not

introduce any new mode of operation or require any physical

modification to the plant; therefore, this change will not create

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

accident previously evaluated.

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

of safety?

The MSSVs ensure the ASME [American Society of Mechanical

Engineers] Code, Section III requirements are maintained to limit

the secondary system pressure to within 110 percent of the design

pressure when passing the design steam flow. This ensures that the

overpressure protection system can cope with all operational and

transient events. Operation with less than the full number of MSSVs

is permitted as long as thermal power is restricted to meet the ASME

Code requirements. This limitation is provided in the proposed

technical specifications along with operability and surveillance

requirements to ensure the level of overpressure protection is

maintained. MSSV operability is defined as the ability to open

within the setpoint tolerances, relieve steam generator

overpressure, and reseat when pressure has been reduced. MSSV

operability is determined by surveillance testing in accordance with

the Inservice Test program which provides assurance that the MSSVs

will perform their designed safety functions to mitigate the

consequences of accidents that could result in a challenge to the

reactor coolant pressure boundary. The proposed change continues to

ensure that the required components are properly maintained and that

the assumed parameters are verified during the applicable conditions

[[Page 43204]]

and on a consistent basis; therefore, this change will not reduce

the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

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

663 Franklin Avenue, Aliquippa, PA 15001.

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

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

NRC Project Director: Robert A. Capra.

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

Generating Station, Ocean County, New Jersey

Date of amendment request: July 21, 1998.

Description of amendment request: The proposed change request would

permit an alternative to the requirement to perform Control Rod Drive

(CRD) scram time testing with the reactor pressurized prior to resuming

power operation. The change would permit: (1) scram time testing with

the reactor depressurized prior to resuming operation, and (2) a second

scram time test with the reactor pressure above 800 psig, prior to

exceeding 40% reactor power.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Involve a significant increase in the probability of

occurrence or consequences of an accident previously evaluated; (or)

There will not be an increase in the probability of occurrence

of an accident previously evaluated in the Safety Analysis Report

(SAR) because the requested change provides additional assurance

that the CRD System is able to perform its safety function, and

therefore does not change the probability of occurrence of an

accident.

There will not be an increase in the consequences of an accident

previously evaluated in the Safety Analysis Report (SAR) because the

requested change will ensure that the CRD System is able to perform

its safety function, and therefore does not change the consequences

of an accident.

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

from any accident previously evaluated; (or)

The requested change will not create the possibility of a new or

different kind of accident from any accident previously evaluated.

The first issue associated with the requested change is increased

wear on the CRDs, resulting in increased buffer seal wear or

failure. This wear or failure of the buffer seal would result in

difficulty or inability to withdraw the rod subsequent to the

depressurized scram. The safety function of the rod to insert on a

scram signal, however, would be unaffected by this seal degradation.

Therefore, there is no safety concern with the increased wear due to

performance of the cold scram test.

The other consideration associated with the new requested change

is the possible increased risk of stub tube leakage during the cold

(depressurized) test. Without the download due to reactor pressure,

the momentary upward loading on the CRD stub tube puts the stub tube

into tension. Any flaws in the stub tube could grow and eventually

result in a stub tube leak. The likelihood of flaws in the stub

tubes, however, is very small, based on the extensive repair work on

the stub tube surfaces performed prior to plant operation. The

integrity of the stub tube repairs is verified by the 1000 pound

leak test performed during every startup of the reactor. This test,

therefore, poses very minimal risk of stub tube leakage.

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

The change will not decrease the margin of safety as defined in

the basis of any Technical Specification. This is because the

requested change, like the existing Technical Specification test,

provides assurance that the CRD System is able to perform its safety

function, and therefore does not change 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: Ocean County Library,

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

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

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

NRC Project Director: Cecil O.Thomas.

GPU Nuclear, Inc., et al., Docket No. 50-289, Three Mile Island Nuclear

Station, Unit No. 1, Dauphin County, Pennsylvania

Date of amendment request: June 11, 1998

Description of amendment request: The proposed amendment would

incorporate an alternative high radiation area control for Three Mile

Island Nuclear Station, Unit No. 1 (TMI-1) in accordance with 10 CFR

20.1601(c). The alternative would modify Technical Specification 6.12

to allow for a conspicuously posted barricade and flashing light in

individual high radiation areas that are located within large areas

where no enclosure exists for locking, and no enclosure can be

reasonably erected. A minor clarification to indicate that the

requirement of paragraph 6.12.1.a also applies to 6.12.1.b and an

editorial change were added.

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

amendment would not involve a significant increase in the

probability or consequences of an accident previously evaluated. The

proposed amendment involves changes to the TMI-1 Technical

Specifications, which are consistent with Regulatory Guide 8.38.

This change does not involve any change to system or equipment

configuration. The proposed amendment incorporates an alternative

high radiation area control, which has been previously found to be

acceptable by the NRC. The reliability of systems and components

relied upon to prevent or mitigate the consequences of accidents

previous evaluated is not degraded by the proposed changes.

Therefore, this change does not increase the probability or

consequences of an accident previously evaluated.

2. Operation of the facility in accordance with the proposed

amendment would not create the possibility of a new or different

kind of accident from any previously evaluated. This change only

involves controls for access to high radiation areas. Access to

plant equipment during normal or accident conditions will not be

affected by utilizing this alternate method. Therefore, the proposed

amendment does 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

amendment would not involve a significant reduction in a margin of

safety. The proposed amendment is consistent with Regulatory Guide

8.38. The proposed amendment involves high radiation area access

control and is not related to the margin of safety associated with

any plant operation or transients. Therefore, it is concluded that

operation of the facility in accordance with the proposed amendment

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

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

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

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

amendment request involves no significant hazards consideration.

[[Page 43205]]

Local Public Document Room location: Law/Government Publications

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

Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105.

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

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

NRC Project Director: Cecil O. Thomas.

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

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

Date of amendment request: May 20, 1998.

Description of amendment request: The proposed change would revise

the Refueling Water Storage Tank (RWST) setpoint associated with

Automatic Switchover to the Containment Sump. This change would require

a revision to the Engineered Safety Features Actuation System

Instrumentation Trip Setpoints, Table 3.3-4, Functional Unit 8.b, RWST

Level--Low-Low, along with associated Bases Section 3/4.3.2.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The proposed change does not adversely affect accident

initiators or precursors and does not alter the design assumptions

affecting the ability of the RWST and the ECCS [Emergency Core

Cooling System] pumps to mitigate the consequences of an accident.

Revising the RWST Level Low-Low setpoint has a negligible effect

on the operating margin for the RWST. The revised setpoint assures

that the minimum RWST volume assumed in the accident analyses is

injected prior to switchover to the recirculation mode. The effect

on containment flood level, equipment qualification, and pH of the

containment sump and the containment spray fluid, remain within the

limits assumed in the accident analyses.

Therefore, the proposed change does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

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

or different kind of accident from any previously analyzed.

The setpoint change does not affect the function of the level

monitoring channels or any function of the accident mitigation

equipment associated with the RWST. No new components or physical

changes are involved with this change. There are no changes to the

source term, containment isolation or radiological release

assumptions used in evaluating the radiological consequences in the

Seabrook Station [updated final safety analysis report] UFSAR. The

new setpoint will continue to initiate the automatic ECCS transfer

from the injection mode to the recirculation mode and provide the

alarm to alert the operator(s) to begin the manual actions necessary

to complete the transfer to the recirculation mode. Manual operator

action is required to complete the switchover to the recirculation

mode. With the new setpoint, sufficient time remains available for

the operator(s) to complete the transfer prior to receipt of the

RWST EMPTY alarm and reaching the vortexing level in the RWST.

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

new or different kind of accident from any previously analyzed.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The design bases for the RWST Level Low-Low setpoint is to

ensure that the minimum volume of water to support the assumptions

made in the safety analysis is injected prior to switchover and that

there is adequate time available for the operators to complete the

manual actions necessary to complete the switchover to the

recirculation mode prior to actuation of the RWST EMPTY alarm. The

minimum injection volume assumed in the accident analyses, and time

required for the operator(s) to initiate and complete manual actions

to complete switchover to the recirculation mode prior to receipt of

the RWST EMPTY alarm, remains unaffected by this change. Therefore,

the proposed change does not involve a significant reduction in a

margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Exeter Public Library,

Founders Park, Exeter, NH 03833.

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

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

CT 06141-0270.

NRC Project Director: Cecil O.Thomas.

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

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

Date of amendment request: May 21, 1998.

Description of amendment request: The proposed change would revise

selected Technical Specification (TS) surveillance requirements to

accommodate fuel cycles of up to 24 months for surveillances that are

currently performed at each 18-month or other specified outage

interval. Specifically, the following TS surveillance requirements

would be revised by the proposed change: 4.1.3.3, Digital Rod Position

Indication; 4.8.1.1.1.b, A.C. Sources--Operating--Transfer of 1E Bus

Power from Normal to Alternate Source; 4.8.1.1.2.f.1 through 15, A.C.

Sources--Operating--Emergency Diesel Generator Surveillances; 4.8.3.3,

Onsite Power Distribution--Trip Circuit For Inverter I-2A; 4.8.2.1.c, d

& f, D.C. Sources--Operating--125V D.C. Batteries and Chargers;

4.8.4.2.a.1) & a.2), Containment Penetration Conductor Overcurrent

Protective Devices and Protective Devices for Class 1E Power Sources

Connected to Non-Class 1E Circuits; 4.8.4.3, Motor Operated Valves

Thermal Overload Protection. In addition, the components listed in

Technical Specification 4.8.2.2, D.C. Sources--Shutdown--125V DC

Batteries and Chargers, have been evaluated to support an extension in

frequency to 24 months (+25%).

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

the probability or consequences of an accident previously evaluated.

The proposed changes do not adversely affect accident initiators

or precursors nor alter the design assumptions, conditions,

configuration of the facility or the manner in which the plant is

operated. The proposed changes do not alter or prevent the ability

of structures, systems, or components (SSCs) to perform their

intended function to mitigate the consequences of an initiating

event within the acceptance limits assumed in the Updated Final

Safety Analysis Report (UFSAR). The proposed changes are

administrative in nature and do not change the level of programmatic

controls or the procedural details associated with aforementioned

surveillance requirements.

Changing the frequencies of the aforementioned surveillance

requirements from at least once per 18 months to at least once per

refueling interval does not change the basis for the frequencies.

The frequencies were chosen because of the need to perform these

verifications under the conditions that are normally found during a

plant refueling outage, and to avoid the potential of an unplanned

transient if these surveillances were conducted with the plant at

power.

Equipment performance over several operating cycles was

evaluated to determine the impact of extending the surveillance

intervals. This evaluation included a review of surveillance

results, preventative maintenance records, and the frequency and

type of corrective maintenance activities, a

[[Page 43206]]

failure mode analysis, and consultation with the respective system

engineer. The evaluations conclude that the subject SSCs are highly

reliable, that presently do not exhibit time dependent failure modes

of significance, and that there is no indication that the proposed

extension could cause deterioration in the condition or performance

of the subject SSCs. There are no known mechanisms that would

significantly degrade the performance of the evaluated equipment

during normal plant operation. Although there have been generic or

repetitive failures of some components in the past, which may have

affected the ability of the SSCs to consistently and successfully

perform their safety function, those items have been resolved

through design changes and rework such that they have not recurred.

There have been no repetitive failures or time dependent failures

that were significant in nature which would have prevented the SSCs

from performing their intended safety function.

Deletion of the restriction ``during effect on safe operation of

the plant is given prior to conduct of a particular surveillance in

a condition or mode other than shutdown.

Since the proposed changes only affect the surveillance

intervals for SSCs that are used to mitigate accidents [sic], the

changes do not affect the probability or consequence of a previously

analyzed accident. While the proposed changes will lengthen the

intervals between surveillances, the increase in intervals has been

evaluated. Based on the reviews of the surveillance tests,

inspections, and maintenance activities, it is concluded that there

is no significant adverse impact on the reliability or availability

of these SSCs.

Since there are no changes to previous accident analyses, the

radiological consequences associated with these analyses remain

unchanged, therefore, the proposed changes do not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

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

or different kind of accident from any previously analyzed.

The proposed changes do not alter the design assumptions,

conditions, configuration of the facility or the manner in which the

plant is operated. There are no changes to the source term,

containment isolation or radiological release assumptions used in

evaluating the radiological consequences in the Seabrook Station

UFSAR. Existing system and component redundancy is not being changed

by the proposed changes. The proposed changes have no adverse impact

on component or system interactions. The proposed changes are

administrative in nature and do not change the level of programmatic

controls and procedural details associated with the aforementioned

surveillance requirements. Therefore, since there are no changes to

the design assumptions, conditions, configuration of the facility,

or the manner in which the plant is operated and surveilled, the

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

kind of accident from any previously analyzed.

3. The proposed changes do not involve a significant reduction

in a margin of safety.

There is no adverse impact on equipment design or operation and

there are no changes being made to the Technical Specification

required safety limits or safety system settings that would

adversely affect plant safety. The proposed changes are

administrative in nature and do not change the level of programmatic

controls and procedural details associated with the aforementioned

surveillance requirements.

From the evaluations performed on the subject SSCs there are no

indications that potential problems would be cycle-length dependent

or that potential degradation would be significant for the time

frame of interest and, therefore, increasing the surveillance

interval to the bounding limit of 30 months (24 months plus 25%)

will have little, if any, adverse affect on safety.

The proposed changes to the surveillance intervals are still

consistent with the basis for the intervals and the intent and

method of performing the surveillance is unchanged. Deletion of the

restriction ``during shutdown'' where this restriction is stated

will permit performance of certain maintenance and testing

activities during conditions or modes other than shutdown. North

Atlantic will ensure, through the implementation of appropriate

administrative controls, that proper regard to their effect on safe

operation of the plant is given prior to conduct of a particular

surveillance in a condition or mode other than shutdown. In

addition, use of the subject SSCs during normal plant operation,

combined with their previous history of availability and

reliability, provide assurance that the proposed changes will not

affect the reliability of the subject SSCs. Thus, it is concluded

that the subject SSCs would be available upon demand to mitigate the

consequences of an accident and, therefore, there is no impact on

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: Exeter Public Library,

Founders Park, Exeter, NH 03833.

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

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

CT 06141-0270.

NRC Project Director: Cecil O. Thomas.

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

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

Date of amendment request: July 2, 1998.

Description of amendment request: The proposed amendment would

revise the updated Final Safety Analysis Report (FSAR) by changing FSAR

Sections 9.7.2, ``Service Water,'' and 9.4, ``Reactor Building Closed

Cooling Water,'' to discuss the use of various types of internal

protective coatings and liners used in the piping and components of the

systems. The proposed change also indicates that periodic maintenance,

surveillances, and inspections would be conducted to ensure that

coating or liner degradation would be promptly detected and corrected

to provide reasonable assurance that the systems can perform their

safety-related 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:

The proposed change does not involve significant hazards

consideration because the changes would not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The SWS [Service Water System] provides cooling water directly

or indirectly to a multitude of mitigating and support systems such

as safety injection, containment spray, and RBCCW [Reactor Building

Closed-Cooling Water]. Therefore either directly or indirectly, the

SWS is credited in the mitigation of virtually all analyzed

operating events and accidents. However, there are no failures of

the SWS which would directly initiate any of the licensing basis

accidents. Therefore, the probability of occurrence of accidents

previously evaluated is not increased by this activity.

The SWS is comprised of two separate and independent trains,

each capable of providing the cooling capacity required for normal

and accident operation. Therefore, the failure of a single heat

exchanger or train will not influence the consequences of an

accident. Only a common mode loss of SWS function could affect

accident consequences. It can be postulated that lining material

could be released as a result of the SWS response to an accident or

as a result of a seismic event, resulting in heat exchanger blockage

in both trains (common mode). However, the discussion below provides

the basis for concluding that lining degradation will not increase

the consequences of an accident.

In response to a Safety Injection Actuation Signal or a Loss of

Normal Power event, the quantity of flow in safety related SWS heat

exchangers may increase significantly, imparting higher loads on the

pipe linings than are typically present during normal operation. In

spite of this flow increase, it is considered to be much more likely

that any lining degradation will occur and be detected under normal

operating conditions, and will be corrected prior to the occurrence

of an event of the type discussed above. SWS pump flow

surveillances, performed periodically during normal operation,

subject significant portions of the SWS to flow levels which equal

or exceed those expected to occur during accidents. Any degraded

lining material prone to be released during an

[[Page 43207]]

accident is expected to be released during these pump surveillances.

The inspections, operating procedures, and surveillances ensure that

significant lining releases will be promptly detected and

investigated. In addition, SWS design features provide the system

with a significant level of protection against degraded lining

debris (e.g., standby spare RBCCW heat exchanger and EDG [Emergency

Diesel Generator] engine cooler strainers) both during normal

operation and while responding to an accident.

An evaluation was performed to assess the significance of

loading on the linings due to a postulated seismic event. The

importance of seismic loads depends upon their magnitude relative to

normal operating loads, and on their relative frequency of

occurrence. Normal operating loads include steady state flow loads

as well as transients due to pump swaps and realignments for

surveillances. The evaluation determined that normal operating loads

are significantly greater than anticipated seismic loads concurrent

with steady state flow loads. Therefore, if normal operating loads

do not cause lining to become detached, it is very unlikely that a

random seismic event would cause detachment. In addition, while flow

loads are continuously present in most of the system and normal

transients occur many times during an operating cycle, seismic

events at the Millstone site are very infrequent (the repetition

rate of an OBE [Operating Basis Earthquake] is hundred of years).

Should normal operating loads cause lining detachment, it is much

more probable that this released material will be detected, and the

degraded condition corrected, prior to the occurrence of a seismic

event.

Based upon these discussions, and given the random nature of

lining degradation and the scrutiny with which the SWS is operated

and maintained, it is not considered to be credible that the

operability of both SWS trains will be simultaneously impaired by

lining degradation and release.

Therefore, there is no significant increase in the probability

or consequences of an accident previously evaluated.

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

from any accident previously evaluated.

As discussed above, the failure of a single heat exchanger or a

single SWS train will not cause an accident. Only a common mode loss

of SWS function could create the possibility of a previously

unanalyzed accident, and this loss would not directly initiate an

accident. However, for the reasons discussed above, lining

degradatiion will not cause common mode failures to occur.

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

or different kind of accident from any accident previously

evaluated.

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

The margins of safety of the protective boundaries (fuel matrix/

cladding, reactor coolant system pressure boundary, and containment)

would not be impacted by the postulated release of lining material

into the SWS. The accident analyses in the FSAR [Final Safety

Analysis Report] demonstrate the performance of the protective

boundaries. As discussed previously, it is not considered to be

credible that lining degradation will cause a common mode loss of

SWS function. Therefore, since the accident analyses credit only one

SWS train, released lining would not affect accident analyses

assumptions. On this basis, it is concluded that margins of safety

as demonstrated by the accident analyses would not be affected by

postulated lining material release.

Therefore, the change will not involve a significant reduction

in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Learning Resources Center,

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

Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano,

49 Rope Ferry Road, Waterford, Connecticut.

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

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

Connecticut.

NRC Deputy Director: Phillip F. McKee.

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

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

Date of amendment request: July 17, 1998.

Description of amendment request: The proposed amendment would

change the Technical Specifications (TS) surveillance requirements for

the onsite emergency diesel generators (EDGs) to achieve an overall

improvement in the EDGs reliability and availability. The proposed

changes would modify the requirement for operability tests of an EDG

when the other EDG is inoperable, delete the requirement for

operability tests when one or both offsite A.C. sources are inoperable,

eliminate fast loading of the EDGs except for the 18-month testing, and

eliminate fast starts (15 seconds) except for once per 6 months and

during the 18-month testing. These proposed changes are generally

consistent with the guidance provided in Generic Letter (GL) 84-15,

``Proposed Staff Actions to Improve and Maintain Diesel Generator

Reliability,'' dated July 2, 1984, and GL 93-05, ``Line-Item Technical

Specifications Improvements to Reduce Surveillance Requirements for

Testing During Power Operation,'' dated September 27, 1993.

Justification for deviations from the guidance provided in the GLs is

provided in the licensee's submittal.

In addition, the licensee proposes to revise the wording in the TS

requirements for offsite circuits to be consistent with NUREG-0212,

``Standard Technical Specifications for Combustion Engineering

Pressurized Water Reactors,'' Revision 2, fall 1980, and the guidance

provided in GL 91-04, ``Changes in Technical Specification Surveillance

Intervals to Accommodate 24-Month Fuel Cycle,'' dated April 2, 1991.

The associated TS Bases will be updated to reflect the proposed

changes.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The LCOs [Limiting Conditions for Operation] for Technical

Specifications [TSs] 3.8.1.1 and 3.8.1.2 will be changed to require

a transmission network between offsite power and the onsite Class 1E

distribution system, instead of just between offsite and the

switchyard. This change, which will expand the requirement, is

consistent with the current Millstone Unit No. 2 interpretation of

the required distribution system. Therefore, the proposed changes

will not result in a significant increase in the probability or

consequences of an accident previously analyzed.

The diesel generators (DGs) supply power to the emergency busses

at Millstone Unit No. 2 in the event of a loss of normal power

(LNP). The emergency busses supply the vital equipment used to

mitigate the consequences of design basis accidents. Therefore, the

diesel generators are vital equipment used to mitigate the

consequences of design basis accidents. Failure of the DGs will not

cause a design basis accident to occur. However, failure of the DGs

will affect the consequences of design basis accidents if a

concurrent LNP occurs.

The proposed changes will revise the action requirements

regarding operability testing of the DGs. The requirement to test

the DGs if offsite circuits are inoperable will be deleted. An

inoperable offsite circuit, by itself, will not affect the

operability of the DGs. The requirement to test the remaining

operable DG if one DG is inoperable will be modified. Testing will

not be required provided a common cause failure is not the reason

for declaring the DG inoperable. The requirement contained in the

first footnote (*) to Technical Specification 3.8.1.1 to complete

the test of the remaining DG will be deleted. The need to test the

remaining DG will be based on the determination of a common cause

failure. These changes will improve DG reliability by reducing the

number of unnecessary starts and by requiring more appropriate

testing of the DGs when there is a potential for common mode

[[Page 43208]]

failure. The proposed changes to the action requirements will not

change the response of the DGs to an LNP. Therefore, the proposed

changes will not result in a significant increase in the probability

or consequences of an accident previously analyzed.

The requirement contained in the second footnote (**) to

Technical Specification 3.8.1.1 to allow a one time extension of the

allowed outage time to 7 days will be deleted. This provision is no

longer necessary since the Millstone Unit No. 1 work has been

completed. The statements that a successful test of the DG performed

for the current Action Statements c, d, or e will satisfy the

required testing of Action States a or b are no longer necessary

with the proposed changes. These statements will be deleted. The

removal of these items will not change the response of the DGs to an

LNP. Therefore, these proposed changes will not result in a

significant increase in the probability or consequences of an

accident previously analyzed.

The proposed changes to the DG surveillance requirements will

allow an engine prelube period before all DG tests starts, allow

slow starting of the DGs, and allow the DGs to be loaded in

accordance with manufacturer recommendations. This will decrease the

wear on the DGs. The proposed changes will also allow adequate time

for the completion of all manufacturer recommended DG engine prelube

procedures. Modifying starting and loading requirements, consistent

with the manufacturer recommendations, is intended to enhance diesel

reliability by minimizing severe test conditions which can lead to

premature failures. In addition, specifying that the 184 day DG SRs

[surveillance requirements] will satisfy the 31 day DG starting and

loading SRs will eliminate redundant testing. These proposed changes

will minimize unnecessary DG testing while maintaining DG

reliability. The proposed changes will not change the response of

the DGs to an LNP. Therefore, these changes will not result in a

significant increase in the probability or consequences of an

accident previously analyzed.

The ASTM [American Society for Testing and Materials] standards

referenced for diesel fuel oil sampling will be modified in SR

4.8.1.1.2.b. The proposed changes will replace an outdated standard,

and will remove the year of issuance or revision from the ASTM

standards referenced. This will allow use of the current approved

ASTM standard. These proposed changes do not affect the sampling

frequency or acceptance criteria of this SR. Therefore, the proposed

changes will not result in a significant increase in the probability

or consequences of an accident previously analyzed.

The proposed wording changes to eliminate any possible confusion

when SRs 4.8.1.1.1 and 4.8.1.1.2 are referenced by SR 4.8.1.2, to

state that the DGs start from standby conditions instead of ambient

conditions, and to remove the requirement to perform a DG

surveillance only during shutdown will not affect any technical

aspect of the SRs. Therefore, the proposed changes will not result

in a significant increase in the probability or consequences of an

accident previously analyzed.

SRs will be added to test the DGs every 184 days at conditions

similar to the current 31 day SRs. These conditions are more

restrictive than the new proposed 31 day SRs. The 184 day SRs will

require the diesel generators to start and obtain speed and voltage

within 15 seconds and will also require the diesel generators to be

synchronized, loaded, and to maintain the load for at least 60

minutes. However, it will allow gradual loading, based on

manufacturer recommendations, to be used. A 184 day surveillance

interval is sufficient to verify DG fast-start capability, and is

consistent with GL [Generic Letter] 84-15, GL 93-05, and NUREG-1432.

Therefore, the posed changes will not result in a significant

increase in the probability or consequences of an accident

previously analyzed.

The list of SRs, contained in SR 4.8.1.2, that do not have to be

performed for the operable diesel generator in Modes 5 and 6 will be

expanded to take into account the 184 day DG SR that will be added.

This proposed change will exclude the one operable DG from being

loaded when the 184 day SR is performed. This is consistent with the

current SR which excludes performance of SR 4.8.1.1.2.a.3. Loading

the one required operable diesel generator could subject this diesel

generator to grid faults which could adversely affect its ability to

perform its safety function. Therefore, the proposed change will not

result in a significant increase in the probability or consequences

of an accident previously analyzed.

The Bases of these Technical Specifications will be modified and

expanded to discuss the proposed changes, and to provide guidance to

ensure the requirements are correctly applied. Therefore, the

proposed changes will not result in a significant increase in the

probability or consequences of an accident previously analyzed.

These proposed changes do not alter the way any structure,

system, or component functions. The intent of the proposed changes

is to improve the reliability of the DGs by eliminating unnecessary

surveillance testing and allowing most of the surveillance testing

to be performed in accordance with the recommendations of the

manufacturer. There will be no adverse effect on equipment important

to safety. The response of the DGs to an LNP, as described in the

Millstone Unit No. 2 FSAR [Final Safety Analysis Report], will

remain the same. There will be no effect on any of the design basis

accidents previously evaluated. Therefore, this License Amendment

Request will not result in a significance increase in the

probability or consequences of an accident previously evaluated.

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

accident from any accident previously evaluated.

The proposed changes do not alter the plant configuration (no

new or different type of equipment will be installed) or require any

new or unusual operator actions. They do not alter the way any

structure, system, or component functions and do not alter the

manner in which the plant is operated. The proposed changes do not

introduce any new failure modes. Therefore, the proposed changes

will 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.

This License Amendment Request proposes to modify the LCOs for

electrical power sources, DG surveillance requirements and the

required actions for inoperable electrical power sources contained

in the Millstone Unit No. 2 Technical Specifications. The proposed

changes will revise LCO wording to be consistent with the required

offsite power distribution requirements and improve DG reliability

by minimizing excessive wear of the DGs, and changing the starting

and loading requirements of the DGs, in accordance with manufacturer

recommendations, during most DG surveillance and operability tests.

Improving the reliability of the DGs will help ensure the DGs will

respond to an LNP as described in the Millstone Unit No. 2 FSAR.

Therefore, this License Amendment Request will not result in a

significant reduction in the margin of safety as defined in the

Bases for the Technical Specifications addressed by the proposed

changes.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Learning Resources Center,

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

Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano,

49 Rope Ferry Road, Waterford, Connecticut.

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

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

Connecticut.

NRC Deputy Director: Phillip F. McKee.

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

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

Date of amendment request: July 21, 1998.

Description of amendment request: The proposed amendment would

change the Technical Specifications (TS) by changing various Reactor

Protection System (RPS) and Engineered Safety Features Actuation System

(ESFAS) setpoints and allowable values; correct the specified maximum

reactor power level limited by the high power level RPS trip; add new

TS and requirements associated with the automatic isolation of steam

generator blowdown; and make several editorial and changes to correct

various errors

[[Page 43209]]

and to provide needed clarification. The applicable TS Bases sections

would also be changed to reflect the proposed changes, correct previous

errors identified during the licensee's review of the TS, eliminate

redundant information, and expand the TS Bases to discuss the new

requirements for the automatic isolation of the steam generator

blowdown.

Specifically, the proposed changes would modify TS 2.1.1, ``Safety

Limits--Reactor Core,'' TS 2.2.1, ``Limiting Safety System Settings--

Reactor Trip Setpoints,'' TS 3.3.1.1, ``Instrumentation--Reactor

Protective Instrumentation'' TS 3.3.2.1, ``Instrumentation--Engineered

Safety Features Actuation System Instrumentation,'' and would add a new

TS 3.7.1.8, ``Plant Systems--Steam Generator Blowdown Isolation

Valves.'' As previously noted, the applicable TS Bases sections will be

updated to reflect the proposed changes.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed change to correct the maximum reactor power level

from 112% to 111.6% is consistent with the maximum high power trip

setpoint of 106.6%, plus 5% uncertainty, currently used in the

safety analyses. This does not change the Technical Specification

required high power reactor trip setpoint. There will be no adverse

effect on any design basis accident previously evaluated or on any

equipment important to safety. Therefore, the proposed change will

not result in a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed changes to the trip setpoints and allowable values

for the Reactor Protection System (RPS) trips on high pressurizer

pressure, high containment pressure, low steam generator pressure,

and low steam generator level are the result of revisions to the

instrument loop uncertainty and setpoint calculations. These

calculations were revised to incorporate calculation methodology

changes, analytical limit changes, correct errors identified, and to

include the effects of a harsh environment (pressure, temperature,

and radiation), where appropriate. The proposed setpoints and

allowable values will ensure a reactor trip signal is generated at,

or before the analytical limits used in the respective accident

analyses are reached. There will be no adverse effect on any design

basis accident previously evaluated or on any equipment important to

safety. Therefore, the proposed changes will not result in a

significant increase in the probability or consequences of an

accident previously evaluated.

The proposed changes to the trip setpoints and allowable values

for the Engineered Safety Features Actuation System (ESFAS)

actuations on low pressurizer pressure, high containment pressure,

low steam generator pressure, low refueling water storage tank

level, and low steam generator level are the result of revisions to

the instrument loop uncertainty and setpoint calculations. These

changes were revised to incorporate calculation methodology changes,

analytical limit changes, correct errors identified, and to include

the effects of a harsh environment (pressure, temperature, and

radiation), where appropriate. The proposed setpoints and allowable

values will ensure an ESF [engineered safety feature] actuation

signal is generated at, or before the analytical limits used in the

respective accident analyses are reached. There will be no adverse

effect on any design basis accident previously evaluated or on any

equipment important to safety. Therefore, the proposed change will

not result in a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed change to add Technical Specification requirements

for the steam generator blowdown isolation valves will provide

additional assurance that the automatic isolation of steam generator

blowdown will occur as assumed in the loss of main feedwater

accident analysis. There will be no adverse effect on any design

basis accident previously evaluated or on any equipment important to

safety. Therefore, the proposed changes will not result in a

significant increase in the probability or consequences of an

accident previously evaluated.

The proposed change to the value of steam generator pressure

when the steam generator low pressure reactor trip can be bypassed

(from 780 psia to 800 psia) will reduce the range of plant operation

when this trip is required to be available. However, this will not

affect the range of plant operation when this RPS trip is required

to be operable. This RPS trip is required in Modes 1 and 2. The

expected steam generator pressure during a reactor startup (entry

into Mode 2) is approximately 900 psia, which corresponds to a

Reactor Coolant System (RCS) temperature of approximately 532 deg.F.

The proposed change will require the bypass to be automatically

removed prior to exceeding a steam generator pressure of 800 psia.

There will be no adverse effect on any design basis accident

previously evaluated or on any equipment important to safety.

Therefore, the proposed change will not result in a significant

increase in the probability or consequences of an accident

previously evaluated.

The proposed change to the value of pressurizer pressure (from

1750 psia to 1850 psia) when the pressurizer low pressure ESF

actuations (SIAS, CIAS, and EBFAS) [safety injection actuation

system, containment isolation actuation system, and enclosure

building filtration actuation system] can be blocked will reduce the

range of plant operation when these functions are required to be

available. However, since the plant would normally be in Mode 3 when

pressurizer pressure is in this range, automatic actuation of these

ESF functions on high containment pressure, as well as manual

actuation, is required to be operable. In addition, the plant would

not normally maintain pressurizer pressure between 1750 psia and

1850 psia. Therefore, since automatic actuation of these ESF

functions on high containment pressure, as well as manual actuation,

should be operable, and the time the plant will operate between 1750

psia and 1850 psia is small, the ESFAS will continue to function as

before. There will be no adverse effect on any design basis accident

previously evaluated or on any equipment important to safety.

Therefore, the proposed change will not result in a significant

increase in the probability or consequences of an accident

previously evaluated.

The proposed change to the value of steam generator pressure

(from 600 psia to 700 psia) when the steam generator low pressure

ESF actuation (main steam line isolation) can be blocked will reduce

the range of plant operation when this function is required to be

available. However, since the plant would be in Mode 3 when steam

generator pressure is in this range (RCS temperature of

approximately 486 deg.F to 503 deg.F), automatic actuation of this

ESF function on high containment pressure, as well as manual

actuation, is required to be operable. In addition, the plant would

not normally maintain steam generator pressure between 600 psia and

700 psia. Therefore, since automatic actuation of this ESF function

on high containment pressure, as well as manual actuation, should be

operable, and the time the plant will operate between 600 psia and

700 psia is small, the ESFAS will continue to function as before.

There will be no adverse effect on any design basis accident

previously evaluated or on any equipment important to safety.

Therefore, the proposed change will not result in a significant

increase in the probability or consequences of an accident

previously evaluated.

The minor editorial and non-technical changes to correct

spelling errors, correct a capitalization error, add page amendment

numbers, add the specific plant parameter (steam generator pressure)

to use if an RPS or ESF function can be bypassed, change the value

of the parameter (pressurizer pressure) used in action statements,

and a ``[less than or equal to]'' symbol, change ``value'' to

``setpoint,'' and update the index will have no effect on plant

operation. These changes will not result in any technical changes to

the Millstone Unit No. 2 Technical Specifications. There will be no

adverse effect on any design basis accident previously evaluated or

on any equipment important to safety. Therefore, the proposed change

will not result in a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed changes to the Technical Specification Bases will

incorporate the RPS and ESFAS setpoint changes, correct errors,

eliminate redundant information, and expand the Bases to discuss the

new requirements for steam generator blowdown isolation. These

changes will have no effect on equipment operation. There will be no

adverse effect on any design basis accident

[[Page 43210]]

previously evaluated or on any equipment important to safety.

Therefore, the proposed changes will not result in a significant

increase in the probability or consequences of an accident

previously evaluated.

The proposed changes have no adverse effect on any of the design

basis accidents previously evaluated and have no adverse effect on

how the RPS and ESFAS function to mitigate the consequences of

design basis accidents. Therefore, the license amendment request

does not impact the probability of an accident previously evaluated

nor does it involve a significant increase in the consequences of an

accident previously evaluated.

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

from any accident previously evaluated.

The proposed changes will not alter the plant configuration (no

new or different type of equipment will be installed) or require any

new or unusual operator actions. They do not alter the way any

structure, system, or component functions and do not alter the

manner in which the plant is operated. The proposed changes do not

introduce any new failure modes. Therefore, the proposed changes

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

accident from any accident previously evaluated.

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

The proposed changes will correct the maximum reactor power

level specified; change RPS trip setpoints, allowable values, and

bypass setpoints; change ESFAS trip setpoints, allowable values, and

block setpoint changes; add a new Technical Specification and

additional requirements associated with the automatic isolation of

steam generator blowdown; and make various minor editorial and non-

technical changes. There will be no adverse effect on equipment

important to safety. The RPS and ESFAS will continue to function as

designed to mitigate the consequences of design basis accidents.

Therefore, there will be no significant reduction of the margin of

safety as defined in the Bases for the Technical Specifications

affected by the proposed changes.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Learning Resources Center,

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

Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano,

49 Rope Ferry Road, Waterford, Connecticut.

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

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

Connecticut.

NRC Deputy Director: Phillip F. McKee.

Pennsylvania Power and Light Company, Docket No. 50-387, Susquehanna

Steam Electric Station, Unit 1, Luzerne County, Pennsylvania

Date of amendment request: June 19, 1998.

Description of amendment request: The amendment to Unit 1 Technical

Specifications (TS) involves the addition of a new section entitled

``Oscillation Power Range Monitoring (OPRM) Instrumentation'' and

revisions to Section 3.4.1 ``Recirculation Loops Operating'' to remove

the specifications related to thermal power stability which will not be

required after the installation of the OPRM instrumentation. Unit 1 is

currently operating under Interim Corrective Actions (ICAs) defined in

TS 3.4.1 that specify restrictions on plant operation and actions by

operators in response to instability events. The OPRM system provides

an automatic long-term solution to the instability issue and eases the

burden on the operator.

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.

This proposal does not involve an increase in the probability or

consequences of an accident previously evaluated.

The OPRM most directly affects the APRM and LPRM portions of the

Power Range Neutron Monitoring system. Its installation does not

affect the operation of these sub-systems. None of the accidents or

equipment malfunctions affected by these sub-systems are affected by

the presence or operation of the OPRM.

The APRM channels provide the primary indication of neutron flux

within the core and respond almost instantaneously to neutron flux

changes. The APRM Fixed Neutron Flux-High function is capable of

generating a trip signal to prevent fuel damage or excessive reactor

pressure. For the ASME overpressurization protection analysis in

FSAR Chapter 5, the APRM Fixed Neutron Flux-High function is assumed

to terminate the main steam isolation valve closure event. The high

flux trip, along with the safety/relief valves, limit the peak

reactor pressure vessel pressure to less than the ASME Code limits.

The control rod drop accident (CRDA) analysis in Chapter 15 takes

credit for the APRM Fixed Neutron Flux-High function to terminate

the CRDA. The Recirculation Flow Controller Failure event (pump

runup) is also terminated by the high neutron flux trip. The APRM

Fixed Neutron Flux-High function is required to be OPERABLE in MODE

1 where the potential consequences of the analyzed transients could

result in the Safety Limits (e.g., MCPR and Reactor pressure) being

exceeded.

The installation of the OPRM equipment does not increase the

consequences of a malfunction of equipment important to safety. The

APRM and RPS systems are designed to fail in a tripped (fail safe)

condition; the OPRM will have no affect on the consequence of the

failure of either system. An inoperative trip signal is received by

the RPS any time an APRM mode switch is moved to any position other

than Operate, an APRM module is unplugged, the electronic operating

voltage is low, or the APRM has too few LPRM inputs. These functions

are not specifically credited in the accident analysis, but are

retained for the RPS as required by the NRC approved licensing

basis.

The OPRM allows operation under current operating conditions

presently restricted by the current Technical Specifications by

providing automatic suppression functions in the area of concern in

the event an instability occurs. The consequences of any accident or

equipment malfunction are not increased by operating under those

conditions. Although protected by the OPRM from thermal-hydraulic

core instabilities above 30% core power, operation under natural

core recirculation conditions is not allowed. No accidents or

transients of a type not analyzed in the FSAR are created by

operating under these conditions with the protection of the OPRM

system.

This change does not increase the probability of an accident as

previously evaluated. The OPRM is designed and installed to not

degrade the existing APRM, LPRM, and RPS systems. These systems will

still perform all of their intended functions. The new equipment is

tested and installed to the same or more restrictive environmental

and seismic envelopes as the existing systems. The new equipment has

been designed and tested to the electromagnetic interference (EMI)

requirements of Reference 2, which assures correct operation of the

existing equipment. The new system has been designed to single

failure criteria and is electrically isolated from equipment of

different electrical divisions and from non-1E equipment. The

electrical loading is within the capability of the existing power

sources and the heat loads are within the capability of existing

cooling systems. The OPRM allows operation under operating

conditions presently forbidden or restricted by the current

Technical Specifications. No other transient or accident analysis

assumes these operating restrictions.

Based upon the analysis presented above, PP&L concludes that the

proposed action does not involve an increase in the probability or

consequences of an accident previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

This proposal does not create the probability of a new or

different type of accident from any accident previously evaluated.

The OPRM system is a monitoring and accident mitigation system that

cannot create the possibility for an accident.

[[Page 43211]]

The OPRM will allow operation in conditions currently restricted

by the current Technical Specifications. Although protected by the

OPRM from thermal-hydraulic core instabilities above 30% core power,

operation under natural circulation conditions is not allowed. No

accidents or transients of a type not analyzed in the FSAR are

created by operating under these conditions with the protection of

the OPRM system. No new failure modes of either the new OPRM

equipment or of the existing APRM equipment have been introduced.

Quality software design, testing, implementation and module self-

health testing provides assurance that no new equipment malfunctions

due to software errors are created. The possibility of an accident

of a new or different type than any evaluated previously is not

created.

The new OPRM equipment is designed and installed to the same

system requirements as the existing APRM equipment and is designed

and tested to have no impact on the existing functions of the APRM

system. Appropriate isolation is provided where new interconnections

between redundant separation groups are formed. The OPRM modules

have been designed and tested to assure that no new failure modes

have been introduced.

Therefore, the proposed change does not create the possibility

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

evaluated.

3. The proposed change does not involve a significant reduction

in the margin of safety.

There has been no reduction in the margin of safety as defined

in the basis for the Technical Specifications. The OPRM system does

not negatively impact the existing APRM system. As a result, the

margins in the Technical Specifications for the APRM system are not

impacted by this addition.

Current operation under the ICAs provides an acceptable margin

of safety in the event of an instability event as the result of

preventive actions and Technical Specification controlled response

by the control room operators. The OPRM system provides an increase

in the reliability of the protection of the margin of safety by

providing automatic protection of the MCPR safety limit, while the

protection burden is significantly reduced for the control room

operators. This protection is demonstrated as described above, and

in the NRC reviewed and approved Topical Reports NEDO-32465-A and

CENPD-400-P-A.

Replacement of the ICA operating restrictions from Technical

Specifications with the OPRM system does not affect the margin of

safety associated with any other system or fuel design parameter.

Therefore, the change does not involve a reduction in the margin of

safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Osterhout Free Library,

Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701.

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

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

NRC Project Director: Robert A. Capra.

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: July 6, 1998

Description of amendment request: The proposed Technical

Specification (TS) changes represent revisions to the Radiological

Effluent Technical Specification (RETS) Section 3.5.b.1, ``Main

Condenser Steam Jet Air Ejector (SJAE)'' and Table 3.10-1 ``Radiation

Monitoring Systems that Initiate and/or Isolate Systems'' including

associated TS Bases. The existing RETS for radiation monitoring

instrumentation systems that initiate and/or isolate systems will be

changed by adding Allowable Outage Times (AOTs) and incorporating

editorial and administrative changes to clarify requirements.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The inherent redundancy and reliability of the protective

instrumentation trip systems ensure that the consequences of an

accident are not significantly increased. In addition, the

restrictive Allowable Outage Time (AOT) interval limits the

probability of the protective instrument channel being unavailable

and an accident requiring its function from occurring

simultaneously. The requirement that the associated trip function

maintains trip capability for selected instrumentation ensures that

the protective instrumentation response will occur such that the

consequences of an accident are not different from those previously

evaluated. The proposed changes provide AOTs for test and repair of

plant instrumentation. 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 of occurrence of

an accident.

Regarding the consequences of an accident, the GE Licensing

Topical Reports (References 1 and 2) [GE Topical Report NEDC-31677P-

A, ``Technical Specification Improvement Analysis for BWR Isolation

Actuation Instrumentation,'' July 1990 and GE Topical Report GENE-

770-06-1-A, ``Bases for Changes to Surveillance Test Intervals and

Allowed Out-Of-Service Times for Selected Instrumentation Technical

Specifications,'' December 1992] conclude that the proposed AOT for

the safety system instrumentation results in an insignificant change

in the core damage frequency. The AOTs result in a slight increase

in the unavailability of the safety functions. The overall effect on

the probability of an accident is negligible. The NRC concurred in

their SERs [safety evaluation reports] (References 3 and 4) [NRC

Safety Evaluation Report, letter from Charles E. Rossi, NRC to S.D.

Floyd, BWR Owners Group, ``General Electric Company Topical Report

NEDC-31677P, Technical Specification Improvement Analysis for BWR

Isolation Actuation Instrumentation'', June 18, 1990 and NRC Safety

Evaluation Report, letter from Charles E. Rossi, NRC to R.D. Binz,

BWR Owners Group, ``General Electric Company Topical Report GENE-

770-06-1, Bases for Changes to Surveillance Test Intervals and

Allowed Out-Of-Service Times for Selected Instrumentation Technical

Specifications,'' July 21, 1992] with this conclusion. Consequently,

there is not a significant increase in the consequences of an

accident.

Since the editorial and administrative items do not alter the

meaning or intent of any requirements, they do not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

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

from any accident previously evaluated.

The proposed changes to the protective instrumentation trip

system specifications do not create the possibility of a new or

different kind of accident because they do not introduce any new

operational modes or physical modifications to the plant.

For systems with only one channel (Main Control Room

Ventilation) or two-out-of-two logic system (SJAE Radiation

Monitors) a six-hour surveillance AOT is being proposed and a repair

time AOT is not allowed. This is consistent with GE Topical Reports

referenced in current TS Bases 4.2 and STS [Standard Technical

Specifications] and therefore, will not introduce a new or different

kind of accident than previously evaluated.

Since the editorial and administrative items do not alter plant

configurations or operating modes, they do not create the

possibility of a new or different kind of accident.

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

The protective instrumentation surveillance requirements provide

verification of the operability of the trip system instrumentation

channels. In addition, the redundant channel that monitors the

identical Trip Function maintains trip capability for the relatively

short duration of the test or repair time period. This ensures that

protective

[[Page 43212]]

instrumentation reliability is maintained. The proposed change

provides for a specific time period to perform required

surveillances on instrument channels without trips present in

associated trip systems. This time allotment tends to enhance the

margin of safety by decreasing the probability of unnecessary

challenges to safety systems and inadvertent plant transients. The

evaluations presented in the referenced GE Licensing Topical Reports

concluded that the overall effect of the proposed changes provides a

net increase in plant safety.

The only action resulting from the proposed changes to RETS is

to add AOTs for selected instrumentation. Spurious signals during

testing could initiate plant transients. These transients are

bounded by the current transient analysis. These tests do not

subject the instruments to any conditions beyond their design

specifications and are performed in accordance with approved testing

standards. This testing ensures equipment operability by identifying

degraded conditions, initiating corrective action and properly

retesting them. Therefore, the proposed RETS do not involve a

significant reduction in a margin of safety.

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

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

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

amendment 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. David E. Blabey, 1633 Broadway, New

York, New York 10019.

NRC Project Director: S. Singh Bajwa, Director

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

Generating Station, Salem County, New Jersey

Date of amendment request: June 25, 1998.

Description of amendment request: The proposed changes affect

Technical Specification (TS) Surveillance Requirement 4.5.1.d.2.b by

deleting the requirement to perform in-situ functional testing of the

Automatic Depressurization System (ADS) safety relief valves (SRVs)

during startup testing activities. The proposed changes also affect TS

Surveillance Requirement 4.4.2.1.b such that the 18-month channel

calibration for the SRV acoustic monitors will no longer require an

exception to the provisions of TS 4.0.4, nor adjustments to SRV full

open noise levels.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

the probability or consequences of an accident previously evaluated.

The proposed TS change does not involve any physical changes to

plant structures, systems or components (SSC). The ADS will continue

to function as designed. The ADS is an Emergency Core Cooling System

(ECCS) designed to mitigate the consequences of an accident, and

therefore, can not contribute to the initiation of any accident. The

ADS utilizes five of the 14 main steam line SRVs as the primary

method for depressurizing the reactor pressure vessel to permit low

pressure core cooling capability in the event of a small break Loss-

of-Coolant-Accident (LOCA) if the high pressure cooling systems

(i.e., High Pressure Cooling Injection (HPCI) and Reactor Core

Isolation Cooling (RCIC) systems) fail to maintain adequate reactor

vessel water level.

Deleting the TS surveillance requirements to perform the in-situ

testing of the ADS/SRVs during startup, as proposed, should reduce

the probability of an inadvertent opening of an SRV as discussed in

Section 15.1.4 of the Hope Creek [Updated Final Safety Analysis

Report] UFSAR since deleting this testing requirement will eliminate

a known initiator of SRV pilot leakage and subsequent erosion. This

proposed TS change will have a tendency to increase, rather than

decrease, the reliability of the ADS/SRVs by eliminating the in-situ

ADS functional startup testing. The probability of the ADS/SRVs to

open on demand has been demonstrated to be extremely high and is not

measurably improved through the in-situ ADS functional startup

testing.

Using the provisions of 10CFR50.59, PSE&G will establish a

method for performing SRV acoustic monitor channel calibration that

does not require reactor steam pressure or SRV opening. This testing

method will comply with the current TS definition of CHANNEL

CALIBRATION. Since the notes associated with TS Surveillance

Requirement 4.4.2.1 (providing a compliance exception to the

provisions of TS 4.0.4 to allow for proper reactor steam pressure to

perform the test and an allowance for noise level adjustments) are

no longer needed, their removal will not affect plant operation or

testing and will not involve an increase in the probability or

consequences of an accident previously evaluated.

This proposed TS change will not increase the probability of

occurrence of a malfunction of any plant equipment important to

safety. Alternate testing methods at Hope Creek and at the offsite

test facility adequately demonstrate proper ADS valve operation and

assure that the valves will continue to function as designed.

Existing surveillance testing and inspections of the ADS/SRVs at

Hope Creek verify that the ADS initiation logic, solenoid valve

operation, pneumatic gas supply integrity and air operator assembly

(including pilot rod) will operate as designed. Offsite testing

verifies pilot disc operation, setpoint calibration, stroke time and

main valve disc operation.

Deleting the in-situ testing requirement, as proposed, will

reduce the probability of increasing SRV leakage, which should

reduce the probability of an inadvertent opening of an SRV.

Therefore, any SRV pilot leakage that can be eliminated would reduce

the probability of occurrence of a malfunction of that SRV. Deleting

the ADS/SRV in-situ functional test will in no way increase any

consequences of a malfunction of plant equipment important to

safety. The consequences of a malfunction of an ADS/SRV as discussed

in the Hope Creek UFSAR remain unchanged.

In addition, eliminating a known initiator of SRV leakage, as

proposed in this TS change, would help reduce operator workarounds

in the form of suppression pool cooling and letdown operation

activities. As a result, this will reduce the unnecessary operation

of the Residual Heat Removal (RHR) and its supporting systems.

Therefore, the proposed TS change does not involve an increase

in the probability or consequences of an accident previously

evaluated.

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

or different kind of accident from any accident previously

evaluated.

The proposed TS changes do not involve any physical changes to

plant SSC. The design and operation of the ADS/SRVs are not changed

from that currently described in the UFSAR. The ADS will continue to

function as designed to mitigate the consequences of an accident. No

changes of any kind are being made to the valves, auxiliary

components or ADS logic. Deleting the requirement to perform the ADS

in-situ functional test during plant startup as proposed in this TS

change request reduces the likelihood of an SRV developing a leak

and degrading throughout the subsequent operating cycle. Therefore,

there is no possibility that implementing this proposed TS change

would create a different type of malfunction to the ADS/SRVs than

any previously evaluated.

Eliminating the requirement to perform the in-situ testing of

the ADS/SRVs during startup activities does not create a new or

different type of accident than any previously evaluated. There is

no accident scenario associated with testing the ADS/SRVs other than

the inadvertent opening of a relief valve, which is currently

discussed in Section 15.1.4 of the UFSAR. The proposed TS changes do

not alter the conclusions described in the UFSAR regarding an

inadvertent opening of an SRV. No new or different type of accident

will be created as a result of these proposed changes.

Therefore, the proposed TS change does not create the

possibility of a new or different kind of accident from any

previously evaluated.

Using the provisions of 10CFR50.59, PSE&G will establish a

method for performing SRV acoustic monitor channel calibration that

does not require reactor

[[Page 43213]]

steam pressure or SRV opening. This testing method will comply with

the current TS definition of CHANNEL CALIBRATION. Since the notes

associated with TS Surveillance Requirement 4.4.2.1 (providing a

compliance exception to the provisions of TS 4.0.4 to allow for

proper reactor steam pressure to perform the test and an allowance

to perform noise level adjustments) are no longer needed, their

removal will not affect plant operation or testing and will not

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

any previously evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The proposed TS change involves deleting the requirement to

perform in-situ functional testing of the ADS/SRVs during startup

activities. This testing imposes an unnecessary challenge on the

ADS/SRVs and has been linked to SRV degradation (e.g., pilot valve

and/or main valve leakage). This proposed TS change should reduce

SRV leakage and improve ADS/SRV reliability by reducing the

potential for spurious SRV actuation. Since ADS operability can be

readily demonstrated with extremely high confidence by the existing

surveillance tests and inspections performed for the ADS, there will

be no reduction in any margin of safety resulting from this proposed

TS change. Therefore, the proposed TS change does not involve a

significant reduction in a margin of safety.

Using the provisions of 10CFR50.59, PSE&G will establish a

method for performing SRV acoustic monitor channel calibration that

does not require reactor steam pressure or SRV opening. This testing

method will comply with the current TS definition of CHANNEL

CALIBRATION. Since the notes associated with TS Surveillance

Requirement 4.4.2.1 (providing a compliance exception to the

provisions of TS 4.0.4 to allow for proper reactor steam pressure to

perform the test and an allowance to perform noise level

adjustments) are no longer needed, their removal will not affect

plant operation or testing and will not involve a significant

reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Pennsville Public Library, 190

S. Broadway, Pennsville, NJ 08070.

Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business

Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.

NRC Project Director: Robert A. Capra.

Tennessee Valley Authority, Docket No. 50-390 Watts Bar Nuclear Plant,

Unit 1, Rhea County, Tennessee

Date of amendment request: February 18, 1998.

Description of amendment request: The proposed amendment would

revise the Watts Bar Nuclear Plant (WBN) Technical Specifications (TS)

and associated Bases to address a new condition (Condition B) and

associated actions in which one train (consisting of two valves) of

Steam Generator Atmospheric Dump Valves (ADVs), although functional,

would be considered technically INOPERABLE in the event of one train of

the auxiliary control air system (ACAS) was out of service. The action

required for the new condition is to restore the ADV lines to OPERABLE

status within 72 hours. In addition, the proposed amendment would make

a correction to the required action for Condition B (new Condition C)

to clarify that the required action for two or more inoperable ADV

lines (with the exception of new Condition B) is to restore all but one

ADV line to operable status. The current Required Action for Condition

B incorrectly states that only one ADV line must be restored to

operable status.

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:

A. The proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

The addition of the 72 hour completion time and clarification to

existing TS do not increase the probability of an accident

previously evaluated since these changes do not result in hardware

or procedural changes which will affect probability of occurrence of

an accident. The probability of an accident occurring during the 72

hour period as compared to the 24 hour completion time currently in

the TS remains small. Further, addition of the 72 hour completion

time and clarification to existing TS does not increase the

consequences of an accident previously evaluated since sufficient

equipment and procedures remain available to mitigate accidents

previously evaluated. With two ADVs inoperable under this LCO, two

ADVs remain in service. As indicated in the Applicable Safety

Analysis of the TS Basis, two valves are adequate to cool the unit

to the RHR [residual heat removal] entry conditions subsequent to

accidents accompanied by a loss of offsite power. In addition, as

indicated in the background discussion of the Bases of 3.7.4, the

ADVs can be operated by use of a bottled nitrogen system designed to

open the valves in the event of loss of normal and emergency air

supplies. The valves may also be operated manually by using the

valve hand wheels. Consequently, the two inoperable ADVs under this

LCO are still expected to remain functional and could be placed in

service and used to cool the steam generators, if necessary, in the

event of an accident. Based on the above, the addition of the 72

hour completion time and clarifications to existing TS in accordance

with this proposed amendment do not significantly increase the

probability or consequences of an accident previously evaluated.

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

new or different kind of accident from any accident previously

evaluated.

The addition of the 72 hour completion time and clarifications

to existing TS does not cause the initiation of any accident nor

create any new credible limiting failure for safety-related systems

and components. The change does not result in an event previously

deemed incredible being made credible. As such, it does not create

the possibility of an accident different than any evaluated in the

FSAR [Final Safety Analysis Report]. The change has an insignificant

effect on the ability of the safety-related systems to perform their

intended safety functions. Although the period during which a

safety-related function (ACAS air supply) is assumed inoperable is

extended from 24 to 72 hours, sufficient remaining equipment (two

ADVs supplied by the opposite train ACAS) is available to mitigate

the limiting [steam generator tube rupture] SGTR accident, assuming

no single failure occurs. Also, additional redundant and diverse

equipment (normal control air, emergency bottled nitrogen, and the

valve hand wheels) is available and expected to remain functional to

ensure the ADVs accomplish their function following an accident. The

change does not create failure modes that could adversely impact

safety-related equipment. Therefore, the change will not create the

possibility of a malfunction of equipment important to safety

different than previously evaluated in the FSAR. Thus, the proposed

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

of accident from any accident previously evaluated.

C. The proposed amendment does not involve a significant

reduction in a margin of safety.

The TS currently allow two or more ADVs to be out of service for

24 hours, based on low probability of an event occurring during the

period which would require use of the ADVs, and based on

availability of the steam dump valves and the MSSVs [main steam

safety valves]. Providing a 72 hour completion time specifically for

loss of two ADV valves due to loss on one train of ACAS to the ADVs

does not significantly reduce the margin of safety since the

probability of an event occurring during the 72 hour period is still

small, and the capability exists to use the inoperable ADVs by

manually operating the valves using the valve hand wheels, or by

connecting the valve nitrogen bottle system, which was designed to

operate the valves upon loss of air. In addition, the MSSVs, and the

condenser steam dump valves would normally also be available. Thus,

the proposed change does not significantly reduce the margin of

safety.

Further, the NRC staff notes that the proposed change to the TS

action statement for two or more ADV lines inoperable to

[[Page 43214]]

require restoration of all but one of the four ADV lines, instead of

the previous requirement to restore only one ADV line to operable

status, is more restrictive and more conservative than the action

statement as currently written. The change also makes the action

statement consistent with the existing TS Bases in Section B 3.7.4,

Action B.1. Accordingly, the staff proposes to find that this

proposed change does not involve a significant increase in the

probability or consequences of an accident previously evaluated,

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

accident from any accident previously evaluated, and 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 and the staff's additional assessment as provided above, 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: Chattanooga-Hamilton County

Library, 1001 Broad Street, Chattanooga, TN 37402.

Attorney for licensee: General Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET l0H, Knoxville, Tennessee 37902.

NRC Project Director: Frederick J. Hebdon.

Tennessee Valley Authority, Docket No. 50-390 Watts Bar Nuclear Plant,

Unit 1, Rhea County, Tennessee

Date of amendment request: May 6, 1998.

Description of amendment request: The proposed amendment would

modify the Watts Bar Nuclear Plant (WBN) Technical Specifications (TSs)

by revising the allowed enrichment of fuel stored in the new fuel

storage racks from 4.3 to 5.0 weight percent uranium-235 (U-235). The

revision also places limitations on fuel storage locations that may be

utilized in the storage racks and provides additional limits on

k(effective) when flooded with unborated water.

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:

A. The proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

The proposed change to the allowed enrichment of new fuel stored

in the new fuel storage racks does not change the criticality

potential with the proposed fuel arrangement requirements for the

storage racks. The potential keff values are maintained

the same as the current TS requirements. In addition, the storage

racks are not modified and the processes for loading and unloading

fuel in these racks and the controls for these racks remain the same

except for the storage limitations dictated by the criticality

analysis. Additional controls are required with appropriate

verification to assure the fuel is stored within the analysis

assumptions. Handling procedures contain additional steps to

specifically verify prohibited cells remain empty after fuel

movement. This verification assures that the probability of a

criticality event is not increased by the enrichment change. Since

the keff limits and operating processes are unchanged by

the proposed revision, there is no increase in the probability of an

accident previously evaluated. Likewise, there is no impact to the

consequences of an accident or increase in offsite dose limits as a

result of the proposed TS change because the criticality

requirements are unchanged and plant equipment will be utilized and

operated without change considering the fuel storage location limits

imposed by this request.

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

new or different kind of accident from any accident previously

evaluated.

As stated above, the plant equipment and operating processes

will not be altered by the proposed TS change with the exception of

allowed fuel storage locations in the new fuel storage racks. The

limitations on acceptable fuel storage locations in the racks ensure

that the k(effective) limits are maintained at the same limits as

currently required. TVA has not postulated a criticality event at

WBN for the spent or new fuel storage locations because the design

of the associated storage racks, potential moderation, and TS

allowable fuel enrichments do not support the potential for this

condition. Therefore, this change does not create the potential for

a new accident from any previously analyzed.

C. The proposed amendment does not involve a significant

reduction in a margin of safety.

The proposed TS change maintains the existing requirements for

criticality by utilizing limited storage locations in the new fuel

pit storage racks. There is no change to operating practices

associated with the use and control of these racks except for the

storage limitations. For these reasons, there will be no reduction

in the margin [of] the safety as a result of implementing the

proposed TS change.

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: Chattanooga-Hamilton County

Library, 1001 Broad Street, Chattanooga, TN 37402.

Attorney for licensee: General Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET l0H, Knoxville, Tennessee 37902.

NRC Project Director: Frederick J. Hebdon.

The Cleveland Electric Illuminating Company, Centerior Service Company,

Duquesne Light Company, Ohio Edison Company, Pennsylvania Power

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

Plant, Unit 1, Lake County, Ohio

Date of amendment request: July 13, 1998.

Description of amendment request: The proposed license amendment

would revise Perry Nuclear Power Plant Technical Specification 3.4.4,

``Safety/Relief Valves (S/RVs),'' by increasing the present [plus or

minus] 1% tolerance on the safety mode lift setpoint for the safety/

relief valves to [plus or minus] 3%. This change would be performed in

accordance with General Electric Topical Report NEDC-31753P, ``BWROG

In-Service Pressure Relief Technical Specification Revision Licensing

Topical 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

identified.

The proposed change allows an increase in the as-found safety

relief valve (SRV) safety mode setpoint tolerance, determined by

test after the valves have been removed from service, from [plus or

minus] 1% to [plus or minus] 3%. The proposed change does not alter

the Technical Specification requirements on the nominal SRV safety

mode lift setpoints, the SRV relief mode setpoints, the required

frequency for the SRV lift setpoint tests, or the number of SRVs

required to be operable. This change does not involve physical

changes to the SRVs, nor does it change the operating

characteristics or safety function of the SRVs.

Consistent with current requirements, this change continues to

require that the SRVs be adjusted to within [plus or minus] 1% of

their nominal lift setpoints following testing. This change does not

change the behavior and operation of any SRV and therefore has no

significant impact to reactor operation. It also has no significant

impact on response to any perturbation of reactor operation

including transients and accidents previously analyzed in the

Updated Safety Analysis Report. In addition, this change does not

change SRV actuation. Therefore, this change will not increase the

probability of an accident previously evaluated.

Generic considerations related to the change in setpoint

tolerance were addressed

[[Page 43215]]

in NEDC-31753P, ``BWROG In-Service Pressure Relief Technical

Specification Revision Licensing Topical Report,'' and were reviewed

and approved by the NRC. The plant specific evaluations, required by

the NRC's Safety Evaluation for NEDC-31753P and performed to support

this proposed change, are contained in NEDC-32307P, ``Safety Review

for PNPP Safety/Relief Valve Setpoint Tolerance Relaxation/Out-of-

Service Analyses,'' dated May 1994. These analyses and evaluations

show that there is adequate margin to the design core thermal limits

and to the reactor vessel pressure limits using a [plus or minus] 3%

SRV setpoint tolerance. They also show that operation of the high

pressure injection systems will not be adversely affected; and the

containment response from a loss of coolant accident will be

acceptable.

(2) The proposed change would not create the possibility of a

new or different kind of accident from any accident previously

evaluated.

The proposed change to allow an increase in the SRV safety mode

setpoint tolerance from [plus or minus] 1% to [plus or minus] 3%

does not alter the nominal SRV lift setpoints or the number of SRVs

required to be operable. This change does not involve physical

changes to the SRVs, nor does it change the operating

characteristics or the safety function of the SRVs. The proposed

change does not involve a physical alteration of the plant. No new

or different equipment is being installed. The proposed change does

not impact core reactivity nor the manipulation of fuel bundles.

There is no alteration to the parameters within which the plant is

normally operated. As a result no new failure modes are being

introduced. There are no changes in the methods governing normal

plant operation, nor are the methods utilized to respond to plant

transients altered.

Therefore, the proposed change does not create the possibility

of a new or different kind of accident from any previously

evaluated.

(3) The proposed change will not involve a significant reduction

in the margin of safety.

The margin of safety is established through the design of the

plant structures, systems, and components, the parameters within

which the plant is operated, and the establishment of the setpoints

for the actuation of equipment relied upon to respond to an event.

The proposed change does not significantly impact the condition or

performance of structures, systems, and components relied upon for

accident mitigation. The proposed change does not significantly

impact any safety analysis assumptions or results.

Therefore, the proposed change does not involve a significant

reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Perry Public Library, 3753

Main Street, Perry, OH 44081.

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

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

NRC Project Director: Ronald R. Bellamy (Acting).

Previously Published Notices of Consideration of Issuance of

Amendments To Facility Operating Licenses, Proposed No Significant

Hazards Consideration Determination, and Opportunity for a Hearing

The following notices were previously published as separate

individual notices. The notice content was the same as above. They were

published as individual notices either because time did not allow the

Commission to wait for this biweekly notice or because the action

involved exigent circumstances. They are repeated here because the

biweekly notice lists all amendments issued or proposed to be issued

involving no significant hazards consideration.

For details, see the individual notice in the Federal Register on

the day and page cited. This notice does not extend the notice period

of the original notice.

Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee

Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina

Date of amendment request: July 8, 1998.

Description of amendment request: The proposed amendments would

allow temporary noncompliance with the Penetration Room Ventilation

System air flow surveillance requirements of Technical Specification

4.5.4.1.b.1 until modifications can be completed to support testing in

accordance with ANSI Standard N510-1975, as required by the Technical

Specifications.

Date of publication of individual notice in Federal Register: July

16, 1998 (63 FR 38433).

Expiration date of individual notice: August 17, 1998.

Local Public Document Room location: Oconee County Library, 501

West South Broad Street, Walhalla, South Carolina.

Florida Power Corporation, et al., Docket No. 50-302, Crystal River

Unit No. 3 Nuclear Generating Plant, Citrus County, Florida

Date of application for amendment: June 18, 1998.

Brief description of amendment: Amend the Crystal River Unit 3

(CR3) Improved Technical Specifications to allow operation with a

number of indications previously identified as tube end anomalies and

multiple tube end anomalies in the CR3 Once Through Steam Generator

tubes.

Date of publication of individual notice in the Federal Register:

June 30, 1998 (63 FR 35615).

Expiration date of individual notice: July 15, 1998.

Local Public Document Room location: Coastal Region Library, 8619

W. Crystal Street, Crystal River, Florida 32629.

Northern States Power Company, Docket No. 50-263, Monticello Nuclear

Generating Plant, Wright County, Minnesota

Date of amendment request: June 19, 1998 (supersedes April 11,

1997, application), as supplemented July 1, 1998, and information

provided in a letter of May 5, 1997.

Brief description of amendment request: The proposed amendment

would revise Section 3.6.C, Coolant Chemistry, and 3/4.17.B, Control

Room Emergency Filtration System, of the Technical Specifications (TS),

Appendix A of the Operating License for the Monticello Nuclear

Generating Plant. The changes were proposed to establish TS

requirements consistent with modified analysis inputs used for the

evaluation of the radiological consequences of the main steam line

break accident. This amendment request was originally noticed in the

Federal Register on May 6, 1998 (63 FR 25115). On June 19, 1998,

supplemented July 1, 1998, the licensee submitted an application that

superseded in its entirety the licensee's previous submittal dated

April 11, 1997.

Date of publication of individual notice in Federal Register: July

28, 1998 (63 FR 40321).

Expiration date of individual notice: August 27, 1998.

Local Public Document Room location: Minneapolis Public Library,

Technology and Science Department, 300 Nicollet Mall, Minneapolis,

Minnesota 55401.

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

Callaway County, Missouri

Date of application for amendment: February 24 1998, as

supplemented by letter dated May 27, 1998.

Brief description of amendment: The amendment would support a

modification to the Callaway Plant, Unit 1 to increase the storage

capacity of the spent fuel pool.

Date of individual notice in Federal Register: July 13, 1998 (63 FR

37598).

[[Page 43216]]

Expiration date of individual notice: August 12, 1998.

Local Public Document Room location: University of Missouri-

Columbia, Elmer Ellis Library, Columbia, Missouri 65201-5149.

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

Generating Station, Coffey County, Kansas

Date of amendment request: March 20, 1998, as supplemented by

letter dated May 28, 1998.

Brief description of amendment: The amendment would support a

modification to the Wolf Creek Nuclear Generating Station, Unit 1 to

increase the storage capacity of the spent fuel pool.

Date of individual notice in Federal Register: July 13, 1998 (63 FR

37601).

Expiration date of individual notice: August 12, 1998.

Local Public Document Room locations: Emporia State University,

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

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

66621.

Notice of Issuance of Amendments to Facility Operating Licenses

During the period since publication of the last biweekly notice,

the Commission has issued the following amendments. The Commission has

determined for each of these amendments that the application complies

with the standards and requirements of the Atomic Energy Act of 1954,

as amended (the Act), and the Commission's rules and regulations. The

Commission has made appropriate findings as required by the Act and the

Commission's rules and regulations in 10 CFR Chapter I, which are set

forth in the license amendment.

Notice of Consideration of Issuance of Amendment to Facility

Operating License, Proposed No Significant Hazards Consideration

Determination, and Opportunity for A Hearing in connection with these

actions was published in the Federal Register as indicated.

Unless otherwise indicated, the Commission has determined that

these amendments satisfy the criteria for categorical exclusion in

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

no environmental impact statement or environmental assessment need be

prepared for these amendments. If the Commission has prepared an

environmental assessment under the special circumstances provision in

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

it is so indicated.

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

applications for amendment, (2) the amendment, and (3) the Commission's

related letter, Safety Evaluation and/or Environmental Assessment as

indicated. All of these items are available for public inspection at

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

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

the particular facilities involved.

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

Station, Plymouth County, Massachusetts

Date of application for amendment: February 20, 1998.

Brief description of amendment: This amendment changed the Pilgrim

Nuclear Power Station Technical Specification (TS) 3/4.5.B and its

Bases to incorporate the ultimate heat sink (UHS) temperature of 75

deg.F, as required by Amendment No. 173. The introduction of a UHS

temperature restriction requires new specifications, actions, and

surveillances for the salt service water system. The amendment also

replaced existing specification 3.5.B ``Containment Cooling System''

with new Specification 3/4.5.B.1 ``Residual Heat Removal (RHR)

Suppression Pool Cooling'', 3/4.5.B.2 ``Residual Heat Removal (RHR)

Containment Spray'', 3/4.5.B.3 ``Reactor Building Closed Cooling Water

(RBCCW) System'', and 3/4.5.B.4 ``Salt Service Water (SSW) System and

Ultimate Heat Sink (UHS)''.

Date of issuance: July 28, 1998.

Effective date: July 28, 1998.

Amendment No.: 176.

Facility Operating License No. DPR-35: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: April 8, 1998 (63 FR

17221).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated July 28, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Plymouth Public Library, 11

North Street, Plymouth, Massachusetts 02360.

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

Station, Plymouth County, Massachusetts

Date of application for amendment: September 19, 1997, as

supplemented June 15, 1998.

Brief description of amendment: The amendment relocates the

Radioactive Effluent Technical Specifications and the Radiological

Environmental Monitoring Program to the Offsite Dose Calculation

Manual, in accordance with the recommendations of Generic Letter 89-01.

Changes are also being made to other sections of the Technical

Specifications to align them with NUREG-1433, to minimize changes when

converting to the Improved Standard Technical Specifications.

Date of issuance: July 31, 1998.

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

within 30 days.

Amendment No.: 177.

Facility Operating License No. DPR-35: Amendment revised the

Technical Specifications and the license.

Date of initial notice in Federal Register: February 25, 1998 (63

FR 9591).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated July 31, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Plymouth Public Library, 11

North Street, Plymouth, Massachusetts 02360.

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

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

Date of application for amendment: June 26, 1998, as supplemented

July 22, 1998.

Brief description of amendment: The amendment revises Technical

Specification (TS) 3.7.8, ``Ultimate Heat Sink (UHS),'' to permit an 8-

hour delay in the UHS temperature restoration period prior to entering

the plant shutdown required actions. This TS amendment is given as a

one-time amendment change effective until September 30, 1998, after

which the TS will revert back to the original TS provisions.

Date of issuance: July 29, 1998.

Effective date: July 29, 1998.

Amendment No.: 179.

Facility Operating License No. DPR-23. Amendment revised the

Technical Specifications.

Public comments requested as to proposed no significant hazards

consideration (NSHC): Yes (63 FR 36967 dated July 8, 1998). The notice

provided an opportunity to submit comments on the Commission's proposed

NSHC determination. No comments have been received. The notice also

provided for an opportunity to request a hearing by August 7, 1998, but

indicated that if the Commission makes a final NSHC determination, any

such hearing would take place after issuance of the amendment.

[[Page 43217]]

The Commission's related evaluation of the amendment, finding of

exigent circumstances, and final determination of NSHC are contained in

a Safety Evaluation dated July 29, 1998.

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: P. T. Kuo, Acting.

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

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

Date of application for amendments: March 30, 1998.

Brief description of amendments: The amendments will (1) restore

Custom Technical Specifications (CTS) and the associated license

conditions that had been replaced by Improved Technical Specifications

(ITS), (2) change certain management titles and responsibilities to

reflect the permanently shutdown condition of the plant, (3) allow use

of Certified Fuel Handlers in lieu of licensed operators, (4) modify

shift crew composition, and (5) eliminate verbiage that imples the

units are operational.

Date of Issuance: July 24, 1998.

Effective date: Immediately, to be implemented within 30 days.

Amendment Nos.: 179 & 166.

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

revised the Technical Specifications.

Date of initial notice in Federal Register: May 6, 1998 (63 FR

25105). The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated July 24, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Waukegan Public Library, 128

N. County Street, Waukegan, Illinois 60085.

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

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

Date of application for amendment: June 6, 1997, as supplemented

September 25, 1997.

Brief description of amendment: The amendment revises Technical

Specifications (TS) Table 4.1-2, Frequency for Sampling Tests, to

delete the requirement to sample the spray additive tank and delete the

requirement for a sodium hydroxide (NaOH) spray additive in TS Section

5.2.C.1.

Date of issuance: July 29, 1998.

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

30 days.

Amendment No.: 197.

Facility Operating License No. DPR-26: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: January 28, 1998 (63 FR

4310).

The September 25, 1997, letter provided clarifying information that

did not change the initial proposed no significant hazards

consideration.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated July 29, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: White Plains Public Library,

100 Martine Avenue, White Plains, New York 10610.

Duke Energy Corporation, Docket Nos. 50-369 and 50-370, McGuire Nuclear

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

Date of application for amendments: March 3, 1998, as supplemented

by letters dated April 24, May 7, and July 22, 1998.

Brief description of amendments: The amendments revise Figure 5.1-1

of the Technical Specifications (TS) to show the new location of the

meteorological tower. The meteorological tower will be relocated to a

new location to facilitate use of the current location as a

construction site. The proposed TS change does not change the related

TS Section 5.1.1.

Date of issuance: July 30, 1998.

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

30 days.

Amendment Nos.: Unit 1--179; Unit 2--161.

Facility Operating License Nos. NPF-9 and NPF-17: Amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: June 29, 1998 (63 FR

35293).

The July 22, 1998, submittal provided clarifying information that

did not change the scope of the March 3, 1998, application and the

initial proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated July 30, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: J. Murrey Atkins Library,

University of North Carolina at Charlotte, 9201 University City

Boulevard, Charlotte, North Carolina.

Duquesne Light Company, et al., Docket Nos. 50-334 and 50-412, Beaver

Valley Power Station, Unit Nos. 1 and 2, (BVPS-1 and BVPS-2)

Shippingport, Pennsylvania

Date of application for amendments: June 19, 1998, as supplemented

June 23, 1998.

Brief description of amendments: These amendments revise the BVPS-1

and BVPS-2 Technical Specifications (TSs) definitions of a channel

calibration to add two sentences stating that (1) the calibration of

instrument channels with resistance temperature detector or

thermocouple sensors may consist of an inplace qualitative assessment

of sensor behavior and normal calibration of the remaining adjustable

devices in the channel and (2) whenever a sensing element is replaced,

the next required channel calibration shall include an inplace cross

calibration that compares the other sensing elements with the recently

installed sensing element. This change makes the BVPS-1 and BVPS-2 TS

definition of channel calibration consistent with the definition of a

channel calibration contained in the NRC's improved Standard Technical

Specifications for Westinghouse Plants (NUREG-1431, Revision 1).

Date of Issuance: July 28, 1998.

Effective date: Both units, effective immediately, to be

implemented within 30 days.

Amendment Nos.: 216 and 93.

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

revised the Technical Specifications.

Date of initial notice in Federal Register: June 26, 1998 (63 FR

34939).

The June 23, 1998, letter provided minor editorial changes to the

TS pages that did not change the initial proposed no significant

hazards consideration determination or expand the amendment request

beyond the scope of the June 26, 1998 Federal Register notice.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated July 28, 1998.

No significant hazards consideration comments received: No.

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

663 Franklin Avenue, Aliquippa, PA 15001.

Florida Power Corporation, et al., Docket No. 50-302, Crystal River

Unit No. 3 Nuclear Generating Plant, Citrus County, Florida

Date of application for amendment: March 20, 1998, and supplemented

May 22, 1998.

[[Page 43218]]

Brief description of amendment: The amendment proposed to revise

Improved Technical Specification Safety Limits and Administrative

Controls to replace the titles of the Senior Vice President, Nuclear

Operations and the Vice President, Nuclear Production with the position

of Chief Nuclear Officer.

Date of issuance: July 20, 1998.

Effective date: July 20, 1998.

Amendment No.: 168.

Facility Operating License No. DPR-72: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: May 6, 1998 (63 FR

25109).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated July 20, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Coastal Region Library, 8619

W. Crystal Street, Crystal River, Florida 34428.

Florida Power and Light Company, et al., Docket No. 50-389, St. Lucie

Plant, Unit No. 2, St. Lucie County, Florida

Date of application for amendment: December 29, 1997, as

supplemented by June 15, 1998.

Brief description of amendment: The amendment will modify the

Technical Specifications for selected cycle-specific reactor physics

parameters to refer to the St. Lucie Unit 2 Core Operating Limits

Report for limiting values.

Date of Issuance: July 24, 1998.

Effective Date: July 24, 1998.

Amendment No.: 92.

Facility Operating License No. NPF-16: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: February 11, 1998 (63

FR 6985).

The June 15, 1998, supplement provided clarifying information that

did not change the scope of the December 29, 1997 application and the

initial proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated July 24, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Indian River Community College

Library, 3209 Virginia Avenue, Fort Pierce, Florida 34981-5596.

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

Station, Nemaha County, Nebraska

Date of amendment request: February 10, 1997, as supplemented

December 26, 1997, and July 16, and July 28, 1998.

Brief description of amendment: The amendment revised the Technical

Specifications to reflect the adoption of the BWR Owner's Group Long-

Term Solution Stability System Option 1-D in addressing reactor

operation in or near a region of potential thermal hydraulic

instability.

Date of issuance: July 29, 1998.

Effective date: July 29, 1998, to be implemented within 30 days.

Amendment No.: 177.

Facility Operating License No. DPR-46: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: March 26, 1997 (62 FR

14462).

The December 26, 1997, July 16, and July 28, 1998, submittals

provided clarifying information and an administrative change that did

not alter the initial proposed no significant hazards consideration

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

contained in a Safety Evaluation dated July 29, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Auburn Memorial Library, 1810

Courthouse Avenue, Auburn, NE 68305.

Northern States Power Company, Docket Nos. 50-282 and 50-306, Prairie

Island Nuclear Generating Plant, Units 1 and 2, Goodhue County,

Minnesota

Date of application for amendments: January 15, 1998, as

supplemented May 29, 1998.

Brief description of amendments: The amendment allows a reduction

in the required number of incore instrumentation detectors for the

remainder of Unit 1, Cycle 19 operation.

Date of issuance: July 28, 1998.

Effective date: July 28, 1998, with full implementation within 30

days.

Amendment Nos.: 136.

Facility Operating License Nos. DPR-42 and DPR-60. Amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: January 30, 1998 (63 FR

4676) The May 29, 1998, supplement provided clarifying information

within the scope of the Federal Register notice and did not change the

staff's initial proposed no significant hazards considerations

determination.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated July 28, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Minneapolis Public Library,

Technology and Science Department, 300 Nicollet Mall, Minneapolis,

Minnesota 55401.

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

FitzPatrick Nuclear Power Plant, Oswego County, New York

Date of application for amendment: December 12, 1997.

Brief description of amendment: The amendment revises the working

hours for operating personnel to allow 8- to 12-hour work days, nominal

40-hour weeks. In addition, associated changes are being made to

surveillance intervals to maintain the same frequency.

Date of issuance: July 24, 1998.

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

30 days.

Amendment No.: 244.

Facility Operating License No. DPR-59: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: January 28, 1998 (63 FR

4321).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated July 24, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Reference and Documents

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

York 13126.

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

Nuclear Power Plant, Wayne County, New York

Date of application for amendment: March 31, 1997, as supplemented

June 18, 1997, October 10, 1997, October 20, 1997, November 11, 1997,

December 22, 1997, January 15, 1998, January 27, 1998, March 30, 1998,

April 23, 1998, April 27, 1998, May 8, 1998, and May 22, 1998.

Brief description of amendment: This amendment changes the

Technical Specifications to accommodate the modification of the spent

fuel pool by replacing the three Region 1 rack modules with seven new

borated stainless steel rack modules scheduled for implementation in

1998. Six new peripheral modules would be added at some future date.

Two of the seven new modules planned to be installed in 1998 are to be

designated as part of Region 2, effectively increasing the Region 2

area. The other five new modules compose Region 1, resulting in a total

of 294 storage positions in Region 1. Region 2, with 1075 storage

positions, consists of three rack types, Type 1, Type 2, and Type 4.

Type 1 cells are the Boraflex cells that form Region 2 for the existing

license. Two racks of Type 2 cells, containing borated stainless steel

(BSS) absorber plates are be added to increase

[[Page 43219]]

the storage capacity of Region 2. In addition, the capacity of Region 2

could be increased in the future by the addition of Type 4 racks, which

also contain BSS absorber plates. The amendment increases the boron

concentration from 300 ppm to 2300 ppm.

Date of issuance: July 30, 1998.

Effective date: July 30, 1998.

Amendment No.: 72.

Facility Operating License No. DPR-18: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: June 30, 1998 (63 FR

35617).

The May 8 and 22, 1998, letters provided clarifying information

that did not change the proposed no significant hazards consideration.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated July 30, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Rochester Public Library, 115

South Avenue, Rochester, New York 14610.

Southern Nuclear Power Company, Inc., et al. Docket Nos. 50-424 and 50-

425, Vogtle Electric Generating Plant (VEGP), Units 1 and 2, Burke

County, Georgia

Date of application for amendments: May 8, 1998.

Brief description of amendments: The amendments revise VEGP

Technical Specification 5.5.7, ``Reactor Coolant Pump Flywheel

Inspection Program,'' to provide an exception to the examination

requirements of Regulatory Position C.4.b of Regulatory Guide 1.14,

Revision 1, dated August 1975.

Date of issuance: July 21, 1998.

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

30 days.

Amendment Nos.: Unit 1--103; Unit 2--81.

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

revised the Technical Specifications.

Date of initial notice in Federal Register: June 17, 1998 (63 FR

33108).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated July 21, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Burke County Library, 412

Fourth Street, Waynesboro, Georgia.

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

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

Date of application for amendments: February 25, 1998 (TS 97-06).

Brief description of amendments: The amendments change the

Technical Specifications (TS) by revising the surveillance requirements

for the emergency diesel generators.

Date of issuance: July 22, 1998.

Effective date: To be implemented no later than 45 days after

issuance.

Amendment Nos.: Unit 1--234; Unit 2--224.

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

revise the TS.

Date of initial notice in Federal Register: April 8, 1998 (63 FR

17235).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated July 22, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Chattanooga-Hamilton County

Library, 1001 Broad Street, Chattanooga, Tennessee 37402.

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

Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc

County, Wisconsin

Date of application for amendments: May 2, 1995, as supplemented

October 12, 1995, March 26, 1996, December 15, 1997, and May 27, 1998

(TSCR 172).

Brief description of amendments: These amendments revise the

Technical Specifications (TS) Table 15.4.1-1, ``Minimum Frequencies For

Checks, Calibrations, and Tests Of Instrument Channels,'' to change the

test frequency of the containment high range radiation monitor, revise

note 7, and revise item 36 to clarify which monitors in the radiation

monitoring system support current TS or meet the requirements of 10 CFR

50.36. In addition several administrative changes to referenced TS

sections and plant system titles were made to correct omissions from

previous amendments.

Date of issuance: July 17, 1998.

Effective date: July 17, 1998. The TS are to be implemented within

45 days from the date of issuance. Implementation shall also include

relocation of certain TS requirements to licensee-controlled documents,

as described in the licensee's application dated May 2, 1995, as

supplemented October 12, 1995, March 26, 1996, December 15, 1997, and

May 27, 1998, and evaluated in the staff's safety evaluation attached

to these amendments.

Amendment Nos.: 185 and 189.

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

revised the Technical Specifications.

Date of initial notice in Federal Register: May 6, 1998 (63 FR

25122). The May 27, 1998, submittal provided additional clarifying

information and updated TS pages. This information was within the scope

of the original Federal Register notice and did not change the staff's

initial no significant hazards considerations determination.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated July 17, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: The Lester Public Library,

1001 Adams Street, Two Rivers, Wisconsin 54241.

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

Beach Nuclear Plant, Unit 2, Town of Two Creeks, Manitowoc County,

Wisconsin

Date of application for amendments: May 15, 1998 (TSCR 205, NPL-98-

0303).

Brief description of amendment: This amendment revises the schedule

for implementing the boron concentration changes from refueling outage

24 to refueling outage 23 for the planned conversion of Unit 2 to 18-

month fuel cycles.

Date of issuance: July 21, 1998.

Effective date: July 21, 1998, with full implementation within 45

days.

Amendment No.: 190.

Facility Operating License No. DPR-27: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: June 17, 1998 (63 FR

33111).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated July 21, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: The Lester Public Library,

1001 Adams Street, Two Rivers, Wisconsin 54241.

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

Generating Station, Coffey County, Kansas

Date of amendment request: July 17, 1998.

Brief description of amendment: The amendment revised Technical

Specification 3/4.7.5, Ultimate Heat Sink, by adding a new Action

Statement to be used in the event that plant inlet water temperature

exceeds 90 degrees F.

Date of issuance: July 18, 1998.

Effective date: July 18, 1998.

Amendment No.: 118.

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

Technical Specifications.

[[Page 43220]]

Public comments requested as to proposed no significant hazards

consideration: No.

The Commission's related evaluation of the amendment, finding of

emergency circumstances, and final determination of no significant

hazards consideration are contained in a Safety Evaluation dated July

18, 1998.

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

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

Local Public Document Room locations: Emporia State University,

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

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

66621.

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

Generating Station, Coffey County, Kansas

Date of amendment request: March 24, 1995, as supplemented by

letters dated July 26, 1995, and September 5, 1996.

Brief description of amendment: The amendment adds a new action

statement to Technical Specification (TS) 3.5.1 which provides a 72-

hour allowed outage time (AOT) for one accumulator to be inoperable

because its boron concentration did not meet the 2300-2500 parts per

million band. In addition, TS surveillance requirements are changed to

incorporate the guidance of Generic Letter 93-05, ``Line-Item Technical

Specifications Improvements to Reduce Surveillance Requirements for

Testing During Operation'' that is applicable to the accumulators, and

the TS Bases section for TS 3/4.5.1 is revised to reflect the changes

described above. Instrumentation surveillance requirements associated

with the accumulator are being relocated from the technical

specifications to Chapter 16 of the Updated Safety Analysis Report.

Date of issuance: July 21, 1998.

Effective date: July 21, 1998, to be implemented within 30 days

from the date of issuance.

Amendment No.: 119.

Facility Operating License N

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