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 Pub. L. 97-415, the U.S. Nuclear Regulatory Commission

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

notice. Pub. L. 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 September 14, 1998, through September 25,

1998. The last biweekly notice was published on September 23, 1998 (63

FR 50932).

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.

[[Page 53944]]

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 November 6, 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 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.

[[Page 53945]]

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: October 28, 1997, as supplemented March

26, May 20, July 29, and August 13, 1998.

Description of amendment request: The proposed amendments would

revise the current Technical Specifications (CTS) of each unit to

conform with NUREG-1430, ``Standard Technical Specifications--Babcock

and Wilcox Plants.'' The Commission had previously issued a Notice of

Consideration of Issuance of Amendments published in the Federal

Register on December 5, 1997 (62 FR 64405), covering all of the

proposed Improved Technical Specification (ITS) changes that were

within the scope of NUREG-1430 for the Oconee Nuclear Station. However,

the submittals also contained proposed changes that are beyond the

scope of NUREG-1430, which were not included in the staff's December 5,

1997, notice. The following descriptions and proposed no significant

hazards analyses cover only the beyond-scope changes. Associated with

each proposed change are administrative/editorial changes such that the

new or revised requirements would fit into the format of NUREG-1430.

Some changes are ``Less Restrictive'' (meaning that the new

requirements being incorporated into the ITS are less restrictive than

the CTS requirements) and some are ``More Restrictive.'' The basis for

the no significant hazards determination is identical for all of the

more restrictive items and is presented at the end of the following

list of more restrictive beyond-scope items:

A. Certain NUREG and CTS Sections 3.1.3.5, 3.5.2.4.a, 3.5.2.5.b,

3.5.2.5.c, and 3.5.2.6, specify that they are applicable ``except

during Mode 1 physics testing.'' The exception would not be included in

the ITS and, therefore, the Mode 1 requirement would be applicable

during the tests. The proposed change is more conservative since no

exceptions would be allowed for physics tests conducted in Mode 1.

B. CTS 3.1.3.2 requires reactor coolant temperature to be greater

than the criticality values of specified heatup limitation curves. This

requirement would not be retained in the ITS. ITS 3.1.8, Limiting

Condition for Operation (LCO) Part e, would be added to provide a

restriction for loop average temperature to be greater than or equal to

520 deg.F when performing physics tests in Mode 2. ITS LCO 3.1.8 would

permit suspending the requirements of ITS LCO 3.4.2, ``RCS (reactor

coolant system) Minimum Temperature for Criticality,'' during physics

tests initiated in Mode 2. Associated Actions and a surveillance

requirement (SR) would be added to provide an appropriate required

action when outside the limit and to verify operation within the limit

periodically.

C. CTS Table 3.5.1-1 presently requires that the operator place the

plant in hot shutdown (ITS equivalent of Mode 3) within 12 hours when

the minimum channels Operable requirement is not met. The proposed

change to the ITS would provide an equivalent requirement and add a

requirement to open all control rod drive (CRD) trip breakers within 12

hours. ITS 3.3.3 Action B, and ITS 3.3.4 Action D, would be added to

require that the unit be in Mode 3 in 12 hours with all CRD trip

breakers open or that power be removed from all CRD trip breakers when

the required action and associated completion time is not met in Mode

1, 2, or 3. For ITS 3.3.3, Action B would also apply when two or more

reactor trip modules are inoperable in Mode 1, 2, or 3. The CTS

presently requires entry into TS 3.0, which requires that the reactor

be in hot shutdown (equivalent to ITS Mode 3) in 12 hours.

D. Note c would be added to ITS Table 3.3.8-1, Post Accident

Monitoring Instrumentation, and referenced to Item No. 8, Containment

Isolation Valve Position, to specify that position indication

requirements apply only to the Containment Isolation Valves that are

electrically controlled.

E. The applicability of Table 3.5.1-1 would be expanded to require

wide range instruments to be operable in Mode 2, plus Modes 3, 4, and

5, with any control rod drive trip breaker in the closed position and

the control rod drive system capable of rod withdrawal. In addition, a

Note would define the upper limit of the applicable Modes for the

required wide range instrument channels as being 10 percent indicated

neutron power.

F. The applicability of ITS 3.3.14 would be expanded to include

Mode 4 when the steam generator is relied upon for heat removal, which

then would be consistent with the applicability of ITS LCO 3.7.5 for

the emergency feedwater (EFW) system. ITS Specifications 3.3.14 and

3.3.15 would be added to address EFW system initiation circuitry and

main steamline break and main feedwater isolation instrumentation

separately. The specification titles, LCOs, actions, and SRs would be

modified to reflect Oconee-specific terminology and design

requirements. Where appropriate, ITS-required actions would be based on

similar NUREG-required actions. EFW pump initiation circuitry operable

requirement would be changed from 250 deg.F to greater than or equal

to 246 deg.F.

G. ITS LCO 3.4.1, Departure from Nucleate Boiling Ratio (DNBR)

Limits, are specified in the core operating limits report rather than

in the LCO and SRs since they are subject to change with fuel cycle

designs. The ITS LCO 3.4.1 actions would require restoring DNBR

parameters to within limits within 2 hours or exiting the applicability

for the specification within 12 additional hours. ITS SR 3.4.1.1, SR

3.4.1.2, and SR 3.4.1.3 would require verification that each DNBR

parameter is within the limit at a 12-hour frequency. ITS SR 3.4.1.4

would require verification by measurement that total RCS flow is within

limit at an 18-month frequency. Specification 3.4.1 would ensure that

limits on RCS pressure, temperature, and flow rate are met to ensure

that the core operates within the limits assumed for the plant safety

analyses. These changes are more restrictive.

H. The NUREG allowed time to complete the SR after addition to core

flood tank (CFT) of 6 hours would be changed to 12 hours. ITS SR

3.5.1.4 would require CFT boron concentration be sampled every 31 days

or once within 12 hours after each solution volume increase greater

than or equal to 80 gallons that is not the result of addition from a

borated water source that meets CFT boron concentration requirements.

Since the CTS does not specify the time limit following addition, the

proposed ITS change is a more restrictive limit.

I. ITS 3.5.3 LCO Note 3 would be added to explicitly require that

the low pressure injection (LPI) discharge header crossover valves be

operable and capable of being opened manually when in Modes 1, 2, and

3. ITS 3.5.3 Action B would require that the LPI discharge header

crossover valves be restored to operable status within 72 hours of

being discovered incapable of being manually opened when in Modes 1, 2,

and 3. ITS 3.5.3 Action D would require LCO 3.0.3 be entered

immediately when one LPI train is inoperable in Modes 1, 2, and 3

concurrent with discovery that the LPI discharge header crossover

valves are incapable of being opened manually in Modes 1, 2, and 3.

J. ITS 3.5.3 would require the LPI system to be operable in Modes

1, 2, 3, and 4. LCO Note 1 would be added to specify that only one LPI

train is required to be operable in Mode 4. LCO Note 2 would be added

to allow an LPI train to be considered operable during

[[Page 53946]]

alignment, when aligned, or when operating if capable of being manually

realigned to the LPI mode of operation. Action E would be added to

require action be initiated immediately to restore the required LPI

train to operable status and to require the reactor to be placed in

Mode 5 within 24 hours when the required LPI train cannot be restored

to OPERABLE status (provided a decay heat removal loop is available).

K. SR 3.9.4.1 would be modified to eliminate verification of a

specific decay heat removal flow rate to verification every 12 hours

that one decay heat removal loop is in operation.

L. Main feeder bus monitoring panel requirements and allowed outage

time would be added to the ITS.

M. TS Section 3.7 would be revised to include the actual trip

setpoint and/or allowable values for the loss of power sensing relays.

N. Battery performance discharge testing as related to battery

operability would be added.

O. Battery charger testing, cell-to-cell resistance measurements,

and battery discharge and overcharge conditions, surveillances would be

added to ITS Section 3.8.

P. High Pressure Injection System discharge pressure allowable

value in ITS Table 3.3.5-1 would be changed from 1500 pounds per square

inch gauge (psig) to 1590 psig.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration for the More Restrictive Items listed above, as follows:

In accordance with the criteria set forth in 10 CFR 50.92, Duke

Energy has evaluated these proposed Technical Specification changes

and determined that they do not represent a significant hazards

consideration. The following is provided in support of this

consideration.

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The proposed changes provide more stringent requirements than

previously existed in the Technical Specifications. These more

stringent requirements do not result in operation that will increase

the probability of initiating an analyzed event. If anything the new

requirements may decrease the probability or consequences of an

analyzed event by incorporating the more restrictive changes. The

changes do not alter assumptions relative to mitigation of an

accident or transient event. The more restrictive requirements

continue to ensure process variables, structures, systems, and

components are maintained consistent with the safety analyses and

licensing basis. Therefore, the changes do 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 the accidents previously evaluated?

The proposed changes provide more stringent requirements than

previously existed in the Technical Specifications. The changes do

not alter the plant configuration (no new or different type of

equipment will be installed) or make changes in the methods

governing normal plant operation. The changes do impose different

requirements. However, these changes are consistent with the

assumptions in the safety analyses and licensing basis. Therefore,

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

of accident from any accident previously evaluated.

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

of safety?

The proposed changes provide more stringent requirements than

previously existed in the Technical Specifications. Adding more

restrictive requirements either increases or has no impact on the

margin of safety. The changes, by definition, provide additional

restrictions to enhance plant safety. The changes maintain

requirements within the safety analyses and licensing basis. As

such, no question of safety is involved. Therefore, the changes do

not involve a significant reduction in a margin of safety.

For the less restrictive beyond-scope items, the basis for the no

significant hazards consideration is unique for each item. The beyond-

scope item and the licensee's basis supporting its determination that

the proposed changes do not represent a significant hazards

consideration follow:

A. A proposed change to the Note for ITS SR 3.1.4.3 would provide

the additional flexibility for testing control rod drop times with

reactor coolant flow conditions other than full flow, but with at least

one reactor coolant pump (RCP) pump running. This would ensure that the

testing is bounding by restricting operation of the unit to the RCP

combination used during control rod drop testing and represents

adoption of the NUREG rather than the CTS.

Basis for proposed no significant hazards consideration

determination:

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The control rods are used to support mitigation of the

consequences of an accident; however, the control rod drop time

variations are not considered the initiator of any previously

analyzed accident. As such the proposed change in the method of

performing the control rod drop time testing will not increase the

probability of any accident previously evaluated. The proposed

changes allow for testing the control rod drop times with less than

a full complement of reactor coolant pumps operating. However, the

operation of the plant is restricted to the pump combinations

providing maximum flow less than or equal to the pump flow used for

the testing. Therefore, the drop times verified during testing will

remain valid for mitigating the consequences of any accident

previously evaluated. Therefore, this change does not involve an

increase in the consequences of any accident previously evaluated.

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

kind of accident from the accidents previously evaluated?

The proposed change does not necessitate a physical alteration

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

installed) or changes in parameters governing normal plant

operation. The proposed change will continue to ensure that the

control rods are available for insertion of reactivity in the time

frames consistent with the safety analysis. Thus, this change does

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

from any accident previously evaluated.

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

of safety?

The margin of safety provided in the acceptable control rod drop

times continues to be provided since these drop times have not been

changed. The surveillance methodology is revised to allow testing

with one, two, or three pumps operating. However, the operation of

the plant is restricted to the reactor coolant pump combinations

which maintain the margin of safety, i.e., those pump combinations

providing maximum flow less than or equal to the pump flow used for

the testing. Therefore, this change does not involve a significant

reduction in a margin of safety.

B. Required Action B.2.2 of ITS 3.3.11, 12, and 13, would be added

to provide the option of closing the main feedwater control valves

(MFCVs) and startup feedwater control valves (SFCVs) in lieu of

reducing main steam header pressure to less than 700 psig.

Applicability would be changed to Modes 1 and 2, plus Mode 3 when the

main steam header pressure is greater than 700 psig except when all

MFCVs and SFCVs are closed.

Basis for proposed no significant hazards consideration

determination:

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The MSLB (main steamline break) and MFW (main feedwater)

Isolation circuitry is not an initiator of analyzed events.

Therefore, the probability of an accident is independent of the

status of the MSLB and MFW Isolation circuitry. As such the proposed

change does not involve a significant increase in the probability of

an accident previously evaluated. The proposed change eliminates the

requirement for MSLB and MFW Isolation circuitry OPERABILITY when

all the MFCVs and SFCVs are closed. When the MFCVs and SFCVs are

closed the MSLB and MFW Isolation circuitry has no safety function

since its function is to close the MFCVs and SFCVs when conditions

indicate [an] MSLB. Therefore, the change does not involve a

significant increase in the

[[Page 53947]]

consequences of an accident previously evaluated.

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

kind of accident from the accidents previously evaluated?

The proposed change does not necessitate a physical alteration

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

installed) or changes in parameters governing normal plant

operation. Thus, this change does not create the possibility of a

new or different kind of accident from any accident previously

evaluated.

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

of safety?

Since MSLB and MFW Isolation circuitry requirements continue to

require OPERABILITY when the reactor is in a condition that requires

their function, the proposed change does not involve a significant

reduction in a margin of safety.

C. ITS 3.3.15 Action A.1 would be added to allow 1 hour to declare

the turbine stop valves (TSVs) inoperable prior to requiring that the

unit shut down when one or more TSV closure channels is inoperable. ITS

Specifications 3.3.14 and 3.3.15 would be added to address the

emergency feedwater system initiation circuitry and main steamline

break and main feedwater isolation instrumentation separately. The

NUREG specification combines the emergency feedwater system initiation,

main steamline isolation, and main feedwater isolation functions into

one specification. The specification titles, LCOs, actions, and SRs

would be modified to reflect Oconee-specific terminology and design

requirements. Where appropriate, ITS-required actions would be based on

similar NUREG-required actions.

Basis for proposed no significant hazards consideration

determination:

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

This change establishes a 1 hour Completion Time during which

the unit may continue operation with MSLB and MFW Isolation

instrumentation inoperable. This change provides an opportunity to

repair the inoperable instrumentation channel(s) prior to declaring

the equipment supported by it inoperable. The addition of this

allowed condition with a short Completion Time does not result in

any hardware changes. The allowed condition also does not

significantly increase the probability of occurrence for initiation

of any analyzed event since the function of the equipment does not

change (and therefore any initiation scenarios are not changed).

Further, the consequences of an accident are the same during the

additional one hour time period allowed for instrument channel

restoration as it is during the time period currently allowed for

restoring TSVs to OPERABLE status. Therefore, the change does not

significantly increase the probability of occurrence of an accident

previously evaluated.

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

kind of accident from the accidents previously evaluated?

The change does not necessitate a physical alteration of the

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

changes in parameters governing normal plant operation. The change

continues to ensure prompt restoration of compliance with the

limiting condition for operation, or prompt and appropriate

compensatory actions are taken. Thus, this change does not create

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

accident previously evaluated.

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

of safety?

Prompt and appropriate Required Actions have been determined

based on the safety analysis functions to be maintained. The allowed

condition has been determined appropriate based on a combination of

the time required to perform the action, the relative importance of

the function or parameter to be restored, and engineering judgment.

Therefore, this new allowed condition does not involve a significant

reduction in the margin of safety.

D. CTS 3.8.10 and 4.4.4.5 frequency would be changed from ``* * *

immediately prior to refueling operation'' to ``Once each refueling

outage prior to CORE ALTERATIONS or movement of irradiated fuel

assemblies within containment'' in ITS SR 3.3.16.2 for testing

frequency of the radiation monitor associated with the purge system

valve isolation and ITS SR 3.9.3.2 for testing isolation function of

the reactor building purge supply and exhaust valves.

Basis for proposed no significant hazards consideration

determination:

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The proposed change does not involve any physical alteration of

plant systems, structures or components, changes in parameters

governing normal plant operation, or methods of operation. The

isolation function of the radiation monitor associated with the

purge system valves is not assumed to be an initiator of any

analyzed event. As a result, the probability of an accident

occurring is independent of the status of testing the isolation

function of the radiation monitor associated with the purge system

valves. This change eliminates the requirement for testing of this

isolation function immediately prior to refueling operations. The

change continues to require the isolation function to be OPERABLE

and continues to ensure that this function is verified within a

reasonable interval prior to irradiated fuel assembly handling

within containment. This provides reasonable assurance the isolation

function of the radiation monitor associated with the purge system

valves remains OPERABLE. Therefore the consequence of an accident

previously evaluated are not significantly increased.

The proposed change does not involve any physical alteration of

plant systems, structures or components, changes in parameters

governing normal plant operation, or methods of operation. The

isolation function of the Reactor Building Purge supply and exhaust

valves is not assumed to be an initiator of any analyzed event. As a

result, the probability of an accident occurring is independent of

the status of testing the isolation function of the Reactor Building

Purge supply and exhaust valves. This change eliminates the

requirement for testing of the isolation function of the Reactor

Building Purge supply and exhaust valves immediately prior to

refueling operations. The change continues to require the isolation

function of the Reactor Building Purge supply and exhaust valves

train to be OPERABLE and continues to ensure that this function is

verified within a reasonable interval prior to irradiated fuel

assembly handling within containment. This continues to provide

reasonable assurance the isolation function of the Reactor Building

Purge supply and exhaust valves remains OPERABLE. Therefore the

consequence of an accident previously evaluated are not

significantly increased.

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

kind of accident from the accidents previously evaluated?

The proposed change does not necessitate a physical alteration

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

installed) or changes in parameters governing normal plant

operation. The proposed change will still require the isolation

function of the radiation monitor associated with the purge system

valves be OPERABLE. Thus, this change does not create the

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

previously evaluated.

The proposed change does not necessitate a physical alteration

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

installed) or changes in parameters governing normal plant

operation. The proposed change will still require the isolation

function of the Reactor Building Purge supply and exhaust valves be

OPERABLE. Thus, this change does not create the possibility of a new

or different kind of accident from any accident previously

evaluated.

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

of safety?

The isolation function of the radiation monitor associated with

the purge system valves is still required to be OPERABLE. This

change continues to ensure that this function is verified within a

reasonable interval prior to irradiated fuel assembly handling

within containment. Therefore the margin of safety has not been

significantly reduced.

The isolation function of the Reactor Building Purge supply and

exhaust valves is still required to be OPERABLE. This change

continues to ensure that this function is verified within a

reasonable interval prior to irradiated fuel assembly handling

within containment. Therefore the margin of safety has not been

significantly reduced.

E. CTS 3.7.6 and 3.7.7 both require an inoperable voltage sensing

relay to be restored within 72 hours. ITS 3.3.19 Required Action A.1

and ITS 3.3.20 Required Action A.1 would be

[[Page 53948]]

incorporated to require that the inoperable channel be placed in trip

within 72 hours. This change allows operation to continue indefinitely

when the channel is placed in trip and continues to allow 72 hours to

restore an inoperable channel that cannot be placed in trip.

Basis for proposed no significant hazards consideration

determination:

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

This change allows indefinite continued operation with one

voltage sensing channel inoperable, provided the inoperable voltage

sensing channel is placed in trip within 72 hours. This action

leaves the system in a one-out-of-two condition for actuation. Thus,

if another channel were to fail, the DGVP (degraded grid voltage

protection) instrumentation can still perform its function. This

change does not significantly increase the probability of occurrence

for initiation of any analyzed event since the function of the DGVP

instrumentation does not change (and therefore any initiation

scenarios are not changed). Also, the change does not change the

assumed response of the equipment in performing its specified

function from that originally considered. Therefore, the changes do

not significantly increase the consequences of an accident.

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

kind of accident from the accidents previously evaluated?

The proposed change does not necessitate a physical alteration

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

installed) or changes in parameters governing normal plant

operation. The change ensures proper availability for the required

DGVP function. Thus, this change does not create the possibility of

a new or different kind of accident from any accident previously

evaluated.

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

of safety? This change to the DGVP instrumentation requirements does

not involve a change in setpoints and cannot affect any margin of

safety associated with the response to a design basis accident. The

change does not prevent the DGVP instrumentation from performing

their function since the action places the DGVP instrumentation in a

one-out-of-two condition for actuation versus the normal two-out-of-

three logic. Thus, if another channel were to fail, the DGVP

instrumentation could still perform its initiation functions.

Therefore, this change to allow the DGVP initiation functions to

operate indefinitely with one required DGVP instrument channel

inoperable provided the channel is placed in the tripped condition

within 72 hours, is not considered to involve a significant

reduction in the margin of safety.

F. CTS Table 4.1-3 requires that CFT boron concentration be sampled

monthly and after each makeup. ITS SR 3.5.1.4 requires it be sampled

every 31 days and once within 12 hours after each solution increase

greater than or equal to 80 gallons that is not the result of addition

from a borated water source that meets CFT boron concentration

requirements. Therefore, the ITS frequency is less restrictive than

current requirements because sampling will be required once within 12

hours following the volume increase and source requirement. Also, the

source of makeup would be changed from the ``borated water storage

tank'' to ``a source that meets CFT boron concentration requirements.''

Basis for proposed no significant hazards consideration

determination:

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

A less frequent performance of a Surveillance Requirement does

not result in any hardware changes. The Frequency of performance

also does not significantly increase the probability of occurrence

for initiation of any analyzed event since the function of the

equipment does not change (and therefore any initiation scenarios

are not changed) and the proposed Frequency has been determined to

be adequate to demonstrate the tank inventory is within the required

parameter limits. Further, the Frequency of performance of a

surveillance does not significantly increase the consequences of an

accident because a change in Frequency does not change the assumed

response of the equipment in performing its specified mitigation

functions from that considered with the original Frequency. The core

flood tank boron concentration change resulting from volume addition

from a source of known concentration is a readily calculated

quantity and hence, a sample and analysis is not required to be

assured of adequate boron concentration. Therefore, this change does

not involve a significant increase in the probability or

consequences of any accident previously evaluated.

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

kind of accident from the accidents previously evaluated?

The proposed change does not necessitate a physical alteration

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

installed) or changes in parameters governing normal plant

operation. The proposed change will still ensure proper

surveillances are required for equipment considered in the safety

analysis. Thus, this change does not create the possibility of a new

or different kind of accident from any accident previously

evaluated.

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

of safety?

The proposed change continues to provide assurance of acceptable

boron concentration since addition from a source of known

concentration results in a readily identifiable resulting

concentration. Therefore, a change in the Surveillance Frequency

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

G. The proposed change would specify actions to be taken for

Borated Water Storage Tank (BWST) level, boron concentration, or

temperature not being within specifications. Proposed ITS 3.5.4

Required Action C.1 would allow 12 hours to reach Mode 3 (i.e., an

additional 6 hours over what is currently allowed by CTS 3.2.2) under

such conditions.

Basis for proposed no significant hazards consideration

determination:

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The proposed change does not involve any physical alteration of

plant systems, structures or components, changes in parameters

governing normal plant operation, or methods of operation. The time

to be in MODE 3 is not assumed to be the initiator of any analyzed

events. As a result, the probability of an analyzed event is

independent of the time permitted to be in MODE 3. The consequences

of an accident occurring during the 12 hours permitted to be in MODE

3 are no greater than the consequences of an accident occurring

during the 6 hours currently permitted to place the unit in Hot

Shutdown. Therefore, the probability and consequence of an accident

previously evaluated are not significantly increased.

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

kind of accident from the accidents previously evaluated?

The proposed change does not necessitate a physical alteration

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

installed) or changes in parameters governing normal plant

operation. The time to place the unit in MODE 5 is appropriately

limited. Therefore, this change does not create the possibility of a

new or different kind of accident from any accident previously

evaluated.

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

of safety?

The extended time to place the unit in MODE 3 is not

significantly greater than the time currently permitted to place the

unit in Hot Shutdown and represents a reasonable time to accomplish

the shutdown. Therefore, the extended time to place the unit in MODE

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

H. CTS 3.3.4.b requires the BWST minimum boron concentration to be

within the limit specified in the core operating limits report at a

minimum temperature of 50 deg.F and would be changed to 45 deg.F in

ITS SR 3.5.4.1. BWST maximum temperature would be changed from 100

deg.F to 115 deg.F.

Basis for proposed no significant hazards consideration

determination:

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The proposed change does not involve any physical alteration of

plant systems, structures or components, changes in parameters

governing normal plant operation, or methods of operation. BWST

water temperature and volume are not

[[Page 53949]]

assumed to be the initiators of any analyzed events. As a result,

the probability of an analyzed event is independent of these values.

The proposed change from allowable values based on the uncertainties

associated with the instrument channel to an analytical limit for

the parameter being measured continues to ensure that the limits on

volume and pressure are maintained within analyzed values.

Therefore, this change does not involve a significant increase in

the probability or consequences of any accident previously

evaluated.

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

kind of accident from the accidents previously evaluated?

The proposed change does not necessitate a physical alteration

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

installed) or changes in parameters governing normal plant

operation. The analytical limits of variables established by the

safety analysis have not been changed. Thus, this change does not

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

any accident previously evaluated.

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

of safety?

Changing the limits from an allowable value based on the

uncertainties associated with the instrument channel to an

analytical limit for the parameter being measured does not involve a

significant reduction in the margin of safety since the actual

pressure and volume assumed in the safety analyses are not changed.

Based on this analysis, it appears that the three standards of 10

CFR 50.92(c) are satisfied for each of the proposed changes. Therefore,

the NRC staff proposes to determine that the amendment request involves

no significant hazards consideration.

Local Public Document Room location: Oconee County Library, 501

West South Broad Street, Walhalla, South Carolina.

Attorney for licensee: J. Michael McGarry, III, Winston and Strawn,

1200 17th Street, NW, Washington, DC.

NRC Project Director: Herbert N. Berkow.

Illinois Power Company, Docket No. 50-461, Clinton Power Station, Unit

1, DeWitt County, Illinois

Date of amendment request: July 31, 1998.

Description of amendment request: The proposed amendment would

clarify requirements for diesel generator start voltage and frequency.

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) Analyzed events are initiated by the failure of certain

plant structures, systems or components. The proposed changes to the

Clinton Power Stations (CPS) Technical Specifications revise the

acceptance criteria for Surveillance Requirements (SRs) pertaining

to the diesel generators (DGs). The DGs are not considered as

initiators of any analyzed event. Thus, these changes do not

increase the probability of any accident previously evaluated.

The consequences of analyzed events involving the diesel

generators are dependent on the successful functioning of the diesel

generator(s) to mitigate such events when a concurrent loss of

offsite power is postulated. The proposed change in the acceptance

criteria for testing of the DGs per the affected SRs accounts for DG

governor performance in response to a fast start. Notwithstanding,

the revised SRs will continue to ensure that minimum frequency and

voltage are attained within the required time, thus satisfying

permissive conditions required for closure of the DG output breaker.

The SRs will also continue to ensure that proper steady-state

voltage and frequency are attained consistent with proper DG

governor and voltage regulator performance. Additionally,

verification that permanently connected loads are energized within

the required time (in response to a loss of offsite power or in

response to a loss of coolant accident (LOCA) concurrent with a loss

of offsite power) will continue to be performed pursuant to SRs not

affected by the proposed changes. Thus, there is no impact on the

capability of the DGs to perform their required safety function.

Based on the above, IP (Illinois Power Co.) has concluded that

the proposed changes will not result in a significant increase in

the probability or consequences of any accident previously

evaluated.

(2) The proposed changes do not involve a physical alteration of

the plant. No new or different equipment is being installed, and no

installed equipment is being operated in a new or different manner.

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

normally operated or in the set points that initiate protective or

mitigative actions. As a result, no new failure modes are being

introduced.

Additionally, there are no changes in the methods governing

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

plant transients altered.

Based on the above, IP has concluded that the proposed changes

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

accident not previously evaluated.

(3) As noted previously, the proposed changes to the acceptance

criteria for testing of the DGs per the affected SRs accounts for

the characteristics of the DG governor during a fast start, but they

do not impact the effectiveness of such testing to provide assurance

of DG operability. Thus, the proposed changes do not impact expected

DG performance, including the capability for each DG to attain and

maintain required voltage and frequency for accepting and supporting

plant safety loads within the required time, as assumed in the plant

safety analyses.

Margins of safety are established through the design of the

plant structures, systems and components, the parameters within

which the plant is operated, and the establishment of set points for

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

respect to any margins of safety associated with the diesel

generators, and as noted previously, the proposed changes do not

impact diesel generator performance. That is, the SRs as revised

will continue to ensure that proper voltage and frequency are

attained for closure of the DG output breaker, and for steady-state

conditions consistent with proper DG governor and voltage regulator

performance. In addition, the proposed changes involve no changes to

any setpoints or settings associated with the diesel generators. On

this basis, the proposed changes do not involve any changes to any

assumptions of the plant safety analyses with regard to the function

of the diesel generators. Thus, no margins of safety are impacted by

the proposed changes.

Based on the above, IP has concluded that the proposed change

will not result in a reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Vespasian Warner Public

Library, 120 West Johnson Street, Clinton, IL 61727.

Attorney for licensee: Leah Manning Stetzner, Vice President,

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

Decatur, IL 62525.

NRC Project Director: Ronald R. Bellamy (acting).

Illinois Power Company, Docket No. 50-461, Clinton Power Station, Unit

1, DeWitt County, Illinois

Date of amendment request: August 17, 1998.

Description of amendment request: The proposed amendment would

reduce the load at which the diesel generators are tested.

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) Analyzed events (or events bounded by analyzed events) are

initiated by the failure of certain plant structures, systems or

components. The scope of the proposed changes is limited only to the

revision of several Surveillance Requirements (SRs) for testing of

the standby emergency diesel generators (DGs). The DGs are not

considered as initiators of any analyzed event. Thus, the proposed

changes do not impact the probability of any accident previously

evaluated.

The consequences of analyzed events are dependent on the

successful functioning of

[[Page 53950]]

credited equipment to mitigate such events. With respect to the

proposed changes, there is no impact on the capability of credited

equipment, i.e., the diesel generators, to perform as required (in

the event of a loss of coolant accident concurrent with a loss of

offsite power). Testing at reduced load levels reduces stress and

wear on the diesel generators, while still ensuring that the DGs are

adequately challenged at operating temperatures to confirm

operability. In addition, reducing the minimum required load levels

reduces time when, or the probability that, the short-term rating of

any diesel generators is exceeded during testing. The resultant

reduction in stress and wear increases DG availability.

Based on the above, IP (Illinois Power Co.) has concluded that

the proposed changes will not result in a significant increase in

the probability or consequences of any accident previously

evaluated.

(2) The proposed changes do not involve a physical alteration of

the plant. No new or different equipment is being installed, and no

installed equipment is being operated in a new or different manner.

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

normally operated or in the set points that initiate protective or

mitigative actions. As a result, no new failure modes are being

introduced.

Based on the above, IP has concluded that the proposed changes

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

accident not previously evaluated.

(3) The revised Surveillance Requirements are consistent with

the recommendations of RG [Regulatory Guide] 1.9, Revision 3.

Testing at reduced load levels reduces stress and wear on the diesel

generators, while still ensuring that the DGs are adequately

challenged at operating temperatures to confirm operability. In

addition, reducing the minimum required load levels reduces time

when, or the probability that, the short-term rating of any diesel

generators is exceeded during testing. The resultant reduction in

stress and wear increases DG availability.

Margins of safety are established through the design of plant

structures, systems and components, the parameters within which the

plant is operated, and the establishment of set points for the

actuation of equipment relied upon to respond to an event. With

respect to any margins of safety associated with the diesel

generators, the proposed changes do not impact diesel generator

performance, and involve no changes to any setpoints or settings

associated with the diesel generators, nor do the proposed changes

involve any changes to any assumptions of the plant safety analyses

with regard to the function of the diesel generators. Thus, no

margins of safety are impacted by the proposed changes.

Based on the above, IP has concluded that the proposed changes

will not result in a reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Vespasian Warner Public

Library, 120 West Johnson Street, Clinton, IL 61727.

Attorney for licensee: Leah Manning Stetzner, Vice President,

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

Decatur, IL 62525.

NRC Project Director: Ronald R. Bellamy (Acting).

Indiana Michigan Power Company, Docket No. 50-315, Donald C. Cook

Nuclear Plant, Unit 1, Berrien County, Michigan

Date of amendment request: August 28, 1998.

Description of amendment request: The proposed amendment would

grant relief from the steam generator inspection surveillance

requirement described in technical specification No. 4.4.5.3. The

relief would allow the inspection to be deferred from April 8, 1999,

until the next refueling outage for Donald C. Cook Nuclear Plant , Unit

1.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

In accordance with CFR 50.92, the proposed amendment will not

involve a significant hazards consideration if the changes do not:

1. Involve a significant increase in the probability or

consequences of an accident previously analyzed;

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

from any accident previously analyzed or evaluated; or

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

Criterion 1

The last unit 1 surveillance was completed in the spring of 1997

and was the most thorough evaluation of the steam generators to

date. Both standard and enhanced eddy current inspection techniques

were employed to inspect the steam generator tubing. Additionally, a

series of in situ pressure tests were performed to verify tubing

integrity. Tube repairs consisting of hot leg tube end re-rolling

and plugging were performed. Pre- and post-tube bundle pressure

tests were conducted to verify the integrity of the repairs. A tube

pull was also conducted to verify continued comformance with generic

letter 95-05 requirements. The tube pull data did not identify any

unexpected conditions or areas of concern. During the 1997

inspection, select secondary side visual and eddy current

inspections were also performed to provide assurance of continued

secondary side internals integrity.

Following the inspection, a condition monitoring and operational

assessment, using data gathered during the steam generator

inspections and tests, was made to determine whether steam generator

leakage and structural integrity could be maintained throughout the

upcoming cycle (cycle 16).

The unit was subsequently restarted and the steam generators

operated without incident when a unit shutdown occurred in September

of 1997.

Throughout the cycle 16 operating period, a relatively low

reactor coolant temperature was maintained. By maintaining a T-hot

temperature of approximately 586 deg.F during the operating period,

corrosion impact on the steam generator tubes was minimized.

Throughout the operating period, steam generator primary-to-

secondary leakrate monitoring was performed to assure conformance

with T/S requirements. Historically, Unit 1 has not experienced a

forced shutdown because of leakrate concerns.

During the shutdown period, the steam generators have been

maintained under lay-up conditions, which comply with or exceed the

industry standard practice. These practices are designed to mitigate

the corrosive environment within the steam generators.

The previous cycle 16 integrity assessment has been re-visited

to provide reasonable assurance conclusions made remain valid given

the extended shutdown period. This re-assessment considered the

initial cycle runtime, the shutdown period and subsequent operation

through the end of the current fuel cycle. These results confirm the

findings of the initial evaluation (i.e., that adequate steam

generator integrity will be maintained throughout the current

cycle).

The proposed change will not affect the scope, methodology,

acceptance limit, or corrective measures of the existing steam

generator examination program. As adequate integrity will be

maintained, the probability and consequences of an accident

previously analyzed due to leaking or degraded tubes is not

increased by the proposed change.

Criterion 2

We have determined that this extension will not result in a

change in plant configuration or operation. Plant systems and

components will not be operated in a different manner as a result of

this change. No plant modifications or changes in methods of

operation will result from this change. Therefore, the extension

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

accident from what has been previously evaluated or analyzed.

Criterion 3

We have determined that the proposed extension request will not

involve a significant reduction in a margin of safety. Re-assessment

of the cycle 16 steam generator operational assessment report, which

indicates structural and leakage integrity will be maintained

throughout the cycle, has shown that the shutdown period will not

adversely impact overall steam generator integrity.

This assessment concluded that when the reactor is shut down and

the reactor coolant system is at a reduced temperature, the steam

[[Page 53951]]

generators are not subject to conditions that lead to tube

degradation. The actual number of days that the steam generators

will be subjected to an environment conducive to tube degradation is

not being increased under this request. Therefore, this request is

judged not to 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: Maud Preston Palenske Memorial

Library, 500 Market Street, St. Joseph, MI 49085.

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

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

NRC Project Director: Ronald R. Bellamy (Acting).

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: August 12, 1998.

Description of amendment request: The proposed amendment would

change the Technical Specifications (TS) by updating the list of

documents specified in TS 6.9.1.8b that describe the analytical methods

used to determine the core operating limits. The changes can be

categorized as: (1) The analysis methodology is unchanged, but the

reference has been clarified by identifying the specific revision,

supplements, and dates for the revision; (2) the analysis methodology

is unchanged and the reference is being added for completeness and; (3)

the analysis methodology is being changed. 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 in reference 4 of Technical Specification

Section 6.9.1.8b revises the steam line break analysis methodology

to be applied to Millstone Unit No. 2 and clarifies the references

to the Siemens topical reports. The other changes are clarifications

or additions for completeness and do not represent a change in the

approved methodology for Millstone Unit No. 2. The change in

methodology is associated with the interference between XTGPWR, the

neutronics code, and XCOBRA-IIIC, the thermal hydraulics code. It

has no impact on plant equipment operation. Since the change only

affects the analysis of the events, it cannot affect the likelihood

or consequences of these events. Therefore, this change will not

significantly increase the probability or consequences of an

accident previously evaluated.

The sentence on page 6-19, starting with ``The acceptable

Millstone 2 * * *.'' and ending with ``* * * dated October, 1988,''

references the document ANF-88-126, ``Millstone Unit 2 Cycle 10

Safety Analysis Report,'' which has been outdated because of the

above mentioned changes in the methodology. The removal of this

sentence is necessary to be consistent with methodology changes.

Therefore, this change will not significantly increase 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 change in reference 4 of Technical Specification

Section 6.9.1.8b revises the steam line break analysis methodology

to be applied to Millstone Unit No. 2 and clarifies the references

to the Siemens topical reports. The other changes are clarifications

or additions for completeness and do not represent a change in the

approved methodology for Millstone Unit No. 2. The proposed change

in reference 4 of Technical Specification Section 6.9.1.8b will not

alter the plant configuration (no new or different type of equipment

will be installed) or require any new or unusual operator actions.

It does not alter the way any structure, system, or component

functions and does not alter the manner in which the plant is

operated.

The sentence on page 6-19, starting with ``The acceptable

Millstone 2 * * *.'' and ending with ``* * * dated October, 1988,''

references an outdated document. The removal of this sentence is

necessary to be consistent with methodology changes. The change does

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

does not alter the manner in which the plant is operated.

The 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 change in reference 4 of Technical Specification

Section 6.9.1.8b revises the steam line break analysis methodology

to be applied to Millstone Unit No. 2 and clarifies the references

to the Siemens topical reports. The other changes are clarifications

or additions for completeness and do not represent a change in the

approved methodology for Millstone Unit No. 2. The change in steam

line break methodology is associated with the interface between

XTGPWR, the neutronics code, and XCOBRA-IIIC, the thermal hydraulics

code. The change will result in a better correlation between the two

computer codes, which is the intent of the iteration process. This

will result in more accurate results while still maintaining a

conservative modeling of the event. The most significant impact is

on the low RCS [reactor coolant system] flow cases associated with

loss of offsite power. These cases are not limiting when compared to

the offsite power available cases. The improved references will

clearly identify the approved Siemens Topical Reports applicable to

Millstone Unit No. 2 and will ensure that methodology changes will

be identified and submitted to the NRC for approval as required. The

sentence on page 6-19, starting with ``The acceptable Millstone 2 *

* *.'' and ending with ``* * * dated October, 1988,'' references an

outdated document. The removal of this sentence is necessary to be

consistent with methodology changes.

Therefore, the proposed changes will not result in a significant

reduction in the margin of safety as defined in the Bases for

Technical Specifications covered in this License Amendment Request.

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, PO Box 270, Hartford,

Connecticut.

NRC Project Director: William M. Dean.

Northern States Power Company, Docket No. 50-263, Monticello Nuclear

Generating Plant, Wright County, Minnesota

Date of amendment request: August 15, 1996, as supplemented March

19, 1998.

Description of amendment request: The proposed amendment revises

the Technical Specifications so that either 8 or 12 hour shifts will be

considered ``normal'' and 40 hours will be considered a ``nominal''

week, changes the wording for surveillances required ``once per shift''

to ``once per 12 hours,'' clarifies the ``once per hour'' wording

related to fire watch patrols, and makes a number of other

typographical corrections and clarifications.

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

[[Page 53952]]

consideration, which is presented below:

(1) The proposed amendment will not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

(With respect to shift definition and editorial changes:) This

change does not affect the physical configuration of the plant or

how it is operated, as such, it is not the initiator of any plant

event. Working a ``normal'' 12-hour shift is no different from

working a ``normal'' 8-hour shift with 4-hours of overtime which has

been an accepted and approved practice for years. Therefore, the

proposed changes will not result in any increase in the probability

of an accident occurring. The intent is still that operators will

not work excessive overtime either on a daily, or weekly basis.

The typographical errors, clarifications and title changes do

not involve technical issues and as such do not involve safety

issues, and therefore do not effect [sic] the chances or

consequences of an accident.

(With respect to surveillance and fire watch patrol interval:)

This change does not affect the physical configuration of the plant

or how it is operated. As such, it is not the initiator of any plant

event. This change clarifies the intervals in which Sensor Checks,

Surveillances, and fire watch patrols must be completed. As

described above [in the supplement], the 12-hour interval has been

determined acceptable for the specified Sensor Checks and

Surveillances based on Monticello and industry experience which

demonstrates instrumentation and channel failures are rare. This

change conforms the Monticello TS (Technical Specifications) to

NUREG-1433 and clarifies the intervals in which checks must be

completed.

Completing fire watch patrols on a one hour +25% interval will

require patrols on an hourly basis, while providing flexibility to

complete the patrols within a 15 minute window. In addition to the

Technical Specification required fire watches, additional

individuals are often in the plant proper, so the required hourly

fire watch patrols are only part of the entire program for fire

detection.

Therefore, the proposed changes will not result in a significant

increase in the probability of an accident occurring.

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

new or different kind of accident from any accident previously

analyzed.

(With respect to shift definition and editorial changes:) This

change does not affect the physical configuration of the plant or

how it is operated. Therefore, revising the length of a ``normal''

shift or correcting minor errors does not create the possibility of

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

As such, it is not the initiator of any plant event.

(With respect to surveillance and fire watch patrol interval:)

Revising the wording to ``once per 12 hours'' or ``once per hour

(+25%)'' does not create the possibility of a new or different kind

of accident from any previously evaluated. No new or different

surveillance activities are proposed, nor are any being deleted. As

such, it is not the initiator of any plant event.

(3) The proposed amendment will not involve a significant

reduction in the margin of safety.

(With respect to shift definition and editorial changes:) This

change does not affect the physical configuration of the plant or

how it is operated. The level of expertise on shift will not be

diminished or changed as a result of this change. Therefore, this

change will not reduce the margin of safety.

(With respect to surveillance and fire watch patrol interval:)

This change does not affect the physical configuration of the plant

or how it is operated. The level of expertise on shift will not be

diminished or changed, nor will it reduce the functionality of plant

equipment. This change requires Sensor Checks, surveillances, and

fire watch patrols be completed within industry guidelines.

The 12 hour interval has been determined acceptable based on

industry experience which demonstrates channel failure is rare. The

one hour interval for fire watch patrols has also been an accepted

industry standard. In addition to the Technical Specification

required fire watches, additional individuals are often in the plant

proper, so the required hourly fire watch patrols are only part of

the entire program for fire detection. The proposed change simply

defines the acceptable interval during which the task must be

performed. Therefore, this change does not constitute a significant

reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Minneapolis Public Library,

Technology and Science Department, 300 Nicollet Mall, Minneapolis,

Minnesota 55401.

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

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

NRC Project Director: Cynthia A. Carpenter.

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

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

California

Date of amendment request: January 14, 1998, as supplemented by

letter dated May 19, 1998.

Description of amendment request: The proposed amendment would

approve a modification to the Diablo Canyon Power Plant, Unit Nos. 1

and 2, 230 kV transmission system. The modifications include

installation of new 230/12kV startup transformers with automatic load

tap changers, along with installation of shunt capacitor banks. The

transformers will assure that voltage on the plant 12 kV and 4 kV buses

is maintained within limits, while the capacitor banks assure adequate

VAR support.

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 replacement of the startup transformers (SUTs) with new

transformers equipped with load tap changers (LTCs) for voltage

control does not alter the original configuration of the electrical

distribution system and hence, will not increase the probability of

occurrence of an accident previously evaluated.

The replacement of the SUTs with new transformers equipped with

LTCs will enhance the capability of the 12 kV and 4 kV electrical

distribution systems to maintain sufficient voltage for successful

transfer of the plant auxiliary loads to the startup source

following a unit trip. This change eliminates the potential for

``double sequencing'' (starting loads from the 230 kV system,

subsequent voltage degradation causes load shedding and restarting

from the diesel generators) of the 4 kV vital loads during an

accident by providing adequate voltage to the 4 kV vital buses from

the 230 kV source. The maintenance of adequate voltage at the 4 kV

vital buses prevents the second level undervoltage relay (SLUR)

action. The LTC will automatically maintain adequate voltage at the

terminals of the vital equipment under design basis accident

conditions. Therefore, engineered safety feature equipment will

function as previously evaluated.

The manual operation of the Unit 2 LTC while in a standby mode

will not increase the probability of an accident since normally none

of the plant loads are energized from the 230 kV system. Plant loads

are only powered from the 230 kV system during short periods of unit

startup and shutdown. Loss of the 230 kV system while the operating

plant loads are fed from the 25/500 kV system cannot initiate an

accident since the system is not connected to plant equipment if the

loads are supplied by the 25/500 kV system. Therefore, the proposed

modifications will not increase the probability of an accident

previously evaluated. The manual operation of the Unit 2 LTC assures

adequate voltage is supplied to Unit 2 safety equipment in the event

of an accident. Therefore, the proposed modification will not

increase the consequences of an accident.

The installation of the shunt capacitors at the Diablo Canyon

Power Plant switchyard and Mesa Substation to replace the VAR

support from Morro Bay Power Plant (MBPP), assuming no MBPP

generation, does not alter the capability or availability of the

offsite

[[Page 53953]]

power source. Since shunt capacitors are considered more reliable

than generators, it adds to the reliability of the 230 kV system and

will not increase the probability of an accident previously

evaluated.

Even if 230 kV voltage were lost or became degraded, the first

or second level undervoltage relays will initiate transfer to the

diesel generators should there be a loss or degraded 230 kV system

while feeding the vital loads from the 230 kV system. This scenario

is evaluated in Final Safety Analysis Report (FSAR) Update Section

15.2.9.1 ``Loss of Offsite Power to the Station Auxiliaries.''

Therefore, the changes will not increase the consequences of an

accident previously evaluated since the safety-related loads will

function as required.

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

or different kind of accident from any accident previously

evaluated.

The change does not result in a change in operation,

maintenance, physical change, or procedural change that could create

the possibility of an accident that is of a new or different type

than previously evaluated.

The replacement SUTs and the installation of the shunt

capacitors to replace MBPP serves the same function as the original

design and do not create the possibility of a new or different type

of accident. Should there be a loss of offsite power, the onsite

power source (diesel generators) will provide power to the loads.

The FSAR already includes an evaluation for station blackout if

there is a total loss of both onsite and offsite power.

Therefore, the proposed changes do not create the possibility of

a new or different kind of accident from any accident previously

evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The replacement transformers and the installation of the shunt

capacitors will not cause a reduction in the margin of safety as

defined in the basis for any Technical Specification (TS). The

minimum voltage required for safe shutdown is defined in TS Table

3.3.4, Functional Unit 7.b, ``Second Level Undervoltage Relay (SLUR)

setting.'' By replacing the existing SUTs with automatic LTC

transformers, the vital 4 kV bus voltage will be automatically

maintained at a sufficiently higher value during normal operation

such that during an accident, the minimum 4 kV vital bus voltages

after the bus transfer will be adequate to prevent SLUR actuation.

The installation of the shunt capacitors will assure adequate VAR

support that was previously provided by operation of the MBPP in the

Los Padres Region of PG&E's service territory for present peak load

and future peak load growth under worse case line outage conditions.

During the interim period between January and February 1998,

when manual control of the Unit 2 SUT LTC will be utilized to

maintain adequate voltage at the 12 kV and 4 kV buses, the margin of

safety is not reduced since the adjustment of the LTC will assure

stable voltage for the vital buses.

Therefore, there is no reduction in a margin of safety as

defined in the basis for any TS.

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

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

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

amendment requests involve no significant hazards consideration.

Local Public Document Room Location: California Polytechnic State

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

Department, San Luis Obispo, California 93407.

Attorney for Licensee: Christopher J. Warner, Esq., Pacific Gas &

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

NRC Project Director: William H. Bateman.

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

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

California

Date of amendment request: March 18, 1998.

Description of amendment request: The proposed amendment would

approve a change in the way passive failures in the auxiliary saltwater

(ASW) and component cooling water (CCW) systems are mitigated during

the long-term recovery period following a loss-of-coolant accident

(LOCA). Specifically, plant procedures would no longer require ASW and

CCW system train separation after the transfer to hot leg recirculation

following a LOCA.

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 changes revise the way passive failures are mitigated in the

auxiliary saltwater (ASW) and component cooling water (CCW) systems.

Specifically, plant procedures would no longer require ASW and CCW

train separation after transfer to hot leg recirculation following a

loss-of-coolant accident. The decision to separate trains would be

made by the Technical Support Center (TSC) after evaluation of plant

conditions. Operation of the ASW and CCW systems during this period

is required to mitigate the accident, therefore, the change in plant

operation would not affect the probability of that accident

occurring.

The change ensures the ASW and CCW systems will be able to

mitigate an active or passive failure without the loss of safety

function during the long-term (beginning 24 hours after the

accident) period of recovery following an accident. Since the ASW

and CCW systems will continue to perform their safety function,

overall system performance is not affected, assumptions previously

made in evaluating the consequences of the accident are not altered,

and the consequences of the accident are not increased as a result

of the change in plant operation.

Therefore, the changes will not increase 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 ASW and CCW systems function to mitigate the consequences of

an accident. The change in operation ensures these system will be

able to mitigate an active or passive failure without loss of safety

function during the long-term (beginning 24 hours after the

accident) period of recovery following an accident. Operation of the

ASW and CCW systems in accordance with plant procedures, and the

guidance on train separation provided to the TSC, ensure the design

basis requirements for the ASW and CCW systems will continue to be

met. Therefore, the ability of the ASW and CCW systems to mitigate

the accident is not degraded. Required operator actions are similar

to other operator actions specified in the FSAR that are considered

acceptable by the NRC.

Therefore, the proposed changes do not create the possibility of

a new or different kind of accident from any accident previously

evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The change ensures the ASW and CCW systems will be able to

mitigate an active or passive failure without loss of safety

function during the long-term (beginning 24 hours after the

accident) period of recovery following an accident. Since the ASW

and CCW systems will continue to perform their safety function,

there is no impact on any acceptance limits for ASW and CCW system

operation assumed in the safety analysis, or on any Technical

Specification (TS).

Therefore, the change does not involve a significant reduction in a

margin of safety as defined in the basis for any TS.

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

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

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

amendment requests involve no significant hazards consideration.

Local Public Document Room Location: California Polytechnic State

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

Department, San Luis Obispo, California 93407.

Attorney for Licensee: Christopher J. Warner, Esq., Pacific Gas &

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

[[Page 53954]]

NRC Project Director: William H. Bateman.

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

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

California

Date of amendment request: August 10, 1998,

Description of amendment request: The proposed amendments would

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

Power Plant, Unit Nos. 1 and 2 to revise TS 3/4.3.2, Table 3.3-5,

``Engineered Safety Features Response Times,'' to add the response

times for closure of the main feedwater regulating valves (MFRVs) and

MFRV bypass valves, and trip of the main feedwater pumps (MFWPs). The

change would also revise TS 3/4.7.1.7 to add a limiting condition for

operation (LCO), actions, and surveillance requirements for the MFWP

turbine stop valves, and would revise the actions and surveillance

requirements for the MFRVs, MFRV bypass valves, and main feedwater

isolation valves (MFIVs) to be consistent with the NUREG-1431

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. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The proposed changes to the Technical Specifications (TS) to add

response time requirements for the main feedwater regulating valve

(MFRV) and associated bypass valves and the main feedwater pump

(MFWP) trip provide more restrictive TS requirements that are

consistent with current plant practice. They do not change the

function or operation of any plant equipment or affect the response

of that equipment if it is called upon to operate. These more

restrictive requirements are imposed to ensure the affected

components are maintained consistent with the safety analyses and

licensing bases.

The proposed changes to: (1) Revise the actions to apply to one

or more main feedwater isolation valves (MFIVs), and MFRVs and

associated bypass valves, (2) extend the action completion time from

4 hours to 72 hours, (3) provide actions when two valves affecting

the feedwater isolation capability for a flow path are inoperable,

(4) add actions for an inoperable MFWP turbine stop valve, and (5)

allow separate action entry for each inoperable valve unless the

feedwater isolation capability for a flow path is affected, do not

change the function or operation of any plant equipment or affect

the response of that equipment if it is called on to operate. The

actions account for the redundancy provided by the remaining valves

and the MFWP trip, and the low probability of an event occurring

during this time period that would require isolation of the main

feedwater flow path. A probabilistic risk assessment, performed to

assess the increase in annual core damage frequency (CDF) associated

with the increase in allowable outage time, determined the increase

in annual CDF to be approximately 1.5 percent. That increase in

annual CDF is considered non-risk significant per the Electric Power

Research Institute ``PSA Application Guide.''

The addition of the limiting condition for operation, actions,

and surveillance requirements for the MFWP turbine stop valves, and

the addition of the surveillance requirement for the MFIVs, MFRVs,

and MFRV bypass valves are more restrictive requirements that ensure

these components are operable and capable of performing their safety

function. They do not change the function or operation of any plant

equipment or affect the response of that equipment if it is called

on to operate. The proposed surveillance intervals are supported by

the operating, maintenance, and surveillance histories of the

valves.

Therefore, the proposed changes do not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

The proposed changes do not necessitate a physical alteration of

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

or changes in the parameters governing normal plant operation. The

changes imposed are consistent with the assumptions made in the

accident analyses and licensing basis.

Therefore, the proposed changes do not create the possibility of

a new or different kind of accident from any accident previously

evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The proposed changes to the TS impose requirements consistent

with the assumptions in the safety analyses and current licensing

bases, and reflect current plant practice. They do not alter the

margins of safety established in previous accident and transient

analysis.

Therefore, none of the proposed changes involves a significant

reduction in a margin of safety.

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

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

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

amendment requests involve no significant hazards consideration.

Local Public Document Room Location: California Polytechnic State

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

Department, San Luis Obispo, California 93407.

Attorney for Licensee: Christopher J. Warner, Esq., Pacific Gas &

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

NRC Project Director: William H. Bateman.

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

Generating Station, Salem County, New Jersey

Date of amendment request: September 8, 1998.

Description of amendment request: The proposed amendment would

revise Technical Specification (TS) Appendix C, ``Additional

Conditions,'' to authorize the use of non-Class 1E single cell battery

chargers, with proper electrical isolation, for charging connected

cells in OPERABLE Class 1E batteries. The single cell chargers would be

used to restore individual cell float voltage to the normal limit

specified in TS Table 4.8.2.1-1.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

in the probability or consequences of an accident previously

evaluated.

The proposed change permits the use of an industry accepted

method to restore a battery cell to its design basis from an

OPERABLE but degraded condition or to prevent a cell from becoming

degraded. IEEE Std 450-1995, ``IEEE Recommended Practice for

Maintenance, Testing, and Replacement of Vented Lead Storage

Batteries for Stationary Applications,'' states that single cell

charging is an acceptable method of correcting low cell voltage or

low specific gravity conditions for a single cell or for a small

number of cells.

At least two class 1E fuses in series will be used on both the

positive and negative leads between the battery and the charger to

protect the battery if a fault should develop in the charger. The

battery charger design includes diodes, a power transformer and

control circuitry to prevent draining the connected cells in the

event of a short circuit in the 120 Volt ac source or a loss of

charger input or output voltage. Charger output is controlled

automatically to prevent overcharging the connected cells.

In the event of a controller failure resulting in charger

overvoltage, procedural controls governing the use of the charger

ensure the condition is detected and corrected before failure of a

connected cell occurs. While the single cell charger is connected,

procedures will require periodic checks to verify proper charger

operation and to measure electrolyte level, temperature and specific

gravity for the cells being charged. Monitoring will be

[[Page 53955]]

performed at least once every eight hours, a frequency sufficient to

ensure compliance with the ACTION requirements of Technical

Specification 3.8.2.1.

An insulating material will be used to minimize the possibility

of shorting leads or clips at the battery. Administrative controls

governing the use and storage of transient loads are sufficient to

ensure the use of single cell battery chargers does not create a

potential missile hazard to safety related systems, structures and

components.

The Class 1E dc system is not an accident initiator. It supports

the operation of safety related equipment required for the safe

shutdown of the plant and for the mitigation of accident conditions.

Therefore, the proposed change does not increase the probability of

an accident previously evaluated.

The station's dc systems will be operable to mitigate the

consequences of an accident previously evaluated. Single cell

charging would be limited to one OPERABLE class 1E battery bank at a

time. Therefore, failure of a class 1E battery as a result of single

cell charging would be limited to a single channel and would not

reduce the number of OPERABLE dc sources below that required to

safely shutdown the plant. Administrative controls would also

prohibit the use of single cell charging for an OPERABLE class 1E

battery if less than the minimum number of class 1E batteries

required by Technical Specifications are OPERABLE.

The proposed change does not cause the capability of the class

1E dc system to be degraded below the level assumed for any accident

described in the (safety analysis report) SAR. It would enhance the

availability of safety related equipment required for the safe

shutdown of the plant and for the mitigation of accident conditions.

Therefore the radiological consequences of an accident will remain

inside the design basis while single cell charging is performed on

an OPERABLE battery.

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

or different kind of accident from any accident previously

evaluated.

The potential to adversely affect the Class 1E batteries is

minimized by the use of Class 1E fuses and by appropriate

administrative controls. Failure modes associated with the proposed

change are bounded by the loss of a Class 1E battery bank which was

previously evaluated. Therefore, the proposed change does not create

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

accident previously evaluated.

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

in a margin of safety.

The proposed change permits the use of non-Class 1E single cell

battery chargers, with proper electrical isolation, for charging

connected cells in OPERABLE class 1E batteries. This would allow

parameters for an individual cell or for a small number of cells to

be restored to the normal values specified in Technical

Specifications without affecting the remainder of the cells in the

battery. Increased cell monitoring after single cell charging,

together with PSE&G's corrective action program which requires

degraded and non-conforming conditions to be documented and

evaluated, provides assurance that the use of single cell charging

will not cause long-term cell degradation to go undetected. Since

all battery cells are required to be maintained within the allowable

values specified in Technical Specifications, and since the use of

the single cell charger will not adversely affect battery capacity

or capability, the proposed change does not involve a significant

reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: 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.

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

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

Unit No. 1

Fairfield County, South Carolina.

Date of amendment request: July 1, 1998.

Description of amendment request: The proposed amendment would

revise Virgil C. Summer Nuclear Station (VCSNS) Technical

Specifications (TS) Surveillance Requirement 4.7.7.e to remove the

``during shutdown'' condition from the specified test interval.

Removing the ``during shutdown'' wording from the TS would allow VCSNS

to perform on-line snubber testing, and would make the up to 25 percent

allowable interval extension in Surveillance Requirement 4.0.2 apply to

the specified snubber surveillance interval. The proposed amendment

would also make administrative changes to Surveillance Requirement

4.7.7.g and BASES 3/4.2.2 and 3/4.2.3 to correct typographical errors.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The probability or consequences of an accident previously

evaluated is not significantly increased.

The proposed change will not affect system operation or

performance, nor do they affect any Engineered Safety Features

actuation setpoints or accident mitigation capabilities. NUREG/CR-

6027 supports the determination that piping failure due to a snubber

single failure is considered low. Therefore, the proposed changes

will not significantly increase the consequences of an accident or

malfunction of equipment important to safety previously evaluated in

the FSAR.

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

different type than any previously evaluated is not created.

The changes to the situational testing requirements will not

affect the method of operation of any system to which a snubber is

attached. The proposed changes only address the plant mode at which

a surveillance activity may be performed. No new or different

accident scenarios, transient precursors, failure mechanisms, or

limiting single failures will be introduced as a result of these

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

accident other than those already evaluated will not be created by

this change.

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

This proposed change will not have an impact on the overall

reliability of the snubber population. This is due, in part, to the

fact that the snubber test plans are self correcting. As functional

test failures are identified, additional snubbers are required to be

tested. Thus, the reliability of the snubber population is

maintained. The proposed change does not alter the intent or method

by which the surveillances are conducted, does not involve any

physical changes to the plant, does not alter the way any structure,

system, or component functions, and does not modify the manner in

which the plant is operated. Therefore the proposed change will not

degrade the ability of the snubbers to perform their safety function

or significantly decrease 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: Fairfield County Library, 300

Washington Street, Winnsboro, SC 29180.

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

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

NRC Acting Project Director: P. T. Kuo.

Southern Nuclear Operating Company, Inc., et al., Docket Nos. 50-424

and 50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke

County, Georgia

Date of amendment request: June 26, 1998, as supplemented by letter

dated September 18, 1998.

Description of amendment request: The proposed amendments would

change the Technical Specifications (TS) as follows: (1) The

applicability of

[[Page 53956]]

Limiting Condition for Operation (LCO) 3.3.6 would be revised to refer

to TS Tables 3.3.6-1 and 3.3.6-1; the TS Tables would be revised to add

a column entitled ``APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS.''

Then, the applicable modes for Manual Initiation, Automatic Actuation

Logic and Actuation Relays, and Safety Injection functions would be

revised to include only Modes 1, 2, 3, and 4. Consistent with this

proposed change, LCO 3.3.6, Condition C and Required Action C.2 would

be revised to reflect that system level manual initiation and automatic

actuation would not be required during core alterations and/or during

movement of irradiated fuel assemblies within containment. Appropriate

Bases changes are included to reflect the proposed changes; (2) LCO

3.9.4 would be revised to allow the equipment hatch and the emergency

air locks to be open during core alterations and/or during movement of

irradiated fuel assemblies within containment. In addition, the LCO

statement would be revised to reflect that containment ventilation

isolation (CVI) would be accomplished by manually closing the

individual CVI valves as opposed to a system level manual or automatic

initiation, consistent with the proposed changes to LCO 3.3.6. The

surveillance requirements (SRs) would be revised to reflect the

proposed change to the CVI and to reflect that the equipment hatch

would be allowed to be open. Appropriate Bases changes are included to

reflect the proposed changes; (3) LCO 3.7.6a, ``Condensate Storage Tank

(CST)--(Non-redundant CSTs),'' would be deleted. This LCO was created

to address a design condition that rendered the CSTs nonredundant. A

note was added stating that this LCO was only applicable to the unit(s)

that have not completed design modifications required for redundant

CSTs and that the LCO would no longer be required when both units

completed the design modifications. These design modifications have

been completed; therefore, LCO 3.7.6a is no longer applicable, and LCO

3.7.6, ``Condensate Storage Tank (CST)--(Redundant CSTs),'' would be

revised to delete the words ``(Redundant CSTs)'' from the title.

Appropriate Bases changes are included 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. Do the proposed changes involve a significant increase in the

probability or consequences of an accident previously evaluated?

No. The proposed changes would revise the VEGP [Vogtle Electric

Generating Plant] Unit I and Unit 2 TS by removing requirements for

automatic and system level manual containment ventilation isolation,

and allow the emergency air lock and the equipment hatch to be open

during core alterations and movement of Irradiated fuel assemblies

inside containment. The containment penetrations affected by the

proposed changes are not initiators for any accident previously

evaluated. Allowing these penetrations to be open under the

conditions specified will not affect the probability of any accident

previously evaluated.

The existing VEGP TS allow the personnel air look doors to be

open during core alterations and movement of irradiated fuel

assemblies inside containment. The radiological consequences of a

fuel handling accident inside containment have been determined to be

below the Standard Review Plan (SRP) section 15.7.4 criteria and

General Design Criteria (GDC) 19 criteria with the personnel air

lock doors open. The proposed changes will not alter these

previously determined consequences. The existing dose analysis

bounds the proposed changes. Therefore, the proposed changes will

not increase the consequences of any accident previously evaluated.

The proposed deletion of LCO 3.7.6a is an administrative change

only. The requirements of LCO 3.7.6a applied only during the time

that the condensate storage tanks (CSTs) were not redundant. Due to

the implementation of design changes which make the CSTs redundant

for each unit, the requirements of LCO 3.7.6a are no longer

applicable. The CSTs (redundant or not) are not initiators for any

accident previously evaluated. Now that the CSTs are redundant, the

requirements of LCO 3.7.6a are no longer necessary to ensure the

capability of the auxiliary feedwater system to perform its safety

function. Therefore, the proposed deletion of LCO 3.7.6a will not

affect the probability or consequences of any accident previously

evaluated.

2. Do the proposed changes create the possibility of a new or

different kind of accident from any accident previously evaluated?

No. The proposed change does not create any new failure modes

for any system or component, nor does it adversely affect plant

operation. The previously determined radiological consequences of a

fuel handling accident inside containment with the personnel air

lock doors open remain bounding for operation under the proposed

changes. No new single failure scenarios are created, and the

proposed changes do not introduce any new challenges to components

and systems that could result in a new or different kind of accident

from any previously evaluated.

The proposed deletion of LCO 3.7.6a is an administrative change

only. The requirements of LCO 3.7.6a applied only during the time

that the condensate storage tanks (CSTs) were not redundant. Due to

the implementation of design changes which make the CSTs redundant

for each unit, the requirements of LCO 3.7.6a are no longer

applicable. Now that the CSTs are redundant, the requirements of LCO

3.7.6a are no longer necessary to ensure the capability of the

auxiliary feedwater system to perform its safety function. No new

single failure scenarios are created, and the proposed changes do

not introduce any new challenges to components and systems that

could result in a new or different kind of accident from any

previously evaluated. Therefore, the proposed deletion of LCO 3.7.6a

will not create a new or different kind of accident from any

accident previously evaluated.

3. Do the proposed changes involve a significant reduction in a

margin of safety?

No. The margin of safety for fission product release is 300 rem

thyroid and 25 rem whole body as defined by 10 CFR (Part) 100. The

previously determined radiological dose consequences for a fuel

handling accident inside containment with the personnel air lock

doors open remain bounding for operation under the proposed changes.

These previously determined dose consequences were determined to be

well within the limits of 10 CFR (Part) 100 by virtue of the fact

that they meet SRP Section 15.7.4 and GDC 19 acceptance criteria.

Therefore, the proposed changes do not involve a significant

reduction in a margin of safety.

The proposed deletion of LCO 3.7.6a is an administrative change

only. The requirements of LCO 3.7.6a applied only during the time

that the condensate storage tanks (CSTs) were not redundant. Due to

the implementation of design changes which make the CSTs redundant

for each unit, the requirements of LCO 3.7.6a are no longer

applicable. Now that the CSTs are redundant, the requirements of LCO

3.7.6a are no longer necessary to ensure the capability of the

auxiliary feedwater system to perform its safety function.

Therefore, LCO 3.7.6a is not necessary to maintain margin of safety

and the proposed change will not involve a reduction in a margin of

safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Burke County Public Library,

412 Fourth Street, Waynesboro, Georgia.

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

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

Georgia.

NRC Project Director: Herbert N. Berkow.

[[Page 53957]]

Southern Nuclear Operating Company, Inc., Georgia Power Company,

Oglethorpe Power Corporation, Municipal Electric Authority of Georgia,

City of Dalton, Georgia, Docket Nos. 50-424 and 50-425, Vogtle Electric

Generating Plant, Units 1 and 2, Burke County, Georgia

Date of amendment request: July 13, 1998

Description of amendment request: The proposed amendment would

change Technical Specification (TS) Section 1.1 Definitions for

``Engineered Safety Feature (ESF) Response Time'' and ``Reactor Trip

System (RTS) Response Time'' to provide for verification of response

time for selected components provided that the components and the

methodology for verification have been previously reviewed and approved

by the NRC. Changes to the TS Bases have also been proposed.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

(1) The proposed license amendment does not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

This change to the Technical Specifications does not result in a

condition where the design, material, and construction standards

that were applicable prior to the change are altered. The same RTS

[reactor trip system] and ESFAS [engineered safety features

actuation system] instrumentation is being used; the time response

allocations/modeling assumptions in the Chapter 15 analyses are

still the same; only the method of verifying time response is

changed. The proposed change will not modify any system interface

and could not increase the likelihood of an accident since these

events are independent of this change. The proposed activity will

not change, degrade or prevent actions or alter any assumptions

previously made in evaluating the radiological consequences of an

accident described in the SAR [safety analysis report]. Therefore,

the proposed amendment does not involve a significant increase in

the probability or consequences of an accident previously evaluated.

(2) The proposed license amendment does not create the

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

previously evaluated.

This change does not alter the performance of the pressure and

differential pressure transmitters and switches, Process Protection

racks, Nuclear Instrumentation, and Logic Systems used in the plant

protection systems. Applicable sensors, Process Protection racks,

Nuclear Instrumentation, and Logic Systems will still have response

time verified by test before placing the equipment into operational

service and after any maintenance that could affect the response

time. Changing the method of periodically verifying instrument

response times for certain equipment (assuring equipment

operability) from time response testing to calibration and channel

checks will not create any new accident initiators or scenarios.

Periodic surveillance of these instruments will detect significant

degradation in the equipment response time characteristics.

Implementation of the proposed amendment does not create the

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

previously evaluated.

(3) The proposed license amendment does not involve a

significant reduction in margin of safety.

This change does not affect the total system response time

assumed in the safety analysis. The periodic system response time

verification method for selected pressure and differential pressure

sensors and for Process Protection racks, Nuclear Instrumentation,

and Logic Systems is modified to allow use of actual test data or

engineering data. The method of verification still provides

assurance that the total system response time is within that assumed

in the safety analysis, since calibration tests will detect any

degradation which might significantly affect equipment response

time. Based on the above, it is concluded that the proposed license

amendment request does not result in a significant reduction in

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: Burke County Public Library,

412 Fourth Street, Waynesboro, Georgia.

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

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

Georgia.

NRC Project Director: Herbert N. Berkow.

Southern Nuclear Operating Company, Inc., Georgia Power Company,

Oglethorpe Power Corporation, Municipal Electric Authority of Georgia,

City of Dalton, Georgia, Docket Nos. 50-424 and 50-425, Vogtle Electric

Generating Plant, Units 1 and 2, Burke County, Georgia

Date of amendment request: September 3, 1998.

Description of amendment request: The proposed amendments would

change the Technical Specifications (TS) to: (1) Support the

replacement of the Nuclear Instrumentation System Source Range and

Intermediate Range Channels and Post-Accident Neutron Flux Monitoring

System; and (2) delete the requirement for performing response time

testing of the source range channels and power range detector plateau

voltage determinations.

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 power range low trip, the intermediate range trip, and

the source range trip are designed to provide protection against

power excursions during reactor startup or low-power operation. The

source and intermediate range trips provide redundant protection

during reactor startup or low-power operation. The changes to the

source range and intermediate range instrumentation and setpoints,

as well as the deletion of source range response time testing, do

not affect any safety analysis conclusions because the source range

and intermediate range trips are not explicitly credited in any

design basis accident. Only the power range low trip setpoint is

assumed to actuate to mitigate the uncontrolled rod cluster control

assembly withdrawal accident. The high flux at shutdown alarm

function during a boron dilution event will continue to be provided

by the new source range detector system. No changes have been made

to the setpoint assumed in the safety analyses. The new detector

system is qualified in compliance with Regulatory Guide 1.97 and

will also be used to provide post-accident monitoring. The

functional and operability requirements for the power range channels

are not affected by deleting the requirement for determining

detector voltage plateaus.

Therefore, based on the conclusions of the above evaluation, the

proposed changes will not involve a significant increase in the

probability or consequences of an accident previously evaluated.

2. The functional and operability requirements for the new

detector system are the same as for the existing system as defined

by the Technical Specifications. No credit is taken for the source

and intermediate range trips in any of the design basis accidents.

The high flux at shutdown alarm and post-accident monitoring

functions continue to be met. The functional and operability

requirements for the power range channels are not affected by

deleting the requirement for determining detector voltage plateaus.

Therefore, the proposed changes will not create the possibility

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

evaluated.

3. The functional and operability requirements for the new

detector system are the same as for the existing system. The

functional and operability requirements for the power range channels

are not affected by deleting the requirement for determining

detector voltage plateaus. The margin of safety provided by the

previous Technical Specifications is not significantly affected

because the proposed changes are based on the same accident analysis

acceptance limits.

[[Page 53958]]

Therefore, the proposed changes in this license amendment will

not result in a significant reduction in the plant's 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: Burke County Public Library,

412 Fourth Street, Waynesboro, Georgia.

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

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

Georgia.

NRC Project Director: Herbert N. Berkow.

Tennessee Valley Authority, Docket No. 50-260 Browns Ferry Nuclear

Plant Unit 2 Limestone County, Alabama

Date of amendment request: September 8, 1998.

Description of amendment request: The proposed amendment would

revise the Browns Ferry Nuclear Plant (BFN) Unit 2 technical

specifications (TS) to include provisions for enabling the Oscillation

Power Range Monitor (OPRM) Upscale trip function in the Average Power

Range Monitor (APRM). The APRM is part of the Power Range Neutron

Monitoring (PRNM) system. The OPRM Upscale trip function provides

protection from exceeding the fuel Minimum Critical Power Ratio (MCPR)

safety limit in the event of thermal-hydraulic power oscillations, and

thereby, provides compliance with Title 10 Code of Federal Regulations,

Part 50, Appendix A, General Design Criteria (GDC) 10 and 12.

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 amendment is to enable the OPRM Upscale trip

function which is contained in the previously installed PRNM

equipment. Enabling the OPRM hardware provides the long term

stability solution required by Generic Letter 94-02.

This hardware incorporates the Option III detect and suppress

solution reviewed and approved by the NRC in NEDO-31960, ``BWROG

[Boiling Water Reactor Owners Group] Long Term Stability Solutions

Licensing Methodology.'' The OPRM is designed to meet all

requirements of GDC 10 and 12 by automatically detecting and

suppressing design basis thermal-hydraulic power oscillations prior

to violating the fuel MCPR Safety Limit. The OPRM system provides

this protection in the region of the power-to-flow map where

instabilities can occur, including the region where ICAs (interim

corrective actions) restricted operation because of stability

concerns. Thus, the ICA restrictions on plant operations are deleted

from the TS, including region avoidance and the requirement for the

operator to manually scram the reactor with no recirculation loops

operating. Operation at high core powers with low core flows may

cause a slight, but not significant, increase in the probability

that an instability can occur. This slight increase is acceptable

because subsequent to the automatic detection of a design basis

instability, the OPRM Upscale trip provides an automatic scram

signal to the RPS [reactor protection system] which is faster

protection than the operator-initiated manual scram required by the

current ICAs. Because of this rapid automatic action, the

consequences of an instability event are not increased as a result

of the installation of the OPRM system because it eliminates

dependence on operator actions.

Based on the above discussion, the proposed amendment does not

involve a significant increase in 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 proposed amendment permits BFN to enable the OPRM power

oscillation detect and suppress function provided in previously

installed PRNM hardware, and it simultaneously deletes certain

restrictions which preclude operation in regions of the power-to-

flow map where oscillations potentially may occur. Enabling the OPRM

Upscale trip function does not create any new system hardware

interfaces nor create any new system interactions. Potential

failures of the OPRM Upscale trip result either in failure to

perform a mitigation action or in spurious initiation of a reactor

scram. These failures would not create the possibility of a new or

different kind of accident. Based on the above discussion, the

proposed amendment does not create the possibility of a new or

different kind of accident from any accident previously.

C. The proposed amendment does not involve a significant

reduction in a margin of safety.

The OPRM Upscale trip function implements BWROG Stability Option

III, which was developed to meet the requirements of GDC 10 and GDC

12 by providing a hardware system that detects the presence of

thermal-hydraulic instabilities and automatically initiates the

necessary actions to suppress the oscillations prior to violating

the MCPR Safety Limit. The NRC has reviewed and accepted the Option

III methodology described in Licensing Topical Report NEDO-31960 and

concluded this solution will provide the intended protection.

Therefore, it is concluded that there will be no reduction in the

margin of safety as defined in TS as a result of enabling the OPRM

Upscale trip function and simultaneously removing the operating

restrictions previously imposed by the ICAs.

Based on the above discussion, 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

its 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: Athens Public Library, 405 E.

South Street, Athens, Alabama 35611.

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: August 27, 1996, and as supplemented on

July 22, 1998.

Description of amendment request: The amendment request removes the

Technical Specification requirements for the Main Steam Isolation Valve

Leakage Control System, and increases the allowable leak rate specified

for the main steam lines. The Perry facility is a pilot plant in the

collaborative efforts of the Nuclear Regulatory Commission, the Nuclear

Energy Institute, and the Electric Power Research Institute for

implementation of the NRC research documented in NUREG-1465, ``Accident

Source Terms for Light-Water Nuclear Power Plants.'' The proposed

changes are based on reanalysis of the design basis Loss of Coolant

Accident using the revised accident source term from NUREG-1465 and the

NEI document entitled ``Generic Framework for Application of Revised

Accident Source Term to Operating Plants.''

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.

[[Page 53959]]

The proposed change removes the Technical Specification

requirements for the Main Steam Isolation Valve Leakage Control

System (MSIV-LCS), and increases the allowable leak rate specified

for the main steam lines. Although the requirements for the MSIV-LCS

are being removed (since credit is no longer taken for the system as

part of the design basis accident analysis), OPERABILITY

requirements on the Main Steam Shutoff Valves are being retained

since the valves meet Criterion 3 of 10 CFR 50.36(c)(2)(ii).

Removing the Technical Specification requirements of the MSIV-LCS

and increasing main steam line allowable leakage rates has been

addressed in the Loss of Coolant Accident (LOCA) reanalysis and does

not adversely affect operation of other equipment or systems

important to safety. These changes do not affect the precursors for

accidents or transients analyzed in Chapter 15 of the Perry Nuclear

Power Plant (PNPP) Updated Safety Analysis Report (USAR). Therefore,

there is no increase in the probability of accidents previously

evaluated.

The spectrum of LOCAs was considered to determine which would be

most limiting with respect to radiological consequences. The worst

case LOCA (i.e., main steam line break upstream of the inboard MSIV)

off-site and Control Room doses have been reanalyzed using the

revised design basis accident (DBA) source term (from NUREG-1465 and

the Nuclear Energy Institute (NEI) document ``Generic Framework for

Application of Revised Accident Source Term to Operating Plants'')

in order to assess the radiological consequences of the increased

main steam line leak rates, and not taking credit for the MSIV-LCS.

The radiological analysis used conservative assumptions and

analytical techniques. These conservatisms in the LOCA reanalysis

have been determined to be comparable to the conservatisms utilized

in the original analyses.

The results of the off-site and Control Room dose reanalysis are

provided below.

Dose Results (REM)

----------------------------------------------------------------------------------------------------------------

Existing USAR Regulatory

Proposed USAR dose* dose limit **

----------------------------------------------------------------------------------------------------------------

Control Room......................... Whole Body............... 0.1 0.4 5

Thyroid.................. 16.2 29.2 30

Skin..................... 4.8 2.5 30

EAB.................................. Whole Body............... 1.9 3.6 25

Thyroid.................. 157.9 140.8 300

LPZ.................................. Whole Body............... 1.7 1.9 25

Thyroid.................. 130.3 144.7 300

----------------------------------------------------------------------------------------------------------------

* Rounded to nearest tenth.

** Exclusion Area Boundary (EAB) and Low Population Zone (LPZ) dose limits are per 10 CFR 100.11. Control Room

dose limits are per 10 CFR part 50 Appendix A, General Design Criterion (GDC) 19 and NUREG 0800 Standard

Review Plan (SRP) Section 6.4.

As noted in the NEI Generic Framework Document (``Generic

Framework for Application of Revised Accident Source Term to

Operating Plants,'' EPRI TR-105909, Interim Report, November 1995),

the acceptability of applications utilizing the revised accident

source terms ``may be judged by the same licensing acceptance limits

(e.g., dose limits in 10 CFR part 100) in use with the TID-14844

source term. That is, the licensee would show that the revised

design basis, with either selective or essentially complete

application of NUREG-1465 together with the plant changes under

evaluation, results in doses no greater than these licensing

acceptance limits.'' The off-site dose licensing acceptance limit

for PNPP is 10 CFR part 100.11 (see Question 3 for details on the

source of this PNPP licensing acceptance limit). The newly

calculated radiological doses were lower for six of the seven

factors evaluated. For the one factor which was higher, i.e., at the

EAB for thyroid dose (from 140.8 REM to 157.9 REM), the dose

remained significantly below the 10 CFR part 100 limit of 300 REM to

the thyroid. This analysis demonstrated that the resulting off-site

and Control Room doses were well below the regulatory limits

contained in 10 CFR part 100, Reactor Site Criteria, and 10 CFR part

50, Appendix A, General Design Criterion (GDC) 19, Control Room.

Therefore, the proposed changes do not involve a significant

increase in the consequences of previously evaluated accidents.

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

or different kind of accident from any previously evaluated.

The proposed change removes the Technical Specification

requirements for the MSIV-LCS, retains the Technical Specification

requirements for the Main Steam Shutoff Valves, and increases the

allowable leak rate specified for the main steam lines.

Removing the Technical Specification requirements for the MSIV-

LCS is based on reanalysis of off-site and Control Room doses, where

the MSIV-LCS is not credited in the calculation. As noted above, the

reanalysis utilizes the revised design basis accident (DBA) source

terms. The limiting reanalysis case assumes that main steam line

leakage is attenuated in the main steam line from the reactor vessel

out to the outboard MSIV. This is the limiting scenario since the

worst case single failure, and hence the most limiting analysis

case, involves a failure to close the valve downstream of the

outboard MSIV in each main steam line, i.e., the Main Steam Shutoff

Valves (1N11F0020A,B,C AND D). Although this most limiting analysis

case assumes a failure to close the Main Steam Shutoff Valves,

retention of OPERABILITY requirements on these valves is appropriate

to ensure the single failure analysis remains valid.

Not crediting the MSIV-LCS in the design basis accident analysis

is consistent with the approach taken by several BWR licensees,

which have applied for NRC approval of this change using an approach

developed by the Boiling Water Reactor Owners Group (BWROG). The

BWROG methodology involves seismically qualifying the main steam

lines out to and including the non-safety related, non-seismic drain

line and main condenser, and then using that volume to attenuate

leakage past the MSIVs. At PNPP, the existence of safety related,

seismically qualified piping leading to the safety related, Class 1E

powered Main Steam Shutoff Valves (downstream of the outboard MSIV),

together with the characteristics of the revised accident source

term (i.e., predominantly aerosol which is largely retained in the

drywell, containment and main steam lines) provides the option of

taking credit only for the volume within the main steam lines for

leakage attenuation.

Knowledge of the more physically correct source term timing and

chemical form permits use of more appropriate mitigation techniques.

Specifically, natural forces such as gravitational settling of

aerosol (particulates) has been credited inside the drywell and in

portions of the main steam lines, which significantly reduces the

amount of radionuclides that could escape from the containment and

into the environment. Also, based on a high radiation signal in the

Control Room, the Containment Spray system would be operated post-

LOCA for up to 24 hours (previous analyses assumed 6 hours of spray

operation), in order to scrub released radionuclides from the

containment atmosphere and into the suppression pool, and thus

reduce the post-LOCA off-site and Control Room dose. Once the

containment sprays have been successful in sweeping the iodine to

the suppression pool, the iodine must be retained in the water. To

achieve this, the pH level of the suppression pool will now be

raised to 7 or above following the accident, and then maintained at

7 or above. This prevents significant fractions of the dissolved

iodine from being converted to elemental iodine and then re-evolving

to the containment atmosphere. During the course of the accident the

pH of the suppression pool can decrease due to radiolysis of reactor

coolant and chloride-bearing electrical insulation, which would

create acids. The

[[Page 53960]]

method for pH control will use the existing Standby Liquid Control

(SLC) system for raising (and maintaining) long-term post-accident

pH levels to 7 or above. Calculations have shown that the contents

of one tank of the Standby Liquid Control solution will be effective

in raising and maintaining pH levels for 30 days following the DBA.

Post-accident operator actions are minimized. The operator

action associated with initiating the Containment Spray system does

not change. Containment Spray is initiated via a push button in the

Control Room. The previously required manual initiation of the MSIV-

LCS involved multiple operator actions to open and close numerous

valves and start the blowers, which will no longer be required.

Replacing these actions, the new analysis simply assumes the

operator closes the Main Steam Shutoff Valves (which was previously

one of the steps in manually initiating the MSIV-LCS system), and

based on post-accident pH samples of the suppression pool, initiates

the Standby Liquid Control system, which is accomplished via two key

lock switches in the Control Room. These operator actions are less

complex than those previously required, and minimize the probability

of an error.

Other accidents, as described in USAR 15, were reviewed. The

original methodology, input parameters and overall conclusions

contained within these accident evaluations were found to be

unaffected by the changes proposed by this activity. Removing the

Technical Specification requirements of the MSIV-LCS and increasing

MSIV allowable leakage rates has been addressed in the LOCA

reanalysis and does not adversely affect operation of other

equipment or systems important to safety. This activity does not

alter or impact plant systems, structures or components which were

not appropriately addressed in the LOCA reanalysis. No new accident

initiator or failure mode is introduced. The physical isolation of

the MSIV-LCS from the Main Steam system will eliminate leakage

pathways. This modification will be performed as part of the PNPP

design change process.

With respect to the change in main steam line leakage limits,

the BWROG has concluded, based on an in-depth evaluation of MSIV

leakage (as discussed in NEDC-31858 ``BWROG Report for Increasing

MSIV Leakage Rate Limits and Elimination of Leakage Control

Systems,'' Revision 2, and summarized in NUREG-1169 ``Technical

Findings Related to Generic Issue C-8; Boiling Water Reactor Main

Steam Isolation Valve and Leakage Treatment Methods''), that leakage

rates of up to 500 scfh are not indicative of substantial mechanical

defects in the valves which would challenge the capability of the

valves to fulfill their safety function of isolating the steam

lines. Therefore, as demonstrated in the design basis LOCA

radiological reanalysis, the proposed increased allowable MSIV

leakage rate (i.e., each line less than or equal to 100 scfh and

total leakage less than or equal to 250 scfh when tested at Pa) will

not affect each MSIV's isolation function capability. Additionally,

no new operator actions or errors are introduced as a result of the

increased main steam line leakage limits, other than those addressed

above.

Based on the above discussions, the proposed change would 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 worst case LOCA (i.e., a main steam line break upstream of

the inboard MSIV) has been reanalyzed using the revised DBA source

term (NUREG-1465 and the NEI generic framework document) in order to

assess the radiological consequences of the increased MSIV leak

rate, and not taking credit for MSIV-LCS. The radiological analyses

used conservative assumptions and analytical techniques. The results

of the revised DBA source term dose calculations should be

determined acceptable using the current licensing basis acceptance

limits (those that were used for initial plant licensing).

As noted in the NEI Generic Framework Document (``Generic

Framework for Application of Revised Accident Source Term to

Operating Plants,'' EPRI TR-105909, Interim Report, November 1995),

``to demonstrate that an adequate margin of safety is maintained,

the licensee may show that the doses associated with the revised

design basis (resulting from the revised source term together with

the plant change under evaluation) are less than the licensing

acceptance limits for the plant.''

The licensing acceptance limits for off-site dose are discussed

in Supplement 8 to the NRC Safety Evaluation Report (SER) for PNPP,

Section 15.3, ``Radiological Consequences of Design Basis

Accidents.'' The licensing acceptance limits are the guideline

values of 10 CFR 100.11, ``Reactor Site Criteria.'' The SER states

``The doses computed for this accident are less than the guideline

values of 10 CFR 100.11 and the staff concludes that the Perry plant

is adequately designed to mitigate the off-site consequences arising

from a LOCA.'' For Control Room doses, the licensing acceptance

limit is discussed in Supplement 10 to the NRC SER, Section 6.4,

``Control Room Habitability.'' The licensing acceptance limits are

as stated therein, i.e., ``The staff's LOCA analysis indicates that

the Control Room doses are within the guidelines of General Design

Criterion (GDC) 19 of Appendix A to 10 CFR part 50 and of Section

6.4 of the Standard Review Plan (SRP, NUREG-0800).''

The revised PNPP design basis calculations (i.e., the revised

DBA source term coupled with the plant changes under evaluation)

demonstrated that the resulting off-site and Control Room doses were

below the licensing acceptance limits contained in 10 CFR part 100,

10 CFR part 50, Appendix A, General Design Criterion 19, and SRP

Section 6.4. An acceptable margin of safety is inherent in these

licensing acceptance limits. The improvement in the technical

knowledge base and in the analytical techniques that are part of the

revised accident source term, and the modeling of the increased MSIV

leakages without taking credit for MSIV-LCS, do not alter the

acceptability of the margin. Therefore, the resulting calculated

Control Room and off-site doses, which are well within regulatory

limits, ensure that the proposed change does not involve a

significant reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: 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 Acting Project Director: Ronald R. Bellamy.

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: September 3, 1998.

Description of amendment request: The proposed license amendment

increases the present Division 3 Diesel Generator (High Pressure Core

Spray System) fuel level requirements to account for (1) a rounding

error in the calculation, and (2) the unusable volume due to vortex

formation at the eductor nozzles located in the fuel oil storage tank.

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 revises the Division 3 Diesel Generator (DG)

7-day fuel oil supply requirement and the 6-day fuel oil supply

requirement due to a rounding error in the calculation and due to

the consideration of vortex formation near the eductor suction

nozzle located near the bottom of the fuel oil storage tank. The

proposed change ensures a sufficient DG fuel oil volume to maintain

submergence of the eductor suction nozzle so that a vortex formation

does not occur. Eliminating the concerns of a vortex formation will

provide assurance that the DG fuel oil system will perform its

intended function. Analyzed events are initiated by the failure of

plant structures, systems, or components. The DGs are not considered

as initiators of any analyzed event. The proposed change does not

have a detrimental

[[Page 53961]]

impact on the integrity of any plant structure, system, or component

that initiates an analyzed event. The proposed change will not alter

the operation of, or otherwise increase its failure probability of

any plant equipment that initiates an analyzed event. As such, the

probability of occurrence for a previously analyzed accident is not

significantly increased.

The consequences of a previously analyzed event are dependent on

the initial conditions assumed for the analysis, the availability

and successful functioning of the equipment assumed to operate in

response to the analyzed event, and the setpoints at which these

actions are initiated. The proposed change ensures a sufficient DG

fuel oil volume to maintain submergence of the eductor suction

nozzle so that a vortex formation does not occur. The proposed

change continues to ensure that the DG fuel oil system will

adequately support the design basis performance and mitigative

function of the DG. The proposed change does not affect the

performance of any credited equipment. As a result, no analyses

assumptions are violated and there are no adverse effects on the

factors that contribute to offsite or onsite dose as the result of

an accident. The proposed change does not affect setpoints that

initiate protective or mitigative actions. The proposed change

ensures that plant structures, systems, or components are maintained

consistent with the safety analysis and licensing bases. Based on

this evaluation, there is no significant increase in the

consequences of a previously analyzed event.

Therefore, this change will not involve a significant increase

in the probability or consequences of any accident previously

evaluated.

(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 revises the Division 3 DG 7-day fuel oil

supply requirement and the 6-day fuel oil supply requirement due to

a rounding error in the calculation and due to the consideration of

vortex formation near the eductor suction nozzle located near the

bottom of the fuel oil storage tank. The proposed change ensures a

sufficient DG fuel oil volume to maintain submergence of the eductor

suction nozzle so that a vortex formation does not occur.

Eliminating the concerns of a vortex formation will provide

assurance that the DG fuel oil system will perform its intended

function. The proposed change does not involve a physical change to

the DG fuel oil system or tank, nor does it change the operating

characteristics or the safety function of the DG. The proposed

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

or different equipment is being installed and no installed

equipment, which might initiate a new or different kind of accident,

is being operated in a different manner. The proposed change does

not impact core reactivity or the manipulation of fuel bundles. The

DG performs a mitigative function. There is no alteration to the

parameters within which the plant is normally operated or in the

setpoints that initiate protective or mitigative actions. 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 revises the Division 3 DG 7-day fuel oil supply

requirement and the 6-day fuel oil supply requirement due to

rounding error in the calculation and due to the consideration of

vortex formation near the eductor suction nozzle located near the

bottom of the fuel oil storage tank. The margin of safety is being

maintained by the proposed change from the margin of safety

established by the original design. The proposed change ensures a

sufficient DG fuel oil volume to maintain submergence of the eductor

suction nozzle so that vortex formation does not occur. Eliminating

the concerns of a vortex formation will provide assurance that the

DG fuel oil system will perform its intended function. The proposed

change does not significantly impact the condition or performance of

structures, systems, and components relied upon for accident

mitigation. The proposed change, in fact, provides assurance of the

DG's ability to perform its intended function as previously

evaluated. 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 Acting Project Director: Ronald R. Bellamy.

Toledo Edison Company, Centerior Service Company, and The Cleveland

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

Power Station, Unit 1, Ottawa County, Ohio

Date of amendment request: September 8, 1998.

Description of amendment request: The proposed amendment would

change Technical Specification (TS) Section 5.3.1, ``Design Features--

Reactor Core--Fuel Assemblies.'' A different type of fuel rod cladding

would be added. The associated bases would also be changed.

Basis for proposed no significant hazards consideration

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

provided their analysis of the issue of no significant hazards

consideration, which is presented below:

The Davis-Besse Nuclear Power Station has reviewed the proposed

changes and determined that a significant hazards consideration does

not exist because operation of the Davis-Besse Nuclear Power Station

in accordance with these changes would:

1a. Not involve a significant increase in the probability of an

accident previously evaluated because it has been demonstrated that

the material properties of the M5 alloy are not significantly

different from those of Zircaloy-4. Further, there are no evaluated

accidents in which the fuel cladding or fuel assembly structural

components are assumed to arbitrarily fail as an accident initiator.

The fuel handling accident assumes that the cladding does, in fact,

fail as a result of an undefined fuel handling event. However, the

probability of that undefined initiating event is independent of the

properties of the fuel rod cladding.

1b. Not involve a significant increase in the consequences of an

accident previously evaluated because it has been demonstrated that

the material properties of the M5 alloy are not significantly

different from those of Zircaloy-4. Therefore, in both non-LOCA and

LOCA accident scenarios, there will be no significant increase in

cladding failure or fission product release.

2. Not create the possibility of a new or different kind of

accident from any accident previously evaluated because it has been

demonstrated that the material properties of the M5 alloy are not

significantly different from those of Zircaloy-4. Therefore, M5 fuel

cladding and fuel assembly structural components will perform

similarly to those fabricated from Zircaloy-4, thus precluding the

possibility of the fuel becoming an accident initiator and causing a

new or different kind of accident.

3. Not involve a significant reduction in a margin of safety

because it has been demonstrated that the material properties of the

M5 alloy are not significantly different from those of Zircaloy-4.

The M5 alloy is expected to perform similarly to Zircaloy-4 for all

normal operating and accident scenarios, including both non-LOCA and

LOCA scenarios. For LOCA scenarios, where the slight differences in

M5 material properties relative to Zircaloy-4 could have

[[Page 53962]]

some impact on the overall accident scenario, plant-specific LOCA

analyses will be performed prior to the use of batch quantities of

fuel assemblies containing either fuel rod cladding, fuel rod end

plugs, or fuel assembly structural components fabricated from M5.

These plant-specific LOCA analyses, required by TS 6.9.1.7, ``Core

Operating Limit Report,'' will either demonstrate that all current,

applicable, and appropriate margins of safety will be maintained

during the use of the M5 alloy or their results will be submitted

for NRC review and approval prior to use of the M5 alloy.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: University of Toledo, William

Carlson Library, Government Documents Collection, 2801 West Bancroft

Avenue, Toledo, OH 43606.

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

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

NRC Acting Project Director: Ronald R. Bellamy.

Yankee Atomic Electric Company, Docket No. 50-029, Yankee Nuclear Power

Station, Franklin County, Massachusetts

Date of amendment request: August 20, 1998.

Description of amendment request: By letter dated August 20, 1998,

the licensee submitted a License Amendment request related to three

Technical Specification (TS) administrative changes. The first is to

remove a definition from the DEFINITIONS section of the TS that is

provided in 10 CFR part 20. The second change is to transfer the site

map from Section 5.0 of the TS to the Final Safety Analysis Report and

to replace the map with a textual description of the site location.

Lastly, to delete TS 5.1.1--EXCLUSION AREA.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

The proposed changes are administrative in nature and in no way

affect the safety of the Yankee Nuclear Power Station (YNPS). The

proposed deletion of the definition for SITE BOUNDARY in no way

reduces or eliminates any regulatory requirement which Yankee Atomic

Electric Company must currently satisfy. Likewise, the relocation of

the YNPS site map from the YNPS Technical Specifications to the YNPS

Final Safety Analysis Report is devoid of any safety implications.

Therefore, the proposed changes will not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated. The administrative

nature of the changes will not affect safety related systems or

components and, therefore, involve no significant increase in the

probability or consequences of an accident previously evaluated.

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

any previously evaluated. The proposed changes do not modify any

plant systems or components and, therefore, do not create the

possibility of a new or different accident from any previously

evaluated.

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

proposed changes do not involve any physical changes to the plant

nor any changes in plant procedures. Therefore, there will be no

reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Greenfield Community College,

1 College Drive, Greenfield, Massachusetts 01301.

Attorney for licensee: Thomas Dignan, Esquire, Ropes and Gray, One

International Place, Boston, Massachusetts 02110-2624.

NRC Project Director: Seymour H. Weiss.

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.

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

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

Carolina

Date of amendment request: August 27, 1998.

Brief description of amendment request: The amendment revises

Technical Specifications 3.0.4 and 4.0.4 to be consistent with the

guidance provided in Generic Letter 87-09 dated June 4, 1987.

Date of publication of individual notice in Federal Register:

September 8, 1998 (63 FR 47529).

Expiration date of individual notice: October 8, 1998.

Local Public Document Room location: Cameron Village Regional

Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.

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

Generating Station, Ocean County, New Jersey

Date of amendment request: August 21, 1998.

Description of amendment request: The amendment would remove the

requirement for the Automatic Depressurization System function of the

Electromatic Relief Valves to be operable during Reactor Vessel

Pressure Testing. Additionally, note h of Table 3.1.1 will be corrected

due to a typographical error introduced in the issuance of Amendment

75.

Date of publication of individual notice in Federal Register:

September 10, 1998 (63 FR 48527).

Expiration date of individual notice: October 13, 1998.

Local Public Document Room location: Ocean County Library,

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

Wisconsin Public Service Corporation, Wisconsin Power and Light Company

and Madison Gas and Electric Company, Docket No. 50-305, Kewaunee

Nuclear Power Plant, Kewaunee County, WI

Date of application for amendment: April 8, 1998, modified by

letter dated August 27, 1998.

Brief description of amendment request: The proposed amendment

would reduce the maximum allowable level of reactor coolant system

activity (dose equivalent 1-131) to provide a means of accepting higher

projected leak rates for steam generator tubes while still meeting

offsite and control room dose criteria. Also included is a change to

the secondary coolant activity level for which an increased sampling

frequency applies.

[[Page 53963]]

Date of publication of individual notice in Federal Register:

September 14, 1998 (63 FR 49137).

Expiration date of individual notice: October 14, 1998.

Local Public Document Room location: University of Wisconsin,

Cofrin Library, 2420 Nicolet Drive, Green Bay, WI 54311-7001.

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 regula

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