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

NUCLEAR REGULATORY COMMISSION

Biweekly Notice; Applications and Amendments to Facility

Operating Licenses Involving No Significant Hazards Considerations

I. Background

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

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

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

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

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

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

the Commission the authority to issue and make immediately effective

any amendment to an operating license upon a determination by the

Commission that such amendment involves no significant hazards

consideration, notwithstanding the pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all notices of amendments issued, or

proposed to be issued from July 31, 1999, through August 13, 1999. The

last biweekly notice was published on August 11, 1999 (64 FR 43764).

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

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

publication of this notice will be considered in making any final

determination.

Normally, the Commission will not issue the amendment until the

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

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

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

the Commission may issue the license amendment before the expiration of

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

the amendment involves no significant hazards consideration. The final

determination will consider all public and State comments received

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

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

opportunity for a hearing after issuance. The Commission expects that

the need to take this action will occur very infrequently.

Written comments may be submitted by mail to the Chief, Rules and

Directives Branch, Division of Administration Services, Office of

Administration, U.S. Nuclear Regulatory Commission, Washington, DC

20555-0001, and should cite the publication date and page number of

this Federal Register notice. Written comments may also be delivered to

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

Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of

written comments received may be examined at the NRC Public Document

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

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

is discussed below.

By September 24, 1999, 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

[[Page 46426]]

document room for the particular facility involved.

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 2, 1999.

Description of amendment request: The proposed amendment would

revise Technical Specification 6.2.2.e to require either the Operations

Manager or an off-shift Operations superintendent to hold a senior

reactor operator (SRO) license. This revision would delete the option

which allows the Manager-Operations to have at one time held a Senior

Reactor Operator License for a similar unit and replaces it with the

requirement for an off-shift Operations superintendent who holds an SRO

license to supervise shift work and licensed activities.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

The change to Technical Specification 6.2.2.e to require the

Manager-Operations or an off-shift Operations superintendent to hold

an SRO license is administrative in nature and does not directly

affect plant operations. The change does not physically alter the

facility in any manner and, as such, does not affect the means in

which any safety-related system performs its intended safety

function.

Therefore, there would be no increase in the probability or

consequences of an accident previously evaluated.

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

new or different kind of accident from any accident previously

evaluated.

As stated above, the proposed change is administrative in

nature. There is no physical alteration to any plant system, nor is

there a change in the method in which any safety related system

performs its function.

Therefore, the proposed change does not create the possibility

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

evaluated.

3. The proposed amendment does not involve a significant

reduction in the margin of safety.

The proposed amendment does not reduce the margin of safety as

defined in the Safety Analysis Report or the bases contained in the

Technical Specifications. The requirement to have a licensed SRO

management position responsible for plant operations is maintained

within the proposed amendment. The proposed amendment is consistent

with (1) 10 CFR 50.54(l), which requires individuals responsible for

directing the licensed activities of licensed operators to hold an

SRO license, (2) Revision 1 of NUREG-1431, ``Standard Technical

Specifications Westinghouse Plants,'' and Technical Specification

Traveler Form (TSTF) 65, Revision 1, and (3) the intent of ANSI/ANS-

3.1, ``Standard for Selection and Training of Personnel for Nuclear

Power Plants,'' (September 1979 Draft).

Therefore, 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: Cameron Village Regional

Library, 1930 Clark Avenue, Raleigh, North Carolina 27605

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

Corporate Secretary, Carolina Power & Light Company, Post Office Box

1551, Raleigh, North Carolina 27602

NRC Section Chief: Sheri R. Peterson.

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 4, 1999.

Description of amendment request: The proposed amendment would

revise Technical Specification 6.9.1.6.2 to incorporate analytical

methodology references which are used to determine core operating

limits. The analytical methodologies to be referenced are documented in

topical reports which have been accepted by the Nuclear Regulatory

Commission for referencing in licensing applications.

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.

The proposed changes incorporate additional references to

methodologies used to evaluate core operating limits. These

methodologies have been approved for use by the NRC. Plant

structures, systems, and components will not be operated in a

different manner as a result of these proposed changes and no

physical modifications to equipment are involved. Adding these

references to the Core Operating Limits Report section of Technical

Specifications does not increase the probability or consequences of

an accident previously evaluated.

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

new or different kind of accident from any accident previously

evaluated.

The proposed changes incorporate additional references to

methodologies used to evaluate core operating limits. These

methodologies have been approved for use by the NRC. Plant

structures, systems, and components will not be operated in a

different manner as a result of these proposed changes and no

physical modifications to equipment are involved. Adding these

references to the Core Operating Limits Report section of Technical

Specifications does not create the possibility of a new or different

type of accident from any previously evaluated.

3. The proposed amendment does not involve a significant

reduction in the margin of safety.

The proposed changes incorporate additional references to

methodologies used to evaluate core operating limits. These

methodologies have been approved for use by the NRC. Plant

structures, systems, and components will not be operated in a

different manner as a result of these proposed changes and no

physical modifications to equipment are involved. Adding these

references to the Core Operating Limits Report section of Technical

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

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Cameron Village Regional

Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.

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

Corporate Secretary, Carolina Power & Light Company, Post Office Box

1551, Raleigh, North Carolina 27602.

NRC Section Chief: Sheri R. Peterson.

Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden

Nuclear Power Station, Units 2 and 3, Grundy County, Illinois

Date of amendment request: May 20, 1999.

Description of amendment request: The proposed amendments would

revise Technical Specification (TS) 3.8.A to identify the specific

Containment Cooling Service Water (CCSW) equipment required to support

operation of the Control Room Emergency Ventilation System (CREVS). The

proposed amendment would also

[[Page 46427]]

revise TS 3/4.5.C.2 to ensure that the suppression pool water level is

adequate to prevent vortexing in the Low Pressure Coolant Injection and

Core Spray pump suctions.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated because of the

following:

The proposed changes to the technical specifications provide

clarity in the support system relationship and requirements for the

CCSW system support of the CREVS operation. [Neither] [t]he CCSW

system nor the CREVS system are assumed to be accident precursors

for previously evaluated accident[s]. Therefore, the proposed

changes have no effect on the probability or consequences of

accidents previously evaluated.

The proposed change to the allowable suppression chamber level

does not involve a significant increase in the probability or

consequences of an accident previously evaluated. The proposed

change revises a Technical Specification acceptance value to [a]

more conservative value and serves to ensure operability of

equipment important to safety. By ensuring equipment availability,

the probability or consequences of an accident previously evaluated

are not increased. In addition, the proposed changes have no impact

on any initial condition assumptions for accident scenarios. Onsite

or offsite dose consequences resulting from an event previously

evaluated are not affected by this proposed amendment request.

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

from any accident previously evaluated because:

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

different kind of accident from that previously evaluated. The

changes to the CCSW specifications more appropriate[ly] reflect the

design requirements and clarify the support role of the CCSW system

as it relates the CREVS. Neither the CCSW system nor the CREVS will

be operated differently with the proposed change. Therefore new or

different failure modes will not be created. Therefore, the

possibility of new and different accidents has not been created with

the proposed change. The proposed change to the suppression pool

allowable level restores margin to the Technical Specifications and

ensures equipment operability. The proposed change is conservative

with respect to current requirements. The proposed amendment does

not involve any plant physical changes that would create the

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

previously evaluated.

Therefore, the proposed amendment does not create the

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

previously evaluated.

3. Involve a significant reduction in the margin of safety

because:

The proposed change to the CCSW technical specification will not

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

proposed change has greater consistency with the current design

requirements for CSSW support of CREVS operation. Therefore, the

margin of safety has been not been altered. [Therefore, the margin

of safety has not been altered. SIC]

The proposed changes for suppression pool level does not involve

a significant reduction in a margin of safety. In fact, the proposed

changes restore margin and ensure equipment operability. Since the

changes maintain the necessary level of system reliability, they do

not involve a significant reduction in the margin of safety.

The proposed amendment for Dresden will not reduce the

availability of systems required to mitigate accident conditions;

therefore, the proposed changes do not involve a significant

reduction in the margin of safety.

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

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

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

requested amendments involve no significant hazards consideration.

Local Public Document Room location: Morris Area Public Library

District, 604 Liberty Street, Morris, Illinois 60450.

Attorney for licensee: Ms. Pamela B. Stroebel, Senior Vice

President and General Counsel, Commonwealth Edison Company, P.O. Box

767, Chicago, Illinois 60690-0767.

NRC Section Chief: Anthony J. Mendiola.

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

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

Date of amendment request: July 14, 1999.

Description of amendment request: The proposed amendments would

allow the units to operate at an uprated power level of 3489 MWt, an

increase of 5 percent rated core thermal power.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

A. Evaluation of the Probability of Previously Evaluated Accidents

The proposed power uprate imposes only minor increases in plant

operating conditions. No change is made to the reactor operating

pressure. Operation at uprated conditions will result in moderate

flow increases in those systems associated with the turbine cycle in

that steam flow increases by approximately six (6)% and feed flow

increases by approximately six (6)%. The increase in flow in the

carbon steel piping systems was evaluated for the effect on flow

induced erosion and corrosion rates and it was confirmed that power

uprate has no significant effect on flow induced erosion or

corrosion. The affected systems are currently monitored by the Flow

Accelerated Corrosion (FAC) program that addresses erosion and

corrosion concerns. Continued monitoring of the systems provides a

high level of confidence in the integrity of potentially susceptible

high energy piping systems.

Plant systems and components have been verified to be capable of

performing their intended design functions at uprated power

conditions. Where necessary, some components will be modified prior

to implementation of uprated power conditions to accommodate the

revised operating conditions. The review has concluded that

operation at power uprate conditions will not affect the reliability

of plant equipment, and that current Technical Specifications (TS)

surveillance requirements ensure adequate monitoring of system

operability. Systems continue to be operated in accordance with

current design requirements under uprated conditions, therefore no

new components or system interactions were identified that could

lead to an increase in accident probability. Changes to reactor

scram setpoints are such that no significant increase in scram

frequency due to operation at uprated conditions will occur.

B. Evaluation of the Consequences of Previously Evaluated Accidents

The radiological consequences due to the Loss of Coolant

Accident (LOCA) were calculated and are found to be below the

applicable regulatory limits. The results are presented in Table 9-3

of Attachment E [of the July 14, 1999 submittal].

The LOCA radiological consequences have not significantly

increased due to power uprate, and radiological consequences

continue to meet established regulatory limits.

The radiological evaluations for other non-LOCA Design Basis

Accidents (DBAs) were also performed and the dose consequences for

these events did not significantly increase. These changes are

outlined in Section 9.2 of Attachment E and they demonstrate that

LaSalle County Station (LCS), Units 1 and 2 still meets the

applicable regulatory limits.

Non-DBA Radiological Doses

All of the other radiological releases discussed in Updated

Final Safety Analysis Report (UFSAR) are either unchanged because

they are not power-dependent, or increase approximately in linear

proportion to the amount of the uprate. The dose consequences for

all of the non-LOCA radiological release accident events did not

significantly increase, and are bounded by the ``LOCA Radiological

Consequences''

[[Page 46428]]

events discussed above and were shown to meet the current dose

acceptance limits. These events are discussed in Section 9.2 of

Attachment E.

Therefore, the proposed change does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

Does the change create the possibility of a new or different

kind of accident from any accident previously evaluated?

The configuration, operation and event response of the LCS,

Units 1 and 2 systems, structures or components are [unchanged] by

operation at uprated power conditions. Analysis of transient events

has confirmed that the same transients remain limiting and that no

transient event results in a new sequence of events that could lead

to a new accident scenario.

An increase in power level will not create a new fission product

release path, or result in a new fission product barrier failure

mode. The current fission product barriers consisting of the reactor

fuel rod cladding, the reactor coolant pressure boundary, and the

containment structure remain in place. Fuel rod cladding integrity

is ensured by operating within thermal, mechanical, and exposure

design limits, and was confirmed for a representative core by

performance of transient and accident analysis. Cycle specific

analysis will continue to be performed for each fuel reload to

demonstrate the compliance with the applicable transient analysis

criteria and to establish the cycle specific Minimum Critical Power

Ratio (MCPR) safety limit and fuel operating limits. The integrity

of the reactor coolant pressure boundary was confirmed by evaluation

of the bounding overpressurization event and ensuring that the

corresponding pressure remained below the American Society of

Mechanical Engineers (AMSE) Boiler and Pressure Vessel (B&PV) Code,

Section III, ``Rules for Construction of Nuclear Power Plant

Components,'' overpressure protection requirements. Similarly,

analysis of the primary containment structure has demonstrated under

worst case design basis accident conditions that the containment

structure remains below the containment design pressure.

The effect of operation at uprated conditions on plant equipment

has been evaluated. No new operating mode, safety-related equipment

lineup, accident scenario, or equipment failure mode was identified

as a result of operating at uprated conditions. In addition,

operation at power uprated conditions does not create any new

sequence of events or failure modes that lead to a new type of

accident. Plant modifications required to support implementation of

power uprated conditions will be made to existing systems rather

than by adding new systems of a different design, which might

introduce new failure modes or accident sequences.

Therefore, the proposed change does not create the possibility

of a new or different kind of accident from any previously

evaluated.

Does the change involve a significant reduction in a margin of

safety?

The power uprate analysis for LCS, Units 1 and 2 assures that

the power dependent safety margin will be maintained by meeting the

appropriate regulatory criteria as prescribed by the applicable

regulations. Similarly, factors of safety specified by application

of the regulatory required design rules have been maintained, as

have other acceptance criteria used to judge the acceptability of

current plant operation.

No change is required in the basic duel deign to achieve the

uprated power levels, or to maintain current operating and safety

margins. No increase in the allowable peak bundle power is requested

as a result of operation at uprated conditions. The abnormal

transients have been evaluated for a representative core

configuration and confirmed that operation at uprated conditions

does not have an adverse effect on the operating limit MCPR. No

change to the Safety Limit MCPR results, thus the margin of safety

as assured by the safety limit MCPR is maintained. The fuel

operating limits related to heat generation rate would still be met

at uprated conditions. Cycle specific analysis will continue to be

performed for each fuel reload to demonstrate the compliance with

the applicable transient analysis criteria and to establish the

cycle specific safety limit and fuel operating limits.

The Emergency Core Cooling System (ECCS)-LOCA performance has

been evaluated at power uprated conditions using methodologies that

have been approved by the NRC for 10CFR50.46, ``Acceptance Criteria

for Emergency Core Cooling Systems for Light-Water Nuclear Power

Reactors,'' analysis. The current ECCS performance requirements were

used in the power uprate analysis. The ECCS-LOCA analysis was

conducted at 102% of the proposed uprated thermal power in

accordance with regulatory guidance. The necessary analysis for

operation of General Electric (GE) fuel under uprated conditions and

the determination that the peak cladding temperature (PCT) remains

below the 10CFR50.46 limit of 2200 deg.F have been performed.

However, LCS Unit 2 currently contains a mixed core of GE and

Siemens Power Corporation (SPC) fuel. LCS obtained [a] TS amendment

that allows operation with SPC fuel, and approved the use of the SPC

analytical methodology. The ECCS-LOCA analysis performed to support

use of the SPC fuel was conducted at a power level that bounds 102%

of the proposed uprated power level and determined that the PCT, for

SPC fuel, remains below the 10CFR50.46 limit of 2200 deg.F. The

analysis for both GE and SPC fuel types demonstrate all 10CFR50.46

criteria are met. Therefore, there is no reduction in margin with

respect to maintaining ECCS performance.

The margin of safety of the reactor coolant pressure boundary is

maintained under power uprated conditions. The design pressure of

the RPV and reactor pressure coolant pressure boundary remains at

1250 psig. The ASME B&PV Code allowable peak pressure is 1375 psig

(i.e., 110% of design value), which is the acceptance limit for

pressurization events. The limiting pressurization event is a Main

Steam Isolation Valve (MSIV) closure with a failure of valve

position scram and this event results in a calculated peak RPV

pressure of 1332 psig at the bottom of the RPV. The peak pressure

remains below the 1375 psig ASME limit. Therefore, there is no

decrease in margin of safety in the reactor coolant pressure

boundary.

The margin of safety of the containment structure is maintained

under power uprated conditions. The analyses were conducted using a

newer NRC-reviewed methodology. The pre-uprated cases were run using

the new methodology and the re-baselined cases were compared to the

uprated cases. The short-term containment peak pressure analysis re-

baseline result was 39.3 psig compared to the original analysis of

39.6 psig. At uprated conditions the peak containment drywell

pressure would be 39.9 psig, and is below the design value of 45

psig. The long-term containment suppression pool temperature

analysis re-baseline result was 190 deg.F compared to the original

analysis result of 200 deg.F. At uprated conditions the analysis

concluded that in the event of a LOCA, the calculated peak bulk

suppression pool temperature would be 193 deg.F. This is less than

the design temperature of the suppression pool of 275 deg.F, and the

criteria used to ensure adequate Net Positive Suction Head (NPSH) to

the ECCS pumps which is 212 deg.F. Therefore, power uprate does not

challenge the structural integrity of the containment structure and

ECCS NPSH is assured.

Therefore, operation at power uprated conditions 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

requested amendments involve no significant hazards consideration.

Local Public Document Room location: Jacobs Memorial Library, 815

North Orlando Smith Avenue, Illinois Valley Community College, Oglesby,

Illinois 61348-9692

Attorney for licensee: Ms. Pamela B. Stroebel, Senior Vice

President and General Counsel, Commonwealth Edison Company, P.O. Box

767, Chicago, Illinois 60690-0767

NRC Section Chief: Anthony J. Mendiola.

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

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

Date of amendment request: August 6, 1999

Description of amendment request: The proposed amendments would

revise Technical Specification 3/4.6.4, ``Vacuum Relief'' to remove

specific operability requirements related to position indication for

the suppression chamber-drywell vacuum breakers. The amendments also

reformat the action statements for inoperable vacuum breakers, increase

the surveillance

[[Page 46429]]

interval for verifying that the vacuum breakers are closed, and delete

the requirement to verify that the manual isolation valves are closed

for an inoperable and open vacuum breaker.

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:

Do the proposed changes involve a significant increase in the

probability or consequences of an accident previously evaluated?

The proposed changes do not change the hardware configuration of

the suppression chamber-drywell vacuum breakers, and the vacuum

breakers are not considered an initiator in any accident scenario.

The removal of specific indication requirements and the extension of

the surveillance interval does not impact the ability of the vacuum

breakers to perform their safety function. The vacuum breakers

continue to meet their intended design function. The proposed

changes do not impact the assumed source term for any analyzed

accident. Therefore, no increases in the probability of an accident

or consequences will result due to this proposed change.

Do the proposed changes create the possibility of a new or

different kind of accident from any accident previously evaluated?

The proposed changes do not involve any physical alterations to

the suppression chamber-drywell vacuum breakers, or cause any

changes in the method by which the vacuum breakers or the

containment vacuum relief system performs their associated design

basis functions. Therefore, the proposed changes do not create the

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

previously evaluated.

Do the proposed changes involve a significant reduction in a

margin of safety?

The proposed changes do not impact the design function assumed

for the containment vacuum relief system. The proposed changes do

not require the vacuum breakers to operate in a condition not

previously assumed in the facility accident analysis. The

containment vacuum relief system will continue to operate and

provide the protection assumed in the accident analysis. In order to

limit bypass, the vacuum breakers are in a normally closed position.

These vacuum breakers cannot be permanently placed in the open

position. The proposed decrease in the surveillance frequency

verifying the closed vacuum breakers will not increase the risk of

the vacuum breakers being in the open position, since they will only

open in response to a pressure differential or manual cycling.

Therefore, the assurance of the operability of the containment

vacuum breakers would be the same as provided under current

Technical Specifications. The containment response analysis is

unchanged, in that the vacuum breakers protect the containment

structure, the peak containment pressure remains as calculated, and

the vacuum breakers continue to maintain bypass leakage rates as

assumed. Therefore this proposed change does not cause a reduction

in the margin of safety.

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

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

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

requested amendments involve no significant hazards consideration.

Local Public Document Room location: Jacobs Memorial Library, 815

North Orlando Smith Avenue, Illinois Valley Community College, Oglesby,

Illinois 61348-9692.

Attorney for licensee: Ms. Pamela B. Stroebel, Senior Vice

President and General Counsel, Commonwealth Edison Company, P.O. Box

767, Chicago, Illinois 60690-0767.

NRC Section Chief: Anthony J. Mendiola.

Commonwealth Edison Company, Docket Nos. 50-254 and 50-265, Quad Cities

Nuclear Power Station, Units 1 and 2, Rock Island County, Illinois

Date of amendment request: July 16, 1999.

Description of amendment request: The proposed change to Technical

Specification Section 3/4.7.D is to eliminate the limit for any one

main steam line isolation valve (MSIV) leakage of less than or equal to

11.5 standard cubic feet per hour (scfh), and to replace that with an

aggregate value of less than or equal to 46 scfh for all four MSIVs.

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:

Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The proposed changes to the Technical Specifications, Appendix

A, modifies the allowed leakage limit to an aggregate value with no

change to the total allowed leakage rate. This change does not

affect either the automatic or manual features that would close the

MSIVs. There are no physical changes to the plant and plant

operations remain unchanged. Therefore, this proposed amendment does

not involve a significant increase in the probability or

consequences of an accident previously evaluated.

Does the change create the possibility of a new or different

kind of accident from any accident previously evaluated?

The safety function of the MSIVs is to provide a timely steam

line isolation to mitigate the release of radioactive steam and

limit reactor inventory loss under certain accident and transient

conditions. The MSIVs are designed to automatically close whenever

plant conditions warrant main steam line isolation. Changing the

leakage limits to include an aggregate value does not affect the

isolation function. No new equipment will be installed or utilized,

and no new operating conditions will be initiated as a result of

this change. Therefore, the proposed change does not create the

possibility of a new or different kind of accident from any

previously evaluated.

Does the change involve a significant reduction in a margin of

safety?

The total allowed leakage rate for all MSIVs remains unchanged

at 46 scfh. Therefore, there will be no change in the types or

significant increase in the amounts of any effluents released

offsite, and, thus, the radiological analyses remain unchanged and

within the guidelines of 10 CFR 100 and General Design Criteria 19.

Therefore, these changes do not involve a significant reduction in

the margin of safety.

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

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

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

requested amendments involve no significant hazards consideration.

Local Public Document Room location: Dixon Public Library, 221

Hennepin Avenue, Dixon, Illinois 61021.

Attorney for licensee: Ms. Pamela B. Stroebel, Senior Vice

President and General Counsel, Commonwealth Edison Company, P.O. Box

767, Chicago, Illinois 60690-0767.

NRC Section Chief: Anthony J. Mendiola.

Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee

Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina

Date of amendment request: July 27, 1999.

Description of amendment request: The proposed amendments would add

a surveillance requirement to verify the Keowee out-of-tolerance logic

trips and blocks closure of the appropriate overhead or underground

power path breakers. This logic is being added as part of a

modification to provide voltage and frequency protection for the Keowee

Hydro Units to protect them from being exposed to out-of-tolerance

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:

[[Page 46430]]

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated:

This change does not create any conditions or events, which lead

to accidents previously, evaluated in the SAR. The Keowee Hydro

units are used for mitigation of loss of power scenarios. The

proposed changes do not change the current function of the Keowee

Hydro Units. Therefore, the proposed change does not involve a

significant increase in the probability of an accident previously

evaluated. The Keowee Hydro units and their role in the Oconee

emergency power system currently meet the design/licensing basis

requirements for the system. There is no adverse affect on

containment integrity and no new release paths are created. The

proposed changes do not cause any adverse effects to the Keowee

single failure design or adversely affect the Keowee start time of

23 seconds. Therefore, the proposed changes do not involve a

significant increase in the consequences of an accident previously

evaluated.

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

from any accident previously evaluated:

The Keowee Hydro units are used for mitigation of loss of power

scenarios. No accidents new or different than already evaluated in

the SAR are postulated as a result of the proposed change. No

setpoints for parameters, which initiate protective or mitigative

action, are being changed. Therefore, this proposed amendment does

not create the possibility of any new or different kind of accident.

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

The proposed change does not adversely affect any plant safety

limits, set points, or design parameters. The change also does not

adversely affect the fuel, fuel cladding, Reactor Coolant System, or

containment integrity. Therefore, the proposed change does not

involve a significant reduction in a margin of safety.

Duke has concluded, based on the above, that there are no

significant hazards considerations involved in this 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: Oconee County Library, 501

West South Broad Street, Walhalla, South Carolina.

Attorney for licensee: Anne W. Cottington, Winston and Strawn, 1200

17th Street, NW., Washington, DC.

NRC Section Chief: Richard L. Emch, Jr.

Energy Northwest, (formerly known as the Washington Public Power Supply

System), Docket No. 50-397, WNP-2, Benton County, Washington

Date of amendment request: July 29, 1999.

Description of amendment request: The proposed amendment would

change the applicability of Section 3.4.9 of the Technical

Specifications (TS) from ``Mode 3 with steam drum pressure less than

the RHR [residual heat removal] cut in permissive'' to ``Mode 3 with

steam drum pressure less than 48 psig.'' Notes associated with TS

Surveillance Requirements 3.4.9.1 and 3.5.1.2 would be changed to

reflect the proposed 48 psig limit.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

This change involves further restrictions on the use of RHR in

the shutdown cooling mode of operation during hot shutdown

conditions. Chapter 15 of the FSAR [Final Safety Analysis Report]

defines the start of hot shutdown as the point when generated power

is below one percent rated power. During entry into hot shutdown

conditions the RHR system will be aligned in the Low Pressure

Coolant Injection (LPCI) mode of operation. Thus, it will be aligned

to provide water to the Reactor Pressure Vessel in the event the

high pressure systems (HPCS and RCIC) are not able to perform this

function. The change being proposed here has no impact on loss of

coolant accidents (LOCAs) requiring mitigation using RHR aligned in

the LPCI mode of operation.

During the high pressure portion of the hot shutdown condition,

intersystem (LOCAs) are a concern. The purpose of the RHR SDC

Isolation Reactor Pressure--High (cut-in permissive) at 135 psig is

to prevent over-pressurization of portions of the RHR system. This

protection is not being modified by this change. The instrumentation

that provides this protection will continue to function as designed.

This change only impacts the applicability of Technical

Specification 3.4.9 and when RHR SDC is required to be operable.

During hot shutdown the reactor is normally cooled down through

use of the main steam system and the condenser. Other means of

cooling are also available using the reactor water cleanup system or

a combination of emergency core cooling system (ECCS) pumps and

safety relief valves (SRVs). The RHR system aligned in the SDC mode

is used at the end of this cooling process to reach cold shutdown

conditions of less than or equal to 200 deg.F. The change being

proposed results in the RHR SDC being manually initiated at a lower

pressure and temperature. This change will have no significant

impact on the capability to cool the reactor.

FSAR Chapter 15, ``Accident Analysis,'' describes two events

associated with the RHR system. FSAR section 15.1.6, ``Inadvertent

Residual Heat Removal Shutdown Cooling Operation,'' describes the

impact of system operation during startup or cool-down when the

reactor is near critical. The proposed change involves the point at

which RHR is started in the SDC mode with the reactor sub-critical

with control rods inserted. Therefore, there will be no change in

the probability or consequences of this accident.

FSAR section 15.2.9, ``Failure of Residual Heat Removal Shutdown

Cooling,'' describes the failure of the RHR system to function in

SDC mode. This evaluation assumes a failure of the SDC mode of

operation but does not disable the remaining modes of RHR operation.

The alternate shutdown cooling paths involve the use of the SRVs

[safety relief valves] to establish a cooling path to the

containment suppression pool. This evaluated accident does not

result in any fuel failure. The proposed change will not result in

any fuel failures. The evaluated accident does result in normal

coolant activity being released to the suppression pool through the

safety relief valves. The proposed activity will not result in a

significant change in the release of this coolant activity.

The proposed change will not cause a significant increase in the

probability of a loss of SDC accident. This change proposes a delay

in the use of SDC because of temperature limitations. During this

time other means of decay heat removal would be used. This will

result in a decrease in use of RHR in SDC mode and a decrease in the

probability of failure of the system by restricting operation to be

within analyzed temperature limits. The proposed change will not

involve a significant increase in the consequences of the loss of

shutdown cooling accident. The accident evaluated in the FSAR

assumes SDC does not operate at any time and alternate means of

cooling are evaluated. Section 15.2.9.6 states there is no fuel

failure and release is limited to normal primary coolant activity to

the suppression pool. The proposed change results in a short delay

in the use of SDC because of temperature limitations. The accident

described in FSAR section 15.2.9 bounds this condition and, as a

result, there will be no increase in accident consequences.

With multiple means of reactor water makeup and heat removal

available the restriction in the use of RHR caused by this change

will not result in 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 change will not cause any new inadvertent shutdown

cooling startup, loss of water inventory or loss of cooling

accidents. New or different inadvertent RHR SDC startup accidents

are not possible because this change is only a further restriction

on when the system is operated. The LOCA accidents during Mode 3 are

[[Page 46431]]

bounded by the LOCAs defined for Modes 1 and 2. No new primary sytem

LOCAs can be initiated because of this change. The purpose of the

RHR cut-in permissive at 135 psig is to prevent overpressurization

of portions of the RHR system that could cause an intersystem LOCA.

This change will not result in a new or different kind of

intersystem LOCA because this is only a further restriction on RHR

SDC operation. The use of RHR in the SDC mode is restricted to

operation at a lower pressure and temperature but other systems are

available to remove the decay heat. No new or different accidents

are created because of this change.

The FSAR section 15.2.9 accident, ``Failure of Resident Heat

Removal Shutdown Cooling,'' is bounding for all other accidents

which postulate failure of the capability to remove decay heat. No

additional accidents resulting in the loss of decay heat removal

capability will be caused by this change.

Therefore, the operation of WNP-2 in accordance with the

proposed amendment will 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 amendment will increase the reliability of the RHR

system when operated in shutdown cooling mode by providing assurance

that the temperature limits of the piping and pipe supports will not

be exceeded. The ability to protect against an intersystem LOCA is

unchanged. The ability to remove decay heat from the reactor is not

changed by this modification as alternate means of heat removal are

available. Therefore, operation of WNP-2 in accordance with the

proposed amendment will not involve a reduction in the margin of

safety.

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

this review, it appears that the three standards of 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: Richland Public Library, 955

Northgate Street, Richland, Washington 99352.

Attorney for licensee: Perry D. Robinson, Esq., Winston & Strawn,

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

NRC Section Chief: Stephen Dembek.

Energy Northwest (formerly known as the Washington Public Power Supply

System), Docket No. 50-397, WNP-2, Benton County, Washington

Date of amendment request: July 29, 1999.

Description of amendment request: The proposed amendment would

revise Technical Specification Table 3.3.5.1-1, ``Emergency Core

Cooling System (ECCS) Instrumentation Items 1.a, 2.a, 4.a and 5.a,'' to

change the Reactor Vessel Water Level--Low Low Low, Level 1 allowable

value from the current value of -148 inches to a new value of -142.3

inches.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

This change involves the measurement of water level in the

Reactor Pressure Vessel (RPV) used to initiate the ECCS. The

accident evaluated for this condition is the spectrum of loss of

coolant accidents (LOCA) severe enough to decrease the RPV water

inventory by a significant amount.

The additional uncertainty introduced because of harsh

environmental effects could not be accommodated between the existing

Technical Specification allowable value and the analytical limit.

This uncertainty results in a requirement that the ECCS be initiated

at a slightly higher water level than previously calculated.

Therefore, operation of WNP-2 in accordance with the proposed

amendment will 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 change will not create a new or different kind of

accident since it only makes a small change in the RPV water level

at which the ECCS is initiated. This change is in the conservative

direction requiring a greater volume of water in the RPV to

accommodate the uncertainty associated with the harsh environment of

the water level sensors.

The level indicating switches are located on instrument racks in

the Reactor Building. The harsh environment in this building would

have no impact on the initial trip needed to initiate the ECCS on

loss of RPV level since conditions in the Reactor Building would be

benign at the initial stages of the accident. Only if the Level 1

trip was reset and initiated after a significant period of time

would the harsh environmental conditions have an impact on the

accuracy of the level indicating switches. However, increasing the

water level at which the ECCS is initiated results in a more

conservative value that adequately includes post-accident harsh

environment uncertainties and ensures that the associated analytical

limit is met.

Therefore, the operation of WNP-2 in accordance with the

proposed amendment will 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 amendment increases the allowable value for water

level in the RPV. This small increase will result in an increase in

the margin of safety. A review of the plant settings for the Level 1

trip indicated that previous settings were within the new allowable

value.

Therefore, operation of WNP-2 in accordance with the proposed

amendment will not involve a significant reduction in the margin of

safety.

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

this review, it appears that the three standards of 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: Richland Public Library, 955

Northgate Street, Richland, Washington 99352.

Attorney for licensee: Perry D. Robinson, Esq., Winston & Strawn,

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

NRC Section Chief: Stephen Dembek.

Energy Northwest (formerly known as the Washington Public Power Supply

System), Docket No. 50-397, WNP-2, Benton County, Washington

Date of amendment request: July 29, 1999.

Description of amendment request: The proposed amendment would

revise Technical Specification Surveillance Requirement (SR) 3.5.2.2.

This requirement verifies the adequacy of the water supply in the

condensate storage tanks (CSTs) which support operation of the high

pressure core spray (HPCS) system during Modes 4 and 5. Current

Technical Specification SR 3.5.2.2 requires that CST water level be

maintained above 13.25 feet in a single tank or above 7.6 feet in each

tank if the suppression pool level is below its minimum level. It is

proposed that the CST water level be maintained above 14.8 feet in a

single tank or above 9.1 feet in each 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.

During Modes 4 and 5 HPCS may be required to provide water to

the reactor vessel if the water level decreases. The revised

condensate storage tank allowable levels increase the operating

margins by providing an increased water inventory. The previously

evaluated accident involving the loss of decay heat cooling

inventory will not have an increase in probability because the

inventory of water will be increased with the change being proposed.

[[Page 46432]]

The consequences of any accident involving the loss of decay

heat cooling inventory will not change as the consequences are

unaffected by the increased water inventory.

Therefore, operation of WNP-2 in accordance with the proposed

amendment will 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 change will not create a new or different kind of

accident since it only increases the amount of water held in reserve

to support reactor vessel inventory loss. The proposed change does

not introduce any credible mechanisms for unacceptable radiation

release nor does it require physical modification to the plant. The

inventory of water in the CSTs will increase to support any loss of

water inventory in the reactor vessel during shutdown.

The proposed change modifies the monitored values for CST level.

The plant has operated well within the existing allowable values.

The increased margin provided by the increased level will assure no

new or different kinds of accidents result from the proposed change.

Therefore, the operation of WNP-2 in accordance with the

proposed amendment will 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 amendment increases the allowable value for water

level in the CSTs. This results in an increase in the inventory of

water available for cooling and inventory control during reactor

shutdown. This will result in an increase in the margin of safety.

Therefore, operation of WNP-2 in accordance with the proposed

amendment will not involve a significant reduction in the margin of

safety.

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

this review, it appears that the three standards of 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: Richland Public Library, 955

Northgate Street, Richland, Washington 99352.

Attorney for licensee: Perry D. Robinson, Esq., Winston & Strawn,

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

NRC Section Chief: Stephen Dembek.

Energy Northwest (formerly known as the Washington Public Power Supply

System), Docket No. 50-397, WNP-2, Benton County, Washington

Date of amendment request: July 29, 1999.

Description of amendment request: The proposed amendment request

would revise Technical Specification Surveillance Requirement (SR) SR

3.8.4.6 of Technical Specification 3.8.4, ``DC Sources--Operating,''

and SR 3.8.5.1 of Technical Specification 3.8.5, ``DC Sources--

Shutdown.'' The proposed change to SR 3.8.4.6 would prohibit

surveillance testing of Division 1, 2, and 3 125 and 250 volt DC,

battery charger capacity during Modes 1, 2, and 3. However, credit

could be taken for unplanned events that satisfied the surveillance

requirement. The proposed change to SR 3.8.5.1 would include SR 3.8.4.6

as one of the surveillance tests that are not required to be performed.

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 has no impact on previously analyzed

accidents or transients, and has no effect on operation, capacity or

surveillance test details of the DC system battery chargers. The

change only imposes a mode restriction on performance of specified

surveillance testing and allows taking credit for unplanned events

that satisfy the surveillance. Therefore, operation of WNP-2 in

accordance with the proposed amendment will 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 change has no effect on operation, capacity, or

surveillance test details of the DC system battery chargers. The

change only prohibits performing specified battery charger capacity

surveillance testing from being implemented during Mode 1, 2, or 3

and allows taking credit for unplanned events that satisfy the

surveillance. The proposed change to SR 3.8.4.6 of Technical

Specification 3.8.4 and SR 3.8.5.1 of Technical Specification 3.8.5

are consistent with the wording previously evaluated and approved by

the NRC in NUREG-1434 Rev. 1.

Therefore, operation of WNP-2 in accordance with the proposed

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

of accident previously evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The proposed change only imposes a mode restriction, prohibiting

battery charger capacity surveillance testing from being performed

during Modes 1, 2, and 3, allowing credit to be taken for unplanned

events that satisfy the surveillance, and allowing such testing to

be omitted under certain conditions during Modes 4 and 5 and during

movement of irradiated fuel in secondary containment. Performance of

this testing would remove a DC electrical power subsystem from

service and could present a safety risk were an event to occur if

the testing was performed in Modes 1, 2, and 3, or while DC service

is required in other operating conditions. Therefore, operation of

WNP-2 in accordance with the proposed amendment will not involve a

significant reduction in the margin of safety.

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

this review, it appears that the three standards of 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: Richland Public Library, 955

Northgate Street, Richland, Washington 99352.

Attorney for licensee: Perry D. Robinson, Esq., Winston & Strawn,

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

NRC Section Chief: Stephen Dembek.

Entergy Operations, Inc., System Energy Resources, Inc., South

Mississippi Electric Power Association, and Entergy Mississippi, Inc.,

Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne

County, Mississippi

Date of amendment request: July 20, 1998, as supplemented June 29,

1999.

Description of amendment request: The amendment would incorporate

the Technical Specification changes necessary for implementation of the

Boiling Water Reactor Owners' Group Reactor Stability Long-Term

Solution, Enhanced Option 1-A (E1A). E1A consists of modifications to

the plant operating procedures and associated plant components that

provide a means for reliably detecting and avoiding reactor

instabilities. By letter dated February 25, 1998, the Nuclear

Regulatory Commission (NRC) staff recognized E1A as a technically

acceptable implementation of a long-term stability solution satisfying

the requirements of NRC IE Bulletin 88-07, Supplement 1, and Generic

Letter 94-02, ``Long Term Solutions and Upgrade of Interim Operating

Recommendations for Thermal-Hydraulic Instabilities in Boiling Water

Reactors.''

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:

[[Page 46433]]

1. This request does not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The proposed amendments allow the implementation of the Enhanced

Option I-A (E1A) long term solution to the neutronic/thermal-

hydraulic instability issue. Current Technical Specification (TS)

restrictions on power and flow conditions, number of operating

recirculation loops and operator actions implemented to reduce the

probability of neutronic/thermal-hydraulic instability are

eliminated and new stability requirements consistent with NEDO-

32339-A, Supplement 4, Revision 1, are imposed. These requirements

include restrictions on power and flow conditions and actions

associated with the modified Average Power Range Monitor (APRM) flow

biased scram and control rod block functions. Required actions

include adherence to the boiling boundary limit stability control

prior to entry and during operation in the region of the power and

flow operating domain which is potentially susceptible to neutronic/

thermal-hydraulic instability in the absence of the stability

control. In addition, the proposed amendments require operator

actions based upon control room indications generated by a new

Period Based Detection System (PBDS). The PBDS is designed to

provide alarm indication that conditions consistent with a

significant degradation in the stability performance of the reactor

has occurred and the potential for imminent onset of neutronic/

thermal-hydraulic instability may exist. The PBDS also provides

analog indication of the highest and second highest successive

period confirmation count of all of the Local Power Range Monitors

(LPRMs) monitored. This provides the plant operators with continuous

indication of reactor stability operating conditions.

The proposed amendments will permit operation in regions of the

power and flow operating domain postulated to be susceptible to

neutronic/thermal-hydraulic instability. Operation in these regions

does not increase the probability of occurrence of initiators and

precursors of previously analyzed accidents when neutronic/thermal-

hydraulic instability is not possible. The proposed amendments

permit the implementation of the features of the E1A solution which

prevent neutronic/thermal-hydraulic instability including preemptive

reactor scram upon entry into the regions of the power and flow

operating domain most susceptible to neutronic/thermal-hydraulic

instability. The E1A solution also requires implementation of

stability control prior to entry into a region of the power and flow

operating domain which is potentially susceptible, in the absence of

stability control, to neutronic/thermal-hydraulic instability. The

E1A solution prevents neutronic/thermal-hydraulic instability during

operation in regions of the power and flow operating domain

previously excluded from operation and therefore does not

significantly increase the probability of a previously analyzed

accident.

Operation in the regions of the power and flow operating domain

excluded by current TS 3.4.1 and Figure 3.4.1-1 can occur as a

result of anticipated operational occurrences. The severity of these

transients may increase in the absence of operator actions due to

the potential occurrence of neutronic/thermal-hydraulic instability

as a result of operation in these regions. The proposed amendments

will permit the implementation of the E1A long term solution to the

stability issue. Required features of the E1A solution include

adherence to a boiling boundary limit stability control prior to

selection by the operator of APRM flow biased scram and control rod

block function ``Setup'' setpoints which allow operation in a region

of the power and flow operating domain potentially susceptible, in

the absence of the stability control, to neutronic/thermal-hydraulic

instability. Upon entry, as a result of an anticipated operational

occurrence, into the region most susceptible to neutronic/thermal-

hydraulic instability, the preemptive reactor scram prevents

neutronic/thermal-hydraulic instability. Therefore, the consequences

of an accident do not significantly increase while operating with

the stability control met.

After exiting the region requiring the stability control to be

met, the setpoints can be manually reset to their normal values.

Stability controls are required to be in place when setpoints are

``Setup''. As a backup E1A feature, the APRM flow biased setpoints

automatically reset to their normal values above a pre-determined

flow condition. This automatic reset to the more conservative

setpoints ensures that the preemptive reactor scram will prevent

operation as a result of an anticipated operational occurrence into

the region most susceptible to neutronic/thermal-hydraulic

instability should the operator not select the more conservative

setpoints appropriate for operation following exit from the region

requiring stability control.

Other required E1A features, including the PBDS, control rod

block alarms associated with entry into the region susceptible to

neutronic/thermal-hydraulic instabilities in the absence of

stability controls, and required operator actions, including manual

reactor scram, help ensure prevention of neutronic/thermal-hydraulic

instabilities. Therefore, the proposed amendments prevent the

occurrence of neutronic/thermal-hydraulic instability as a

consequence of an anticipated operational occurrence and do not

significantly increase the consequences of any previously analyzed

accident.

2. This request does not create the possibility of a new or

different kind of accident from any accident previously evaluated.

The proposed amendments replace current restrictions on power

and flow conditions with alternative restrictions which permit the

implementation of the E1A long term stability solution. The current

restrictions on the power and flow conditions and operating

recirculation loops in the RUN mode do not automatically prevent the

entry into regions of the power and flow operating domain most

susceptible to neutronic/thermal-hydraulic instability and therefore

the possibility of neutronic/thermal-hydraulic instability exists in

the absence of operator action. The required features of the E1A

solution implement a preemptive scram upon entry into the region

most susceptible to neutronic/thermal-hydraulic instability, without

operator action. The accessible operating domain allowed by the

proposed amendments is a subset of the power and flow operating

domain currently allowed. Current initiators and precursors of

accidents and anticipated operational occurrences [cannot] occur

with new or different initial conditions as a result of this change.

Additionally, there are no new event initiators or precursors of

accidents and anticipated operational occurrences created by this

change. Therefore, the proposed amendments do not create the

possibility of a new or different kind of accident from that

previously evaluated.

Concurrent with the implementation of the proposed amendments, a

modified Flow Control Trip Reference (FCTR) card, the E1A FCTR card,

and a new Period Based Detection System (PBDS) will be installed as

required by the E1A solution. The function of the E1A FCTR card is

to aid the operator in the identification of entry into regions of

the power and flow operating domain potentially susceptible to

neutronic/thermal-hydraulic instability in the absence of stability

controls and to initiate a preemptive scram upon entry into the

regions most susceptible to neutronic/thermal-hydraulic instability.

This is accomplished by altering the existing values of setpoints of

the APRM flow biased scram and the control rod block functions

generated by the E1A FCTR card. The E1A FCTR card design includes

components which may be susceptible to electromagnetic interference

or other environmental effects. The plant specific environmental

conditions (temperature, humidity, pressure, seismic, and

electromagnetic compatibility) have been confirmed to be enveloped

by the environmental qualification values for the E1A FCTR cards.

Therefore, the potential for spurious scrams or common mode failures

induced by environmental effects (e.g., electromagnetic

interference) is considered negligible. The installation of the E1A

FCTR card will therefore not create the possibility of a new or

different kind of accident from any accident previously evaluated.

The function of the PBDS is to provide the operator with an

indication that conditions consistent with a significant degradation

in the stability performance of the reactor has occurred and the

potential for imminent onset of neutronic/thermal-hydraulic

instability may exist. This is accomplished by the installation of a

new PBDS card in the Neutron Monitoring System. The PBDS card takes

inputs from individual local power range monitors and provides

analog indication of the highest and second highest successive

period confirmation count, provides a High Decay Ratio (Hi DR) and

High-High Decay Ratio (Hi-Hi DR) alarms, and INOP status indication

to the operator in the control room. These displays [cannot] create

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

accident previously evaluated. The PBDS card design includes

components which may be susceptible to electromagnetic interference

or other environmental effects. However, the plant specific

environmental conditions (temperature, humidity, pressure, seismic,

[[Page 46434]]

and electromagnetic compatibility) have been confirmed to be

enveloped by the PBDS environmental qualification values. Therefore,

the installation of the PBDS card will not create the possibility of

a new or different kind of accident from any accident previously

evaluated.

3. This request does not involve a significant reduction in a

margin to safety.

The proposed amendments permit the implementation of the E1A

long term solution to the stability issue. Under certain conditions,

existing BWR [boiling water reactor] designs are susceptible to

neutronic/thermal-hydraulic instability. General Design Criterion

(GDC) 12 of 10 CFR 50, Appendix A, requires thermal-hydraulic

instability to be prevented by design or be readily and reliably

detected and suppressed. When the design of the reactor system does

not prevent the occurrence of neutronic/thermal-hydraulic

instability, instability is an anticipated operational occurrence.

GDC 10 of 10 CFR 50, Appendix A, requires that specified acceptable

fuel design limits not be exceeded during anticipated operational

occurrences.

Analyses performed by the BWROG [Boiling Water Reactor Owners'

Group] indicate that neutronic/thermal-hydraulic instability induced

power oscillations could result in conditions exceeding the Minimum

Critical Power Ratio (MCPR) Safety Limit (SL) prior to detection and

suppression by the current design of the Neutron Monitoring System

and Reactor Protection System.

To ensure compliance with GDC 12 the BWROG developed Interim

Corrective Actions (ICAs) to enhance the capability of the operator

to readily and reliably detect and suppress neutronic/thermal-

hydraulic instability. The BWROG ICAs also provided additional

guidance for monitoring local power range monitors beyond the

requirements of current TS 3.4.1 to ensure adequate margin to the

onset of neutronic/thermal-hydraulic instability. Reliance on

operator actions to comply with GDC 12 was accepted on an interim

basis by the NRC pending final implementation of a long term

solution to the stability issue. Neutronic/thermal-hydraulic

instability is prevented by implementation of the E1A solution

through the modified design of the Reactor Protection System (APRM

[average power range monitor] flow biased scram) and the stability

control prior to entry into a region of the power and flow operating

domain which is potentially susceptible, in the absence of stability

control, to neutronic/thermal-hydraulic instability. In addition,

significant backup protection features, including the PBDS, control

rod block alarms associated with entry into the region susceptible

to neutronic/thermal-hydraulic instabilities in the absence of

stability controls, and specified operator actions, including manual

reactor scram, are required to be implemented. As a result, the

margin to the onset of neutronic/thermal-hydraulic instability

provided by the existing TS requirements and BWROG ICAs

recommendations is not significantly reduced by the implementation

of the E1A solution. The E1A solution assures compliance with GDC 12

by the prevention of neutronic/thermal-hydraulic instability and

therefore precludes neutronic/thermal-hydraulic instability from

becoming a credible consequence of an anticipated operational

occurrence. The consequences of anticipated operational occurrences

will not increase and the margin to the MCPR SL will not decrease

upon implementation of the E1A solution. Therefore, the proposed

amendments do not involve a significant reduction in a margin of

safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room Location: Judge George W. Armstrong

Library, 220 S. Commerce Street, Natchez, Mississippi 39120

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

Strawn, 1400 L Street, NW., 12th Floor, Washington, DC 20005-3502.

NRC Section Chief: Robert A. Gramm.

Entergy Operations, Inc., System Energy Resources, Inc., South

Mississippi Electric Power Association, and Entergy Mississippi, Inc.,

Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne

County, Mississippi

Date of amendment request: May 6, 1999.

Description of amendment request: The proposed amendments would

change those Technical Specifications (TS) required to support Grand

Gulf Nuclear Station (GGNS), Cycle 11 operation. The changes would

include a change to the minimum critical power ratio safety limit

(SLMCPR) that would reflect a decrease of the two recirculation loop

SLMCPR limit from 1.11 to 1.09, and the single recirculation loop

SLMCPR limit from 1.12 to 1.10. These values were developed with

General Electric's cycle-specific SLMCPR methodology in GESTAR-II

Amendment 25, which was recently approved by the Nuclear Regulatory

Commission in a Safety Evaluation Report dated March 11, 1999.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

I. The proposed change does not significantly increase the

probability or consequences of an accident previously evaluated.

The Minimum Critical Power Ratio (MCPR) safety limit is defined

in the Bases to Technical Specification 2.1.1 as that limit which

``ensures that during normal operation and during Anticipated

Operational Occurrences (AOOs), at least 99.9% of the fuel rods in

the core do not experience transition boiling.'' The MCPR safety

limit is re-evaluated for each reload and, for GGNS Cycle 11, the

analyses have concluded that a two-loop MCPR safety limit of 1.09,

based on the application of GE's [General Electric's] NRC-approved

cycle-specific MCPR safety limit methodology demonstrates that this

acceptance criterion is satisfied. For single-loop operation, a MCPR

safety limit of 1.10, based on GE's [NRC-approved cycle-specific

MCPR safety limit methodology, also demonstrates that this

acceptance criterion is satisfied. Core MCPR operating limits are

developed to support the Technical Specification 3.2 requirements

and ensure these safety limits are maintained in the event of the

worst-case transient. Since the MCPR safety limit will be maintained

at all times, operation under the proposed changes will ensure at

least 99.9% of the fuel rods in the core do not experience

transition boiling. Therefore, these changes to the Minimum Critical

Power Ratio (MCPR) safety limit do not affect the probability or

consequences of an accident.

GE's NRC-approved GESTAR-II cycle-specific MCPR safety limit

methodology has been applied and has no effect on the probability or

consequences of any accidents previously evaluated. As previously

licensed, one exception to GESTAR is that the mis-oriented and mis-

located bundle events will continue to be analyzed as accidents

subject to the acceptance criteria in the current licensing basis.

The design of the GE11 fuel bundles is such that the bundles are not

likely to be mis-oriented or mis-located and the normal

administrative controls will be in effect for assuring proper

orientation and location. Therefore, the probability of a fuel

loading error is not increased. This analysis ensures that

postulated dose releases will not exceed a small fraction (10

percent) of 10CFR100 limits. Therefore, the probability or

consequences of accidents previously evaluated are unchanged.

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

or different kind of accident from any accident previously

evaluated.

The GE11 fuel to be used in Cycle 11 is of a design compatible

with fuel present in the core and used in the previous cycle.

Therefore, the GE11 fuel will not create the possibility of a new or

different kind of accident. The proposed changes do not involve any

new modes of operation, any changes to setpoints, or any plant

modifications. The proposed revised MCPR safety limits have been

shown to be acceptable for Cycle 11 operation. Compliance with the

applicable criterion for incipient boiling transition continues to

be ensured. The proposed MCPR safety limits do not result in the

creation of any new precursors to an accident.

Therefore, the proposed changes do not create the possibility of

a new or different type of accident from any accident previously

evaluated.

III. The proposed change does not involve a significant

reduction in a margin of safety.

[[Page 46435]]

The MCPR safety limits have been evaluated in accordance with

GE's NRC-approved cycle-specific methodology to ensure that during

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

the core are not expected to experience transition boiling. One

exception to GESTAR is that the mis-oriented and mis-located bundle

events will continue to be analyzed as accidents subject to the

acceptance criteria in the current licensing basis. This analysis

ensures that postulated dose releases for the worst case mis-

oriented and mis-located bundle will not exceed a small fraction (10

percent) of 10CFR100 limits. On this basis, the implementation of

this GE methodology 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: Judge George W. Armstrong

Library, 220 S. Commerce Street, Natchez, Mississippi 39120.

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

Strawn, 1400 L Street, NW., 12th Floor, Washington, DC 20005-3502.

NRC Section Chief: Robert A. Gramm.

Entergy Operations, Inc., System Energy Resources, Inc., South

Mississippi Electric Power Association, and Entergy Mississippi, Inc.,

Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne

County, Mississippi

Date of amendment request: June 23, 1999.

Description of amendment request: The requested Technical

Specification changes would revise those specifications associated with

various engineered safety feature systems, which need no longer be

credited following a design-basis fuel handling accident. The proposed

changes affect conditions where irradiated fuel is handled in the

primary or secondary containment, and also affect certain

specifications related to performing core alterations. These changes

are based on the revised analysis of the design-basis fuel handling

accident for the Grand Gulf Nuclear Station. This requested change is

consistent with the changes approved for the Perry Nuclear Power Plant

Operating License (Amendment 102), and the industry-proposed change to

the Technical Specification NUREGs, TSTF-51.

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 significantly increase the

probability or consequences of an accident previously evaluated.

A new term to describe irradiated fuel is used to establish

operational conditions where specific activities represent

situations where significant radioactive releases can be postulated.

These operational conditions are consistent with the design basis

analysis. Because the equipment affected by the revised operational

conditions is not considered an initiator to any previously analyzed

accident, inoperability of the equipment cannot increase the

probability of any previously evaluated accident. The proposed

requirements bound the conditions of the current design basis fuel

handling accident analysis which concludes that the radiological

consequences are within the acceptance criteria of NUREG 0800,

Section 15.7.4 and General Design Criteria 19. Therefore, the

proposed changes do not significantly increase the probability or

consequences of any previously evaluated accident.

Removing a one time only allowance granted by Amendment 129 to

the Operating License that is no longer in affect is an

administrative change. Therefore, the proposed change does not

significantly increase the probability or consequences of any

previously evaluated accident.

Based on the above, neither the proposed changes to the

Technical Specifications nor that to the Operating License

significantly increase the probability or consequences of any

accident previously evaluated.

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

new or different kind of accident from any previous analyzed.

The new term to describe irradiated fuel is used to establish

operational conditions where specific activities represent

situations where significant radioactive releases can be postulated.

These operational conditions are consistent with the design basis

analysis. The proposed changes do not introduce any new modes of

plant operation and do not involve physical modifications to the

plant. Therefore, the proposed changes do not create the possibility

of a new or different kind of accident from any previous analyzed.

Removing a one time only allowance granted by Amendment 129 to

the Operating License that is no longer in affect is an

administrative change. Therefore, the proposed change does not

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

any previous analyzed.

Based on the above, neither the proposed changes to the

Technical Specifications nor that to the Operating License create

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

accident previously analyzed.

3. The proposed changes do not involve a significant reduction

in a margin of safety.

The new term to describe irradiated fuel is used to establish

operational conditions where specific activities represent

situations where significant radioactive releases can be postulated.

These operational conditions are consistent with the design basis

analysis and are established such that the radiological consequences

are at or below the current GGNS [Grand Gulf Nuclear Station]

licensing limit. Safety margins and analytical conservatisms have

been evaluated and are well understood. Substantial margins are

retained to ensure that the analysis adequately bounds all

postulated event scenarios. The proposed change only eliminates the

unnecessary margin from the analysis. The current margin of safety

is retained.

Specifically, the margin of safety for the fuel handling

accident is the difference between the 10CFR100 limits and the

licensing limit defined by NUREG 0800, Section 15.7.4. With respect

to the control room personnel doses, the margin of safety is the

difference between the 10CFR100 limits and the licensing limit

defined by 10CFR50, Appendix A, Criterion 19 (GDC 19). The

additional margin between the calculated doses for the postulated

events and the corresponding licensing limit provides no useful

purpose.

The proposed applicability continues to ensure that the whole-

body and thyroid doses at both the control room and the exclusion

area and low population zone boundaries are at or below the

corresponding licensing limit. The margin of safety is unchanged;

therefore, the proposed changes do not involve a significant

reduction in a margin of safety.

Removing a one time only allowance granted by Amendment 129 to

the Operating License that is no longer in affect is an

administrative change. Therefore, the proposed change does not

involve a significant reduction in a margin of safety.

Based on the above, neither the proposed changes to the

Technical Specifications nor that to the Operating License result in

a significant reduction in a margin of safety.

Based on the above evaluation, operation in accordance with the

proposed amendment involves no significant hazards considerations.

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: Judge George W. Armstrong

Library, 220 S. Commerce Street, Natchez, Mississippi 39120.

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

Strawn, 1400 L Street, NW., 12th Floor, Washington, DC 20005-3502.

NRC Section Chief: Robert A. Gramm.

Entergy Operations Inc., Docket No. 50-382, Waterford Steam Electric

Station, Unit 3, St. Charles Parish, Louisiana

Date of amendment request: October 6, 1998.

[[Page 46436]]

Description of amendment request: The proposed change modifies the

requirement to perform a Moderator Temperature Coefficient (MTC) test

near the end of each cycle. This request constitutes a lead-plant

submittal, submitted by Waterford 3 on behalf of the Combustion

Engineering Owners Group (CEOG). CE NPSD-911, Amendment 1, ``Analysis

of Moderator Temperature Coefficients in Support of a Change in the

Technical Specifications End of Cycle Negative MTC Limit'' dated

January, 1998 is provided as an Attachment to the application.

Specifically, the proposed change modifies Technical Specification (TS)

4.1.1.3.2c by adding a provision that eliminates the need to determine

the MTC upon reaching two-thirds of core burnup if the results of the

MTC tests required in TS 4.1.1.3.2a and 4.1.1.3.2b are within a

specified tolerance. In addition, some editorial changes are proposed

and the Bases change is included to support the changes in the TS.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Will operation of the facility in accordance with this

proposed change involve a significant increase in the probability or

consequences of an accident previously evaluated?

Response: No.

Under the proposed change, compliance with the TS Limiting

Condition for Operation is achieved through a surveillance program

consisting of beginning-of-cycle (BOC) measurements, plant parameter

monitoring, and end-of-cycle (EOC) MTC predictions. This change

eliminates the performance of the 2/3 Cycle MTC Surveillance when

the BOC MTC Surveillances are within a required tolerance of the

design value.

The probability and consequences of an accident previously

evaluated will not be increased because this change does not modify

any assumptions used in the input to the safety analyses. The

current safety calculations will remain valid because the allowed

range of MTC values will not change.

The Combustion Engineering analysis CE NPSD-911 and CE NPSD-911

Amendment 1, demonstrate that if the startup test program has

established that the core is operating as intended, and if the

isothermal temperature coefficients measured at zero power during

the cycle startup program, and at power prior to 40 EFPD [Effective

Full Power Days], fall within the design value of plus or minus

0.16 x 10-4 delta k/k/ deg.F, then the end-of-cycle best

estimate prediction will also be within plus or minus

0.16 x 10-4 delta k/k/ deg.F of true MTC.

Removing the footnote that was applicable during Cycle 7 and

providing a plus/minus for SR 4.1.1.3.2c is purely an administrative

change.

Therefore, the proposed change will not involve a significant

increase in the probability or consequences of any accident

previously evaluated.

2. Will operation of the facility in accordance with this

proposed change create the possibility of a new or different type of

accident from any accident previously evaluated?

Response: No.

Plant operation and plant parameter TS limits will remain

unchanged. There are no new changes in plant design nor are any new

failure modes introduced. CE NPSD-911 analysis determined that if

the MTC at the beginning-of-cycle is within plus or minus

0.16 x 10-4 delta k/k/ deg.F of the design value then the

MTC at the end-of-cycle will also be within plus or minus

0.16 x 10-4 delta k/k/ deg.F of the design value.

Removing the footnote that was applicable during Cycle 7 and

providing a plus/minus for SR 4.1.1.3.2c is purely an administrative

change.

Therefore, the proposed change will not create the possibility

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

evaluated.

3. Will operation of the facility in accordance with this

proposed change involve a significant reduction in a margin of

safety?

Response: No.

The margin of safety will not be reduced because the range of

allowed temperature coefficients will not be changed. The

surveillance program consisting of beginning-of-cycle measurements,

plant parameter monitoring, and end-of-cycle MTC predictions will

ensure that the MTC remains within the range of acceptable values.

Removing the footnote that was applicable during Cycle 7 and

providing a plus/minus for SR 4.1.1.3.2c is purely an administrative

change.

Therefore, the proposed change will not involve a significant

reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room Location: University of New Orleans

Library, Louisiana Collection, Lakefront, New Orleans, Louisiana 70122.

Attorney for licensee: N.S. Reynolds, Esquire, Winston & Strawn

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

NRC Section Chief: Robert A. Gramm.

FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse

Nuclear Power Station, Unit 1, Ottawa County, Ohio

Date of amendment request: July 26, 1999.

Description of amendment request: The proposed amendment would make

the following line-item Technical Specification (TS) improvements:

(1) Relocate TS Section 3/4.3.3.2, Instrumentation--Incore

Detectors; TS 3/4.3.3.9, Instrumentation--Waste Gas System Oxygen

Monitor; and TS 3/4.4.7, Reactor Coolant System `` Chemistry, to the

Updated Safety Analysis Report (USAR) Technical Requirements Manual

(TRM);

(2) Change to TS 3/4.11.2, Radioactive Effluents--Explosive Gas

Mixture, and TS Bases 3/4.11.2, Explosive Gas Mixture, to reflect the

above proposed relocation of TS 3/4.3.3.9;

(3) Revise the requirements of TS 3/4.4.6.1, Reactor Coolant System

Leakage--Leakage Detection Systems, to require one monitor (gaseous or

particulate) of the containment atmosphere radioactivity monitoring

systems to be operable, rather than requiring both systems to be

operable simultaneously; and

(4) Revise the requirements of TS 3/4.3.3.1, Radiation Monitoring

Instrumentation, to be consistent with the above proposed revision to

TS 3/4.4.6.1.

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 (DBNPS) has reviewed the

proposed changes and determined that a significant hazards

consideration does not exist because operation of the Davis-Besse

Nuclear Power Station, Unit Number 1, in accordance with these

changes would:

1a. Not involve a significant increase in the probability of an

accident previously evaluated because no accident initiator,

conditions or assumptions are affected by the proposed revisions to

Technical Specification (TS) 3/4.3.3.1, Radiation Monitoring

Instrumentation, TS 3/4.3.3.2, Instrumentation--Incore Detectors; TS

3/4.3.3.9, Instrumentation--Waste Gas System Oxygen Monitor; TS 3/

4.4.7, Reactor Coolant System--Chemistry; TS 3/4.11.2, Radioactive

Effluents--Explosive Gas Mixture; and TS 3/4.4.6.1, Reactor Coolant

System Leakage--Leakage Detection Systems, and their associated TS

Bases.

The requirements of TS 3/4.3.3.2, TS 3/4.3.3.9, and TS 3/4.4.7

are proposed to be relocated from the TS to the DBNPS Updated Safety

Analysis Report (USAR) Technical

[[Page 46437]]

Requirements Manual (TRM). These requirements would be relocated

generally intact to the TRM whereby future changes would be subject

to the regulatory controls of 10 CFR 50.59. These relocations are

consistent with the NRC guidance provided in Generic Letter (GL) 95-

10, ``Relocation of Selected Technical Specifications Requirements

Related to Instrumentation,'' or NUREG-1430, Revision 1, ``Standard

Technical Specifications--Babcock and Wilcox Plants,'' dated April

1995.

The proposed revision to TS 3/4.11.2, Radioactive Effluents--

Explosive Gas Mixture, and its Bases is an administration change to

a reference necessitated by the proposed relocation of TS 3/4.3.3.9

to the USAR TRM.

The proposed revision to TS 3/4.3.3.1 and TS 3/4.4.6.1 regarding

the number of Reactor Coolant System (RCS) leakage detection

monitors required and their allowed outage times is based upon the

NRC's guidance of NUREG-1430, Revision 1. This proposed revision

affects the TS only and does not reduce the number, diversity, or

sensitivity of Reactor Coolant System leakage detection systems

inside the containment building or as committed to in the DBNPS

USAR.

1b. Not involve a significant increase in the consequences of an

accident previously evaluated because no accident condition or

assumption is affected by the proposed revisions. As described

above, the revisions are consistent with the guidance of NRC GL 95-

10 or NUREG-1430, Revision 1. The proposed revisions, as described

above, do not alter the source term, containment isolation, or

allowable releases. The proposed changes, therefore, will not

increase the radiological consequences of a previously evaluated

accident.

2. Not create the possibility of a new or different kind of

accident from any accident previously evaluated because no new

accident initiators or assumptions are introduced by the proposed TS

revisions. No new accident scenarios, transient precursors, failure

mechanisms, or limiting failures are introduced as a result of the

proposed changes.

3. Not involve a significant reduction in a margin of safety

because the proposed revisions do not reduce or adversely affect the

capabilities of any plant structures, systems or components. The

proposed relocation of TS 3/4.3.3.2, TS 3/4.3.3.9, and TS 3/4.4.7 to

the USAR TRM is essentially an administrative change to the location

and process by which these requirements are controlled and revised.

Future revisions to these requirements relocated to the USAR TRM

will be subject to the regulatory controls of 10 CFR 50.59.

Therefore, these revisions will not result in a significant

reduction in a margin of safety.

The proposed revision to TS 3/4.11.2 and its Bases is

administrative and reflects the relocation of TS 3/4.3.3.9 to the

USAR TRM. Therefore, this revision will not result in a significant

reduction in a margin of safety.

The proposed revisions to TS 3/4.3.3.1 and TS 3/4.4.6.1 affect

the number of containment atmosphere radioactivity monitors required

by TS to be operable simultaneously. However, redundancy and

diversity requirements are maintained in the TS for detecting

Reactor Coolant System leakage. Although TS-allowed outage times are

proposed to be increased consistent with NUREG-1430, Revision 1

guidance, related compensatory action requirements are also being

increased. Furthermore, the DBNPS commitments made for complying

with Regulatory Guide 1.45, May, 1973, ``Reactor Coolant Pressure

Boundary Leakage Detection Systems,'' are not changed by the

proposed revisions. Along with the applicable revised TS

requirements, 10 CFR 50, Appendix B, Criterion XVI will require

prompt corrective action for inoperable leakage detection systems.

Accordingly, these proposed revisions will not result in a

significant reduction in a margin of safety.

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

this review, it appears that the three standards of 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 Section Chief: Anthony J. Mendiola.

FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse

Nuclear Power Station, Unit 1, Ottawa County, Ohio

Date of amendment request: July 26, 1999.

Description of amendment request: The proposed amendment would

change the Technical Specifications to adopt the performance-based 10

CFR Part 50, Appendix J, Option B approach for Type B and C containment

leakage rate testing, and to relocate certain details of the tests into

a Containment Leakage Testing Program.

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, Unit No. 1, in accordance with these changes would:

1a. Not involve a significant increase in the probability of an

accident previously evaluated because accident initiators,

conditions, or assumptions are not affected by the proposed changes.

The proposed changes to the Technical Specifications and Bases

implement 10 CFR [Part] 50 Appendix J Option B for Type B and C

Local Leak Rate Testing, based on the guidance of Regulatory Guide

1.163,

``Performance-Based Containment Leak-Test Program.'' Provided

that components have performed satisfactorily on a historical basis,

this guidance permits the use of extended testing frequencies. These

proposed changes do not affect accident initiators, conditions, or

assumptions.

1b. Not involve a significant increase in the consequences of an

accident previously evaluated because the proposed changes do not

change the source term or total allowable releases. With the

exception of the proposed increase in the containment air lock

leakage limits, the proposed changes do not affect the total

allowable containment leakage rates presently specified in the

Technical Specifications. Although the air lock leakage limits are

proposed to be increased, the accident analyses are based on the

current TS allowable maximum bypass leakage, which is not proposed

to be changed. Therefore, increases in leakage limits for individual

components, such as the air locks and their door seals, which are

constituents of bypass leakage, will have no effect on the

radiological consequences described in the accident analyses.

The proposed TS changes relating to implementation of 10 CFR

[Part] 50 Appendix J Option B may result in a small, but acceptable

increase in post-accident containment leakage, due to the increased

probability that due to generally increased intervals between tests,

an unacceptable leakage rate could go undetected for a longer length

of time. NUREG-1493, ``Performance-Based Containment Leak-Test

Program,'' September, 1995, which provided the technical basis for

the 10 CFR [Part] 50 Appendix J Option B rulemaking, provides a

detailed evaluation of the expected leakage and its consequences and

concludes that increased test frequencies are workable without

significant risk impacts.

2. Not create the possibility of a new or different kind of

accident from any accident previously evaluated because no new

accident initiators or assumptions are introduced by the proposed

changes. The proposed changes do not affect the methodology used in

conducting containment leak rate testing. The proposed changes do

not involve a change to the plant design or operation and,

therefore, will not introduce any new or different failure modes or

initiators.

3. Not involve a significant reduction in a margin of safety.

The proposed changes relating to implementation of 10 CFR [Part]

50, Appendix J, Option B do not significantly affect the allowable

containment leakage rates presently specified in the Technical

Specifications. The Technical Specifications, under the proposed

changes, will continue to ensure containment reliability by periodic

testing performed in full compliance with 10 CFR [Part] 50, Appendix

J.

[[Page 46438]]

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 Section Chief: Anthony J. Mendiola.

FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse

Nuclear Power Station, Unit 1, Ottawa County, Ohio

Date of amendment request: July 28, 1999.

Description of amendment request: The proposed amendment would

change Technical Specification (TS) Section 3/4.7.5.1, ``Ultimate Heat

Sink,'' to allow operation on Modes 1 through 4 with an Ultimate Heat

Sink water temperature of less than or equal to 90 deg.F, instead of

the current limit of less than or equal to 85 deg.F.

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, Unit No. 1, in accordance with these changes would:

1a. Not involve a significant increase in the probability of an

accident previously evaluated because no accident initiators,

conditions, or assumptions are significantly affected by the

proposed change. The proposed change would increase the allowable

Ultimate Heat Sink (UHS) water temperature, as specified in TS LCO

3.7.5.1.b, from less than or equal to 85 deg.F to less than or equal

to 90 deg.F. This water is used by the Service Water System to

provide cooling to equipment that is used to mitigate accidents such

as a Large Break Loss of Coolant Accident. This increase in Service

Water temperature has been evaluated and the proposed change does

not result in the operation of equipment important to safety outside

their acceptable operating ranges.

1b. Not involve a significant increase in the consequences of an

accident previously evaluated because the proposed change does not

change the source term, containment isolation, or allowable

releases. The proposed increase in the Service Water System

temperature has been evaluated with respect to the containment and

equipment used to mitigate the consequences of accidents previously

evaluated. These evaluations have determined that there are no

significant increases in consequences.

2. Not create the possibility of a new or different kind of

accident from any accident previously evaluated because no new

accident initiators or assumptions are introduced by the proposed

5 deg.F increase in UHS temperature. The proposed change does not

result in installed equipment being operated outside their design

operating ranges. No new or different equipment failure modes or

mechanisms are introduced by the proposed change.

3. Not involve a significant reduction in a margin of safety

because the proposed 5 deg.F increase in UHS temperature does not

result in significant changes to the initial conditions contributing

to accident severity or consequences.

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 Section Chief: Anthony J. Mendiola.

FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear

Power Plant, Unit 1, Lake County, Ohio

Date of amendment request: June 17, 1999.

Description of amendment request: The proposed amendment modifies

multiple surveillance requirements to support implementation of a 24-

month operating cycle.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration which is presented below:

The proposed amendment does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

A. Frequency Extensions

The proposed Technical Specification (TS) changes involve a

change in the surveillance testing intervals to facilitate a change

in the Perry Nuclear Power Plant (PNPP) operating cycle from 18

months to 24 months. The proposed TS changes do not physically

impact the plant, nor do they impact any design or functional

requirements of the associated systems. That is, the proposed TS

changes do not degrade the performance of, or increase the

challenges to, any safety systems assumed to function in the

accident analysis. The proposed TS changes do not impact the TS

surveillance requirements themselves, or the way in which the

surveillances are performed. In addition, the proposed TS changes do

not introduce any accident initiators, since no accidents previously

evaluated have, as their initiators, anything related to the

frequency of surveillance testing. Also, evaluation of the proposed

TS changes demonstrated that the availability of equipment and

systems required to prevent or mitigate the radiological

consequences of an accident are not significantly affected because

of other, more frequent testing that is performed, the availability

of redundant systems and equipment, or the high reliability of the

equipment. Since the impact on the systems is minimal, it is

concluded that the overall impact on the plant accident analysis is

negligible. Furthermore, a historical review of surveillance test

results and associated maintenance records indicated that there was

no evidence of any failures that would invalidate the above

conclusions. Therefore, the proposed TS changes do not significantly

increase the probability or consequences of an accident previously

evaluated.

B. Allowable Value Changes

The proposed changes in Allowable Values for the instrumentation

include in Table 3.3.8.1-1 Items d and e of the Technical

Specifications are the result of application of the Perry Instrument

Setpoint Methodology (ISM) using plant specific drift values.

Application of this methodology results in Allowable Values which

more accurately reflect total instrumentation loop accuracy as well

as that of test equipment and calculated drift between

surveillances. The proposed changes will not result in any hardware

changes. The instrumentation is not assumed to be an initiator of

any analyzed event. Existing operating margin between plant

conditions and actual plant setpoints is not significantly reduced

due to these changes. The role of the instrumentation is in

mitigating and thereby limiting the consequences of accidents. The

Allowable Values have been developed to ensure that the design and

safety analysis limits will be satisfied. The methodology used for

the development of the Allowable Values ensures the affected

instrumentation remains capable of mitigating design basis events as

described in the safety analyses and that the results and

radiological consequences described in the safety analyses remain

bounding. Additionally, the proposed change does not alter the

plant's ability to detect and mitigate events. Therefore, this

change does not involve a significant increase in the probability or

consequences of an accident previously evaluated.

[[Page 46439]]

C. Frequency Reductions to Semiannual

The proposed Technical Specification (TS) changes involve a

change in the surveillance testing intervals from 18 months to

either 6 months or quarterly. The shorter frequencies are based on

PNPP specific results of setpoint drift evaluations. The proposed

more restrictive TS changes do not physically impact the plant, nor

do they impact any design or functional requirements of the

associated systems. That is, the proposed TS changes do not degrade

the performance of, or increase the challenges to, any safety

systems assumed to function in the accident analysis. The proposed

TS changes do not impact the TS surveillance requirements

themselves, or the way in which the surveillances are performed. In

addition, the proposed TS changes do not introduce any accident

initiators, since no accidents previously evaluated have, as their

initiators, anything related to the frequency of surveillance

testing. The proposed TS frequencies will demonstrate that the

equipment and systems required to prevent or mitigate the

radiological consequences of an accident are continuing to meet the

assumptions of the setpoint evaluation, on a more frequent basis.

Since the impact on the systems is minimal, and the assumptions of

the safety analyses will be maintained, it is concluded that the

overall impact on the plant accident analysis is negligible.

Furthermore, a historical review of surveillance test results and

associated maintenance records indicated that there was no evidence

of any failures that would invalidate the proposed test frequencies.

Therefore, the proposed TS changes do not significantly increase the

probability or consequences of an accident previously evaluated.

The proposed amendment would not create the possibility of a new

or different kind of accident from any accident previously

evaluated.

A. Frequency Extensions

The proposed TS changes involve a change in the surveillance

testing intervals to facilitate a change in the PNPP operating cycle

length. The proposed TS changes do not introduce any failure

mechanisms of a different type than those previously evaluated,

since there are no physical changes being made to the facility. No

new or different equipment is being installed. No installed

equipment is being operated in a different manner. As a result, no

new failure modes are being introduced. In addition, the

surveillance test requirements themselves, and the way surveillance

tests are performed, will remain unchanged. Furthermore, a

historical review of surveillance test results and associated

maintenance records indicated there was no evidence of any failures

that would invalidate the above conclusions. Therefore, the proposed

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

of accident from any previously evaluated.

B. Allowable Value Changes

The proposed changes are the result of application of the ISM

using plant specific drift values and do not create the possibility

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

evaluated. This is based on the fact that the method and manner of

plant operation is unchanged. The use of the proposed Allowable

Values does not impact safe operation of PNPP in that the safety

analysis limits will be maintained. The propose Allowable Values

involve no system additions or physical modifications to systems in

the station. These Allowable Values were revised to ensure the

affected instrumentation remains capable of mitigating accidents and

transients. Plant equipment will not be operated in a manner

different from previous operation, except that setpoints may be

changed. Since operational methods remain unchanged and the

operating parameters have been evaluated to maintain the station

within existing design basis criteria, no different type of failure

or accident is created.

C. Frequency Reductions to Semiannual or Quarterly

The proposed TS changes involve a change in the surveillance

testing interval due to the application of the ISM and plant

specific drift analysis results. Also, the quarterly tests reflect

current PNPP calibration practices, since the components are

normally calibrated during the Channel Functional Test. The proposed

TS changes do not introduce any failure mechanisms of a different

type than those previously evaluated, since there are no physical

changes being made to the facility. No new or different equipment is

being installed. No installed equipment is being operated in a

different manner. The proposed change does not impact core

reactivity, or the manipulation of fuel bundles. As a result, no new

failure modes are being introduced. In addition, the surveillance

test requirements themselves, and the way surveillance tests are

performed, will remain unchanged. Furthermore, a historical review

of surveillance test results and associated maintenance records

indicated there was no evidence of any failures that would

invalidate the above conclusions. Therefore, the proposed TS changes

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

from any previously evaluated.

The proposed amendment will not involve a significant reduction

in a margin of safety.

A. Frequency Extensions

Although the proposed TS changes will result in changes in the

interval between surveillance tests, the impact, if any, on system

availability is small, based on other, more frequent testing that is

performed, or the existence of redundant systems and equipment, or

overall system reliability. Evaluations have shown there is no

evidence of time dependent failures that would impact the

availability of the systems. The proposed change does not

significantly impact the condition or performance of structures,

systems, and components relied upon for accident mitigation. The

proposed change does not significantly impact any safety analysis

assumptions or results. Therefore, the proposed change does not

involve a significant reduction in a margin of safety.

B. Allowable Value Changes

The proposed change does not involve a reduction in a margin of

safety. The proposed changes have been developed using a methodology

to ensure safety analysis limits are not exceeded. As such, this

proposed change does not involve a significant reduction in a margin

of safety.

C. Frequency Reductions to Semiannual or Quarterly

The proposed TS changes will result in a shorter interval

between surveillance tests to ensure that the assumptions of the

safety analysis are maintained. The impact, if any, on system

availability is small, as a result of this more frequent testing

that is performed. The proposed change does not significantly impact

the condition or performance of structures, systems, and components

relied upon for accident mitigation. The proposed change does not

significantly impact any safety analysis assumptions or results.

Therefore, the proposed change does not involve a significant

reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Perry Public Library, 3753

Main Street, Perry, OH 44081.

Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts &

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

NRC Section Chief: Anthony J. Mendiola.

FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear

Power Plant, Unit 1, Lake County, Ohio

Date of amendment request: August 4, 1999.

Description of amendment request: The amendment would incorporate

an additional option into the Required Actions for Technical

Specification 3.9.1, ``Refueling Equipment Interlocks.'' The change

would provide additional Required Actions when the refueling interlocks

are inoperable. The alternative would permit continued refueling

activities once control rod withdrawal is blocked and operators verify

that all appropriate controls rods are fully inserted.

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

[[Page 46440]]

consequences of an accident previously evaluated.

The refueling interlocks are explicitly assumed in the Perry

Nuclear Power Plant Updated Safety Analysis Report (USAR) analyses

of the control rod removal error and fuel loading error during

refueling. This analysis evaluates the probability and consequences

of control rod withdrawal during refueling. Criticality and,

therefore, subsequent prompt reactivity excursions are prevented

during the loading of fuel, provided all required control rods are

fully inserted. The refueling interlocks accomplish this by

preventing loading fuel into the core with any control rod

withdrawn, or by preventing withdrawal of a rod from the core during

fuel loading. When the refueling interlocks are inoperable, the

current method of preventing fuel loading when a control rod is

withdrawn, is to prevent fuel movement. This method is currently

required by the Technical Specifications. An alternate method to

ensure that fuel is not loaded into a cell with the control rod

withdrawn is to prevent control rods from being withdrawn and verify

that all control rods required to be inserted are fully inserted.

The proposed Technical Specification Required Actions will require

that a control rod block be placed in effect, thereby ensuring that

control rods are not subsequently inappropriately withdrawn.

Additionally, following placing the control rod withdrawal block in

effect, the proposed actions will require that all required control

rods be verified to be fully inserted. This verification is in

addition to the requirements to periodically verify control rod

position by other Technical Specification requirements. These

proposed actions will ensure that control rods are not withdrawn and

cannot be inappropriately withdrawn, because an electrical or

hydraulic block to control rod withdrawal is in place. Like the

current requirements, the proposed will ensure that unacceptable

operations are blocked (e.g., loading fuel into a cell with a

control rod withdrawn, except when following the requirements of LCO

3.10.6, ``Multiple Control Rod Removal--Refueling,'' which is

unaffected by this change). The proposed additional Required Actions

provide an equivalent level of assurance that fuel will not be

loaded into a core cell with a control rod withdrawn as do the

current Required Action or the Surveillance Requirement. Therefore,

the proposed change does not significantly increase the probability

or consequences of an 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 change in the Technical Specification requirements does not

involve a change in the plant design, or to the status of the

reactor core during refueling. The proposed actions will ensure that

control rods are not withdrawn and cannot be inappropriately

withdrawn, because an electrical or hydraulic block to control rod

withdrawal is in place. Although the exact method by which the

control rod withdrawal block is inserted is revised, the net effect

is equivalent. The requirements will continue to ensure that fuel is

not loaded into the core when a control rod is withdrawn, except

when following the requirements of LCO 3.10.6, ``Multiple Control

Rod Removal--Refueling,'' which is unaffected by this change.

Therefore, no new failure modes are introduced, and 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 will not involve a significant reduction

in the margin of safety.

As discussed in the Bases for the affected Technical

Specification requirements, inadvertent criticality is prevented

during the loading of fuel provided all required control rods are

fully inserted during the fuel insertion. The refueling interlocks

function to support the refueling procedures by preventing control

rod withdrawal during fuel movement and the inadvertent loading of

fuel when a control rod is withdrawn. The proposed change will allow

the refueling interlocks to be inoperable and fuel movement to

continue only if a control rod withdrawal block is in effect and all

required control rods are verified to be fully inserted. These

proposed Required Actions provide an equivalent level of protection

as the refueling interlocks by preventing a configuration which

could lead to an inadvertent criticality event. The refueling

procedures will continue to be supported by the proposed Required

Actions because control rods cannot be withdrawn and as a result

fuel cannot be inadvertently loaded when a control rod is withdrawn,

except when following the requirements of LCO 3.10.6, ``Multiple

Control Rod Removal--Refueling,'' which is unaffected by this

change. Therefore, the proposed changes do not cause 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: Mary E. O'Reilly, Attorney, FirstEnergy

Corporation, 76 South Main Street, Akron, OH 44308.

NRC Section Chief: Anthony J. Mendiola.

Florida Power and Light Company, et al., Docket No. 50-389, St. Lucie

Plant, Unit No. 2, St. Lucie County, Florida

Date of amendment request: February 23, 1999.

Description of amendment request: The proposed license amendment

would remove redundant boron concentration monitoring requirements

specified for operating Modes 3 through 6 by deleting Technical

Specification 3/4.1.2.9, ``Reactivity Control Systems--Boron

Dilution.'' These requirements were interim measures intended to apply

until a permanent boron dilution alarm system was installed and

functional.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

amendment would not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The proposed amendment does not involve changes to previously

evaluated accident initiators. The proposed deletion of the

redundant boron concentration verification requirements do not

impact the results of existing accident analyses, and will have no

adverse impact on any plant system performance. TS 3/4.1.2.9

provides mode and charging pump dependent monitoring requirements

for RCS boron concentration that are designed to detect an unplanned

boron dilution event in MODES 3 through 6 in the absence of an

automatic alarm system, and is based on the time requirements for

operator action specified in Section 15.4.6 of the Standard Review

Plan (SRP). This specification evolved from interim measures that

were proposed by FPL until the boron dilution alarm system (BDAS)

could be made completely functional following initial start-up of

St. Lucie Unit 2. The BDAS is completely functional and provides

redundant control room alarms to alert operators to the occurrence

of an unplanned boron dilution event in Modes 3 through 6. The alarm

setpoints are based on Chemical and Volume Control System (CVCS)

malfunction analyses, and satisfy the same SRP acceptance criteria

upon which the monitoring requirements of TS 3/4.1.2.9 were based.

Therefore, operation of the facility in accordance with the proposed

amendment will not involve a significant increase in the probability

or consequences of an accident previously evaluated.

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

amendment would not create the possibility of a new or different

kind of accident from any accident previously evaluated.

The proposed amendment will not change the physical plant or the

modes of operation defined in the facility license. The amendment

will remove requirements from the facility technical specifications

that were proposed by FPL as interim measures until the boron

dilution alarm system became completely functional. The amendment

will not alter the design of St. Lucie plant systems described in

the Updated Final Safety Analysis Report (UFSAR), and the plant

configuration will continue to remain consistent with assumptions

used in the existing accident analyses. Therefore, operation of the

facility in accordance with the proposed amendment would not create

the possibility of a new or different kind of

[[Page 46441]]

accident from any accident previously evaluated.

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

amendment would not involve a significant reduction in a margin of

safety. The proposed amendment has been evaluated with respect to

the applicable safety analyses. The BDAS provides a continuous,

early warning capability to detect a boron dilution event in Modes

3, 4, 5 and 6, and satisfies the same SRP time requirements for

operator action as the interim TS that is proposed for deletion.

BDAS setpoints are determined and/or validated for each fuel cycle

to ensure they remain consistent with the CVCS malfunction analyses

of record, and changes that may become necessary are controlled

pursuant to 10 CFR 50.59. The minimum required Shutdown Margin is

not changed by this proposal. Therefore, operation of the facility

in accordance with the proposed amendment would not involve a

significant reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Indian River Junior College

Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003.

Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light,

P.O. Box 14000, Juno Beach, Florida 33408-0420.

NRC Section Chief: Sheri R. Peterson.

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

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

Date of amendment request: July 27, 1999.

Description of amendment request: The proposed amendments request

that Turkey Point Unit 3 Technical Specification (TS) 3/4.8.1, A.C.

SOURCES,TS 3/4.4.3, PRESSURIZER, and TS 3/4.5.2, ECCS SUBSYSTEMS--

Tavg GREATER THAN OR EQUAL TO 350 deg.F, be revised on a

one-time basis to extend the Allowed Outage Time (AOT) for an

inoperable Emergency Diesel Generator (EDG) from 72 hours to 7 days.

The proposed one-time AOT extension will be used to replace the Unit 3

EDG engine radiators prior to the Spring 2000 refueling outage.

However, replacement of the radiator is a very labor-intensive

evolution that cannot be performed within the existing 72 hour AOT. The

proposed AOT extension will allow the radiator replacement activity to

be completed successfully in a safe manner. The extended AOT will be

applied to one EDG at a time in a sequential manner. When the radiator

replacement activity is complete on one engine, it will be returned to

service so that work can proceed on the redundant EDG. It should be

noted that although the proposed changes apply only to Unit 3, the Unit

4 TSs are administratively affected since the TSs are combined for both

units.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

amendment would not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The Emergency Diesel Generators (EDG) are part of the on-site

electrical power distribution system. They function as a standby

power source in the event that the preferred A.C. power supply,

i.e., offsite power, is interrupted. While certain failures in the

electrical distribution system can lead to a loss of offsite power

which is a design basis event for the plant, the EDGs are not

assumed to be an initiating condition of any accident evaluated in

the safety analysis report. Therefore, a one-time extension in the

EDG Allowed Outage Time (AOT) does not involve a significant

increase in the probability of an accident previously evaluated.

The purpose of the proposed license amendment is to permit on-

line replacement of the Unit 3 EDG radiators. The radiators are part

of the closed-loop diesel engine cooling water system and do not

interface with any system or component that contains radioactivity.

The EDGs do supply A.C. power to the emergency core cooling and

containment heat removal systems during accidents that involve loss

of offsite power. However, no changes are predicted for the

postulated post-accident releases since adequate EDG capacity will

be available under the conditions of the proposed license amendment

to accommodate any design basis accident condition. Accordingly, the

consequences of accidents previously evaluated in the safety

analysis report are not changed by an extended EDG outage.

Probabilistic Safety Assessment (PSA) techniques were used to

evaluate the impact of a one-time extension of the EDG AOT from 72

hours to 7 days. The results of these analyses indicate that

extending the AOT for the purpose of replacing the engine radiator

cores represents an acceptably small impact on Core Damage

Probability.

Based on the above, FPL concludes that the proposed amendment

does not involve a significant increase in the probability or

consequences of any accident previously evaluated.

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

amendment would not create the possibility of a new or different

kind of accident from any previously evaluated.

The proposed change does not alter the design, physical

configuration, or modes of operation of the plant. Plant

configurations that are prohibited by Technical Specifications will

not be created by the one-time EDG AOT extension. Therefore, the

proposed activity does not create the possibility of a new or

different kind of accident from any previously evaluated.

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

amendment would not involve a significant reduction in a margin of

safety.

The proposed license amendment will extend by 96 hours the

requirement to shutdown the plant when a Unit 3 EDG is removed from

service for maintenance. The one-time AOT extension will not alter

plant equipment, setpoints, or operating practices that provide the

existing margins of safety. Therefore, the change does not involve a

significant reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Florida International

University, University Park, Miami, Florida 33199.

Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light,

P.O. Box 14000, Juno Beach, Florida 33408-0420.

NRC Section Chief: Sheri R. Peterson.

Portland General Electric Company, et al., Docket No. 50-344, Trojan

Nuclear Plant, Columbia County, Oregon

Date of amendment request: August 27, 1998.

Description of amendment request: The amendment would delete the

requirements for an emergency plan from the 10 CFR Part 50 license and

technical specifications after the spent nuclear fuel is transferred to

a Part 72 licensed independent spent fuel storage installation (ISFSI).

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 elimination of the emergency plan requirements from

the 10 CFR 50 license is predicated on completion of transfer of the

spent nuclear fuel to the proposed 10 CFR 72 ISFSI licensed area and

removal of the reactor vessel and internals from the 10 CFR 50

licensed area of the site.

[[Page 46442]]

Removal of the potential radiological source terms for accidents

previously evaluated effectively eliminates the credibility of the

accidents, therefore, elimination of the emergency plan requirements

does not involve a significant increase in the probability or

consequences of an accident previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

The proposed change is deletion of emergency plan requirements

and, as such, has no direct impact on plant equipment or the

procedures for operating plant equipment. Therefore, it 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.

Following the removal of the spent nuclear fuel and the reactor

vessel and internals from the 10 CFR 50 licensed area, the remaining

credible accidents are limited to decommissioning activities. The

potential accidents associated with decommissioning activities are

presented in the TNP [Trojan Nuclear Plant] Decommissioning Plan and

have been shown to have consequences less than the EPA PAGs

[Environmental Protection Agency Protective Action Guidelines].

Following the removal of the spent nuclear fuel and the reactor

vessel (including the internals) from the 10 CFR 50 site, no

credible accidents associated with the remaining decommissioning

activities would require pre-planned emergency measures to avoid

acute radiation doses. The deletion of the Trojan Nuclear Plant

Permanently Defueled Emergency Plan will not result in a reduction

in the margin of safety previously analyzed. Therefore, the proposed

10 CFR 50 license amendment does not involve a significant reduction

in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Branford Price Millar Library,

Portland State University, 934 S.W. Harrison Street, P.O. Box 1151,

Portland, Oregon 97207.

Attorney for licensee: Leonard A. Girard, Esq., Portland General

Electric Company, 121 S.W. Salmon Street, Portland, Oregon 97204.

NRC Section Chief: Michael T. Masnik.

Power Authority of The State of New York, Docket No. 50-286, Indian

Point Nuclear Generating Unit No. 3, Westchester County, New York

Date of amendment request: February 19, 1998, as supplemented July

28, 1999.

Description of amendment request: This application for amendment to

the Indian Point 3 Technical Specifications (TSs) proposes to revise

the Radioactive Effluents Technical Specifications (RETS) in accordance

with Generic Letter 89-01 (GL-89-01), to make changes to implement

revised 10 CFR Part 20 requirements, and to make administrative changes

under 10 CFR 50.36a.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Does the proposed license amendment involve a significant

increase in the probability or consequences of any accident

previously evaluated?

A. The proposed changes involve (1) combining related LCO and

surveillance requirements from Sections 2.0 and 3.0, respectively,

of the Indian Point 3 (IP3) RETS and relocating this text to the new

Radiological Effluent Controls (REC) section of the ODCM, (2)

relocating the bases contained in Section 4.0 of the RETS to the

ODCM REC, (3) relocating the detailed reporting requirements

contained in Section 5.0 of the RETS to the ODCM REC, and (4)

updating references to 10 CFR Part 20. Additional changes include

formatting both the remaining RETS and the new REG to more closely

model Standard Technical Specifications (STS), revising the

frequency of the Radioactive Effluent Release Report in accordance

with 10 CFR 50.36a, relocating all definitions to Appendix A of the

Technical Specifications and adding/deleting definitions as

necessary, and adding a new Special Reports section to the ODCM.

Most of the changes are (1) consistent with the guidance provided in

the generic letter, NUREG-1301, or provisions of 10 CFR; or (2)

editorial. Editorial changes include the relocation of text,

correction of typographical and punctuation errors, renumbering,

reformatting, immaterial wording revisions/deletions/clarifications

which do not change intent, and updating references.

B. The proposed revisions to the liquid and gaseous release rate

limits, the relocation of the old 10 CFR 20.106 requirements to the

new 10 CFR 20.1302, and the revision to the TS bases for the Liquid

Holdup Tank activity will involve no change in the types or amounts

of effluents that will be released, nor will there be an increase in

individual or cumulative occupational radiation exposures.

The changes of definitions, terminology, paragraph references,

and report submittal frequency are necessary to keep IP3 TS

consistent with revised federal regulations (i.e., 10 CFR 20 and 10

CFR 50.36(a)). Record retention and reporting requirements will

continue to meet NRC regulations. These changes are administrative

in nature and do not affect plant hardware or operation.

The changes do not impact the operation, design, configuration,

or testing of plant structures, systems or components. As such, the

proposed changes do not involve a significant increase in the

probability or consequences of any accident previously evaluated.

2. Does the proposed license amendment create the possibility of

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

A. The changes do not impact the operation, design,

configuration, or testing of plant structures, systems or

components. The changes do not result in a change in type or amount

of radiological effluents released. As such, the proposed changes do

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

from any previously evaluated.

3. Does the proposed amendment involve a significant reduction

in a margin of safety?

A. The changes are being made in accordance with NRC guidance

and continue to assure compliance with the applicable regulatory

requirements including 10 CFR 20. The changes do not result in a

change in the types or amounts of effluents released. The current

level of radiological effluent control will be maintained. As such,

the proposed changes do not involve a significant reduction in a

margin of safety.

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

this review, it appears that the three standards of 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: White Plains Public Library,

100 Martine Avenue, White Plains, New York 10601.

Attorney for licensee: Mr. David E. Blabey, 10 Columbus Circle, New

York, New York 10019.

NRC Section Chief: S. Singh Bajwa.

Sacramento Municipal Utility District (the District), Docket No. 50-

312, Rancho Seco Nuclear Station, Sacramento County, California

Date of amendment request: March 18, 1996 (PA-192).

Description of amendment request: The proposed amendment would

update the Rancho Seco cask drop analysis and establish the cask drop

event as the design-basis event for plant operation in the permanently

defueled mode. The proposed amendment would also make editorial changes

to the Permanently Defueled Technical Specifications and Bases by

adding the word ``heavy'' to specification D3.3 and eliminating

references to the MP-187 cask in specification D3.3 and D4.3.3.

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:

[[Page 46443]]

The District has reviewed the proposed changes against each of

the criteria in 10 CFR 50.92, and, based on the above safety

analysis, concludes:

Using the Gantry Crane to handle a fully loaded transfer cask in

the Fuel Storage Building will not create a significant increase in

the probability or consequences of an accident previously evaluated

in the SAR [safety analysis report], because the conservative dose

consequence calculated for the updated, design basis cask drop event

resulted in an exposure (224 mrem) that is:

1. A very small percentage ([approximately] 0.9%) of the 10 CFR

100 design basis accident dose limit of 25 rem total body;

2. A small percentage ([approximately] 3.6%) of the NUREG-0612

control of heavy loads accident dose limit of 6.25 rem total body;

3. Well within ([approximately] 4.5%) of the old EPA

[Environmental Protection Agency] NUREG-0654 plume exposure

Protective Action Guidelines of 500 mrem total body dose;

4. Well within the new EPA 1 to 5 rem Total Effective Dose

Equivalent (TEDE) Protective Action Guidelines (PAGs) specified in

document EPA-400-R-92-001, Table 2-1, May 1992;

5. Less than the maximum hypothetical Rancho Seco Independent

Spent Fuel Storage Installation design basis accident (375 mrem

total body dose);

6. Less than the original Rancho Seco operating design basis for

the Fuel Storage Building FHA [fuel-handling accident] exposure (399

mrem);

7. Less than the original Rancho Seco operating design basis for

the Reactor Building FHA exposure (477 mrem); and

8. Much less than the original Rancho Seco operating design

basis Maximum Hypothetical Accident exposure (3,600 mrem).

Therefore, the conservatively calculated 224 mrem cask drop

design basis accident exposure is (1) relatively small and (2) not

considered a significant hazard.

Also, the probability of occurrence of the FHA, which is the

current design basis accident, is similar to the probability of

occurrence of the updated cask drop event. The FHA is assumed to

occur because the fuel handling bridge is not single failure proof.

Likewise for the cask drop scenario, since the Gantry Crane is not

single failure proof, this Safety Analysis Report evaluates the

Gantry Crane dropping a loaded spent fuel cask.

This Safety Analysis Report analyzes the dropped cask accident

scenario even though the Gantry Crane and fuel handling bridge are:

1. Designed to safely handle their respective loads (i.e., a

loaded transfer cask and a spent fuel assembly, respectively; and

2. In compliance with the design and administrative requirements

addressed in NUREG-0612, ``Control of Heavy Loads at Nuclear Power

Plants.''

A loaded cask transfer drop is a very unlikely event because of

the numerous Gantry Crane safety features described in the above

safety Analysis Report. These features described above include:

1. Gantry Crane Administrative Safety Features;

2. Gantry Crane Design Safety Features;

3. General Gantry Crane Control System Design Safety Features;

4. Gantry Crane Radio Control System Design Safety Features;

5. Hoist Design Safety Features; and

6. Trolly and Bridge Design Safety Features.

The updated cask drop accident scenario will not create the

possibility of a new or different type of accident than previously

evaluated in the SAR, because the DSAR [defueled SAR] currently

evaluates a cask drop event. The cask drop scenario evaluated in the

above Safety Analysis Report just updates the existing cask drop

analysis. The updated cask drop analysis only:

1. Identifies the type of spent fuel cask that Rancho Seco will

use;

2. Results in a change to the calculated dose consequence

associated with the current, bounding, design basis accident (i.e.,

the FHA); and

3. Results in a change to the existing Rancho Seco cask drop

analysis.

The updated, design basis, cask drop event will not involve a

significant reduction in the margin of safety, because the

conservatively calculated dose consequence associated with the

postulated drop of a spent fuel transfer cask is:

1. Relatively small (i.e., 224 mrem) compared to the eight

accident limits and previously calculated accident doses summarized

above;

2. A very unlikely event;

3. Not a significant hazard; and

4. Not a public health and safety concern.

This conclusion is the same for the FHA, which is the current,

bounding, Rancho Seco design basis accident.

Also, the Emergency Planning Zone remains unchanged for this

updated, cask drop accident scenario. No significant changes to the

Rancho Seco Emergency Plan result from this proposed change to the

updated, design basis accident at Rancho Seco.

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. The staff also reviewed the proposed editorial changes for

no significant hazards consideration. The proposed editorial changes do

not affect the design or operation of the facility and also satisfy the

three standards of 10 CFR 50.92(c). Therefore, the NRC staff proposes

to determine that the requested amendment involves no significant

hazards consideration.

Local Public Document Room location: Central Library, Government

Documents, 828 I Street, Sacramento, California 95814

Attorney for licensee: Dana Appling, Esq., Sacramento Municipal

Utility District, P.O. Box 15830, Sacramento, California 95852-1830

NRC Section Chief: Michael T. Masnik

Southern Nuclear Operating Company, Inc, Docket Nos. 50-348 and 50-364,

Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County, Alabama

Date of amendment request: March 12, 1998, as supplemented April

24, August 20 and November 20, 1998, and February 3, 1999

Description of amendment request: The proposed amendments would

revise the Technical Specifications (TS) of each unit to conform with

NUREG-1431, Revision 1, ``Standard Technical Specifications--

Westinghouse Plants.'' The Commission had previously issued a Notice of

Consideration of Issuance of Amendments in the Federal Register on May

25, 1999, (64 FR 28218) covering all the proposed changes that were

within the scope of NUREG-1431. The following descriptions and no

significant hazard analyses cover only those items that are beyond the

scope of NUREG-1431. Associated with each change are administrative/

editorial changes which would make the new or revised requirements fit

into the format of NUREG-1431.

1. The Standard Technical Specification (STS) terms FQW(Z) and

FQC(Z) in Limiting Condition for Operation (LCO) 3.2.1 would be deleted

and the terms FQ(Z), ``steady state'' limit and ``transient'' limit

would be used. (Significant Hazards Evaluation A)

2. The STS wording in Required Action 3.2.4.A to ``reduce'' thermal

power would be revised to ``limit'' thermal power to allow entry into

the LCO applicability during startup when QPTR may be in excess of 1.02

due to transient core conditions which are usually self-correcting. (A)

3. The Applicability of LCO 3.2.4 would be revised to be consistent

with the Applicability for the AFD LCO to eliminate subtle differences

between the two LCO Applications which were previously the same. (M)

4. The Reactor Coolant System Loop Test specified in the TS LCO 3/

4.10.4 would not be included. (L-1)

5. A new Action would be added to the Emergency Core Cooling System

(ECCS)--Shutdown LCO 3.5.3. The new Action deals with the centrifugal

charging subsystem. (L-2)

6. The Reactor Coolant Pump (RCP) seal injection flow requirements

of 3.5.5 would be revised. The requirement to verify a single operating

point would be changed to require verification of a range of values on

an operating curve. (M)

7. The time allowed to reduce the power range neutron flux setpoint

in 3.7.1 to within the required limit would be extended and made

applicable in Mode 1 only. (L-3 and L-3a)

[[Page 46444]]

8. The Actions in 3.7.2 for an inoperable Main Steam Isolation

Valve (MSIV) would be revised to take credit for the redundant MSIVs in

each steam line. (L-4)

9. An Action would be added to the Service Water (SW) LCO 3.7.8

that accounts for the redundant automatic turbine building isolation

valves in each Farley SW train. (L-5)

10. The diesel generator accelerated Test Table 3.8.1-1 would be

deleted. (LA)

11. The AC Sources--Shutdown surveillance 3.8.2.1 would be revised

to more clearly state the required surveillances. (L-6 and L-6a)

12. The Actions 3.8.4 and 3.8.9 for an inoperable SW intake

structure Battery and Distribution System would be revised to more

accurately reflect the Farley design. (L-7)

13. The STS footnote to ESFAS Table 3.3.2-1 would be revised to be

consistent with the design of the Farley main steam system. (L-8)

14. A new Condition C would be added to LCO 3.3.4 to address

actions associated with the source range neutron flux monitor.

This text is long and has been trimmed here. Open the source document for the complete record.