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

Federal RegisterDec 3, 1997

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NUCLEAR REGULATORY COMMISSION

Biweekly Notice; Applications and Amendments to Facility

Operating Licenses Involving No Significant Hazards Considerations

I. Background

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

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

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

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

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

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

the Commission the authority to issue and 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 November 7, 1997, through November 20, 1997.

The last biweekly notice was published on November 19, 1997 (62 FR

61836).

Notice of Consideration of Issuance of Amendments to Facility Operating

Licenses, Proposed No Significant Hazards Consideration Determination,

and Opportunity for a Hearing

The Commission has made a proposed determination that the following

amendment requests involve no significant hazards consideration. Under

the Commission's regulations in 10 CFR 50.92, this means that operation

of the facility in accordance with the proposed amendment would not (1)

involve a significant increase in the probability or consequences of an

accident previously evaluated; or (2) create the possibility of a new

or different kind of accident from any accident previously evaluated;

or (3) involve a significant reduction in a margin of safety. The basis

for this proposed determination for each amendment request is shown

below.

The Commission is seeking public comments on this proposed

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

publication of this notice will be considered in making any final

determination.

Normally, the Commission will not issue the amendment until the

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

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

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

the Commission may issue the license amendment before the expiration of

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

the amendment involves no significant hazards consideration. The final

determination will consider all public and State comments received

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

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

opportunity for a hearing after issuance. The Commission expects that

the need to take this action will occur very infrequently.

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

Directives Branch, Division of Freedom of Information and Publications

Services, Office of Administration, U.S. Nuclear Regulatory Commission,

Washington, DC 20555-0001, and should cite the publication date and

page number of this Federal Register notice. Written comments may also

be delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike,

Rockville, Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays.

Copies of written comments received may be examined at the NRC Public

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

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

intervene is discussed below.

By January 2, 1998, the licensee may file a request for a hearing

with respect to issuance of the amendment to the subject facility

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

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

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

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

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

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

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

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

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

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

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

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

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

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

Secretary or the designated Atomic Safety and Licensing Board will

issue a notice of a hearing or an appropriate order.

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

shall set forth with particularity the interest of the petitioner in

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

the proceeding. The petition should specifically explain the reasons

why intervention should be permitted with particular reference to the

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

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

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

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

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

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

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

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

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

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

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

the specificity requirements described above.

Not later than 15 days prior to the first prehearing conference

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

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

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

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

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

brief explanation of the bases of the contention and a concise

statement of the alleged facts or expert opinion which support the

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

contention at the hearing. The petitioner must also provide references

to those specific sources and documents of which the petitioner is

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

facts or expert opinion. Petitioner must provide sufficient information

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

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

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

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

petitioner who fails to file such a supplement which satisfies these

requirements with respect to at least one contention will not be

permitted to participate as a party.

Those permitted to intervene become parties to the proceeding,

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

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

hearing, including the opportunity to present evidence and cross-

examine witnesses.

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

determination on the issue of no significant hazards consideration. The

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

If the final determination is that the amendment request involves

no significant hazards consideration, the Commission may issue the

amendment and make it immediately effective, notwithstanding the

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

of the amendment.

If the final determination is that the amendment request involves a

significant hazards consideration, any hearing held would take place

before the issuance of any amendment.

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

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

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

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

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

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

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

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

Nontimely filings of petitions for leave to intervene, amended

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

not be entertained absent a determination by the Commission, the

presiding officer or the Atomic Safety and Licensing Board that the

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

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

For further details with respect to this action, see the

application for amendment which is available for public inspection at

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

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

the particular facility involved.

Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-324,

Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North

Carolina

Date of amendments request: November 6, 1997.

Description of amendments request: The proposed amendments change

the Technical Specifications (TS) for the Brunswick Steam Electric

Plant (BSEP) Units 1 and 2 to allow three 18-month diesel generator

(DG) surveillance requirements (SR) to be performed during both plant

operation (Operational Conditions 1 and 2) and shutdown (Operational

Conditions 3, 4, and 5) rather than, as currently required, only during

shutdown. The first SR is an inspection of the DG involving a partial

disassembly. The second ensures that non-critical DG protective

functions are bypassed on an Emergency Core Cooling system actuation

signal. The third verifies that the DG operates for greater than or

equal to 60 minutes while loaded to at least 3500 kw, which bounds the

maximum expected post-accident diesel generator loading. The proposed

amendments additionally remove an expired footnote from the BSEP Unit 2

DG 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:

10 CFR 50.92 provides standards for determining whether a

significant hazards consideration exists. A proposed amendment to an

operating license for a facility involves no significant hazards

consideration if 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, (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. Carolina Power & Light Company has reviewed

these proposed license amendment requests and has concluded that their

adoption would not involve a significant hazards consideration. The

basis for this determination follows.

1. The proposed license amendments do not involve a significant

increase in the probability or consequences of an accident previously

evaluated.

The proposed license amendments add a footnote to SR 4.8.1.1.2.d to

allow performance of SR 4.8.1.1.2.d.1, SR 4.8.1.1.2.d.4, and SR

4.8.1.1.2.d.5 in OPERATIONAL CONDITION 1, 2, 3, 4, or 5 rather than

only during shutdown. The footnote requires the unit to be in

OPERATIONAL CONDITION 3, 4, or 5 when performing SR 4.8.1.1.2.d.2, SR

4.8.1.1.2.d.3, SR 4.8.1.1.2.d.6, and SR 4.8.1.1.2.d.7 for its

associated diesel generators. No such limitation is placed on SR

4.8.1.1.2.d.1, SR 4.8.1.1.2.d.4, or SR 4.8.1.1.2.d.5.

There is no relaxation of any limiting condition for operation

(LCO) and no decrease in surveillance requirements as a result of the

proposed amendments. As such, the proposed license amendments will not

affect the ability of the diesel generators to perform their intended

safety function. Performance of SR 4.8.1.1.2.d.1, SR 4.8.1.1.2.d.4, and

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SR 4.8.1.1.2.d.5, during power operations, will not adversely affect

overall nuclear safety. Diesel generator capacity is such that any

three of the four diesel generators can supply the required loads for

the safe shutdown of one unit and a design basis accident on the other

unit without relying on offsite power. The diesel generator is not tied

to the emergency bus (E bus) during performance of SR 4.8.1.1.2.d.1 or

SR 4.8.1.1.2.d.4. Therefore, performance of SR 4.8.1.1.2.d.1 and SR

4.8.1.1.2.d.4, during power operation, will not affect the operability

of any other safety-related systems nor will it create any

perturbations of the electrical distribution system that could

challenge plant operation.

Performance of SR 4.8.1.1.2.d.5, during power operation, will not

adversely affect overall nuclear safety. SR 4.8.1.1.2.d.5 is performed

in a similar manner to SR 4.8.1.1.2.a.5, which requires that, at least

once per 31 days on a staggered test basis, a diesel generator be

synchronized to the E bus and loaded to 1750 kw for 15 minutes. The

critical portions of these surveillances are when the diesel generators

are being synchronized to the E bus or disconnected from the E Bus. As

such, performance of SR 4.8.1.1.2.d.5 during power operation does not

create an additional opportunity of a perturbation of the electrical

distribution system that could challenge plant operation than currently

exists as a result of the performance of SR 4.8.1.1.2.a.5. The existing

design of the electrical distribution system ensures that a grid

problem will not result in failure of a diesel generator when it is

synchronized to the E bus. The E buses are normally supplied by offsite

power, via a 4160 V balance of plant (BOP) bus, through a master/slave

breaker combination. When performing SR 4.8.1.1.2.d.5, the diesel

generator is started in manual mode and synchronized to the E bus. With

a diesel generator synchronized to the E bus, the diesel generator is

protected from a potential overload condition. Class 1E protective

relaying, at the E bus, is aligned to the trip circuit of the slave

breaker to protect the diesel from an overload condition should the

normal source of power be lost. These relays sense E bus voltage, E bus

frequency, and directional power from the E bus to the BOP bus.

Actuation of any of these relays, with the diesel in manual, will trip

the slave and master breakers to separate the diesel generator from the

BOP bus. This separates the diesel generator from the potential

overload condition. In addition, either a loss of offsite power or loss

of coolant accident results in the diesel generator output breaker

opening, E bus loads stripping, and the diesel generator reverting to

automatic mode. This allows the diesel generator to tie back to the E

bus and carry the E bus loads.

The proposed license amendments reflect the clarification,

previously made to Bases Section 3/4.8, ``Electrical Power Sources,''

in SR 4.8.1.1.2.d itself. Accordingly, SR 4.8.1.1.2.d.2, SR

4.8.1.1.2.d.3, SR 4.8.1.1.2.d.6, and SR 4.8.1.1.2.d.7 are performed for

diesel generator 1 or 2 with BSEP, Unit No. 1 in OPERATIONAL CONDITION

3, 4, or 5 and for diesel generator 3 or 4 with BSEP, Unit No. 2 in

OPERATIONAL CONDITION 3, 4, or 5. Defining the term ``during shutdown''

as ``OPERATIONAL CONDITION 3, 4, or 5'' is consistent with the current

TS requirements of SR 4.8.1.1.2.d. TS Table 1.2, ``OPERATIONAL

CONDITIONS,'' defines five OPERATIONAL CONDITIONS for the BSEP. There

are two OPERATIONAL CONDITIONS applicable to power operation with the

unit critical (i.e., POWER OPERATION and STARTUP) and three OPERATIONAL

CONDITIONS applicable to a subcritical, shutdown unit (i.e., HOT

SHUTDOWN, COLD SHUTDOWN, and REFUELING). Therefore, ``during shutdown''

and ``in OPERATIONAL CONDITION 3, 4, or 5'' have equivalent meaning.

Eliminating the expired BSEP, Unit No. 2 footnote to SR

4.8.1.1.2.d.1 is an administrative change and, therefore, cannot

increase the probability or consequences of an accident previously

evaluated.

Based on the above, the proposed license amendments do not involve

a significant increase in the probability or consequences of an

accident previously evaluated.

2. The proposed license amendments will not create the possibility

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

evaluated.

The proposed license amendments to allow performance of SR

4.8.1.1.2.d.1, SR 4.8.1.1.2.d.4, and SR 4.8.1.1.2.d.5 in OPERATIONAL

CONDITION 1, 2, 3, 4, or 5, rather than only during shutdown, do not

affect the operation or response of any plant equipment, including the

diesel generators, or introduce any new failure mechanism. Plant

systems and equipment will continue to respond in accordance with

design and as analyzed. There will not be a malfunction of a new or

different type introduced by the proposed license amendments.

The proposed license amendments reflect the clarification,

previously made to Bases Section 3/4.8, in SR 4.8.1.1.2.d itself.

Accordingly, SR 4.8.1.1.2.d.2, SR 4.8.1.1.2.d.3, SR 4.8.1.1.2.d.6, and

SR 4.8.1.1.2.d.7 are performed for diesel generator 1 or 2 with BSEP,

Unit No. 1 in OPERATIONAL CONDITION 3, 4, or 5 and for diesel generator

3 or 4 with BSEP, Unit No. 2 in OPERATIONAL CONDITION 3, 4, or 5.

Defining the term ``during shutdown'' as ``OPERATIONAL CONDITION 3, 4,

or 5'' is consistent with the current TS requirements of SR

4.8.1.1.2.d. TS Table 1.2, ``OPERATIONAL CONDITIONS,'' defines five

OPERATIONAL CONDITIONS for the BSEP. There are two OPERATIONAL

CONDITIONS applicable to power operation with the unit critical (i.e.,

POWER OPERATION and STARTUP) and three OPERATIONAL CONDITIONS

applicable to a subcritical, shutdown unit (i.e., HOT SHUTDOWN, COLD

SHUTDOWN, and REFUELING). Therefore, ``during shutdown'' and ``in

OPERATIONAL CONDITION 3, 4, or 5'' have equivalent meaning.

Eliminating the expired BSEP, Unit No. 2 footnote to SR

4.8.1.1.2.d.1 is an administrative change and, therefore, cannot create

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

accident previously evaluated.

Based on the above, the proposed license amendments do not create

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

previously evaluated.

3. The proposed license amendments do not involve a significant

reduction in a margin of safety.

Bases Section 3/4.8, ``Electrical Power Systems,'' states that the

operability of the alternating current (ac) and direct current power

sources and associated distribution systems during operation ensures

that sufficient power will be available to supply the safety-related

equipment required for the safe shutdown of the facility and the

mitigation and control of accident conditions within the facility.

Diesel generator capacity is such that any three of the four diesel

generators can supply the required loads for the safe shutdown of one

unit and a design basis accident on the other unit without relying on

offsite power. Performance of SR 4.8.1.1.2.d.1, SR 4.8.1.1.2.d.4, and

SR 4.8.1.1.2.d.5 during power operation will not affect the operability

of any other safety-related systems, nor will it create any

perturbations of the electrical distribution system that could

challenge plant operation. Class 1E protective relaying, at the E bus,

protects the diesel from an overload condition should the normal source

of power be lost while performing SR 4.8.1.1.2.d.5. There is no

relaxation of any LCO as a result of the

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proposed license amendments. If an additional ac power source becomes

inoperable during the performance of SR 4.8.1.1.2.d.1, SR

4.8.1.1.2.d.4, and SR 4.8.1.1.2.d.5, the units will be placed in the

appropriate OPERATIONAL CONDITION in accordance with TS 3.8.1.1, ``A.C.

Sources Operating.'' Therefore, the diesel generators' ability to

perform their intended safety function, as described in Section

8.3.1.1.6.1 of the BSEP Updated Final Safety Analysis Report, is not

adversely affected by the proposed license amendments.

The proposed license amendments are consistent with the guidance of

Generic Letter 91-04, ``Changes In Technical Specification Surveillance

Intervals To Accommodate A 24-Month Fuel Cycle,'' which concludes that

TSs need not restrict surveillances to only being performed during

shutdown provided that performance of the surveillance during power

operations does not adversely affect safety.

The proposed license amendments reflect the clarification,

previously made to Bases Section 3/4.8, in SR 4.8.1.1.2.d itself.

Accordingly, SR 4.8.1.1.2.d.2, SR 4.8.1.1.2.d.3, SR 4.8.1.1.2.d.6, and

SR 4.8.1.1.2.d.7 are performed for diesel generator 1 or 2 with BSEP,

Unit No. 1 in OPERATIONAL CONDITION 3, 4, or 5 and for diesel generator

3 or 4 with BSEP, Unit No. 2 in OPERATIONAL CONDITION 3, 4, or 5.

Defining the term ``during shutdown'' as ``OPERATIONAL CONDITION 3, 4,

or 5'' is consistent with the current TS requirements of SR

4.8.1.1.2.d. TS Table 1.2, ``OPERATIONAL CONDITIONS,'' defines five

OPERATIONAL CONDITIONS for the BSEP. There are two OPERATIONAL

CONDITIONS applicable to power operation with the unit critical (i.e.,

POWER OPERATION and STARTUP) and three OPERATIONAL CONDITIONS

applicable to a subcritical, shutdown unit (i.e., HOT SHUTDOWN, COLD

SHUTDOWN, and REFUELING). Therefore, ``during shutdown'' and ``in

OPERATIONAL CONDITION 3, 4, or 5'' have equivalent meaning.

Eliminating the expired BSEP, Unit No. 2 footnote to SR

4.8.1.1.2.d.1 is an administrative change and, therefore, cannot

involve a significant reduction in a margin of safety.

Based on the above, the proposed license 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: University of North Carolina

at Wilmington, William Madison Randall Library, 601 S. College Road,

Wilmington, North Carolina 28403-3297.

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

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

Raleigh, North Carolina 27602.

NRC Project Director: James E. Lyons.

Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-324,

Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North

Carolina

Date of amendments request: November 6, 1997.

Description of amendments request: The proposed amendments to

Technical Specification (TS) Limiting Conditions for Operation (LCO)

3.3.5.5, Instrumentation for Control Room Emergency Ventilation System

(CREVS) and 3.7.2, Control Room Emergency Ventilation System, and

associated Bases for the Brunswick Steam Electric Plant (BSEP) Units 1

and 2 would be limited in duration (approximately 3 months) and would

allow operation of both BSEP units to continue while upgrades to the

control building ventilation system, including new air conditioning

(AC) units, are being installed. Part of the planned work requires

opening the ductwork at the evaporative (i.e. cooling) coils. Temporary

barriers will be constructed to preserve the leakage integrity of the

control room pressure boundary; however, the temporary barriers will

not be seismically qualified. While the permanent AC units are out of

service, temporary AC units will be utilized. During the upgrade

installation, the AC for the control room will not be protected from

certain external events (e.g., seismic events, environmental hazards

such as tornadoes and hurricanes, radiological sabotage, and missile

hazards), as required by the system design and licensing basis, and

will not fully meet single failure criteria.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

the probability or consequences of an accident previously evaluated.

The proposed changes do not affect any component of any of the

barriers to radiation release, any of the systems which protect the

core from overheating, nor any system used to shut down the reactor.

The proposed changes do not affect any of the chlorination system

piping or the tank car, which would be the initiating components of a

chlorine release event. The proposed changes affect the CREVS and CREVS

instrumentation, neither of which are accident or event causing

systems. Therefore, the proposed changes do not increase the

probability of an accident or toxic gas release previously analyzed in

the Updated Final Safety Analysis Report (FSAR).

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

mitigate the consequences of a design basis accident or event involving

a release of radioactive material. In addition, the proposed changes do

not significantly affect the ability of the system to mitigate the

consequences of a toxic gas release. The following measures will be

taken to minimize the consequences of accidents and events:

Temporary isolation barriers will be constructed to provide

integrity of the duct during design basis radiation release events.

These temporary barriers will ensure that 10 CFR Part 50, General

Design Criterion 19 for Control Room operator doses is met for all

design basis radiation release accidents.

During the time that the temporary barrier is used, the chlorine

tank car will be removed from the exclusion area. Analyses have shown

that with the chlorine tank car outside of the exclusion area, there is

no threat to Control Room habitability. Removal of the chlorine tank

car from the exclusion area is the current Technical Specification

requirement for inoperability of the Control Room chlorine isolation

mode.

The temporary condensing units for the Control Room Air

Conditioning system will be installed to high quality standards, and a

spare condensing unit will be provided such that two units can be

maintained functional. These units will each be powered from a separate

division of Class 1E power. The operation of the units will be

monitored to ensure that they are in good operating order.

If two or more of the condensing units should fail, instructions

have been provided to the operators for increased monitoring of

temperatures, and mitigating actions are available to the operators if

temperatures rise above a predetermined limit.

Therefore, the consequences of an accident or an event involving a

release

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of radiation, toxic gas, or smoke will not be significantly increased.

In addition, the change will not significantly affect the consequences

of a seismic event or other severe natural phenomena, as previously

analyzed in the Updated FSAR.

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

new or different kind of accident from any accident previously

evaluated.

The proposed changes involve adjustments to the LCO requirements

for CREVS relative to protection from severe natural phenomena. The

proposed changes do not introduce any new modes of plant operation. The

proposed changes do not involve any new modes of system operation,

except that temporary condensing units will be used in place of the

permanent condensing units. The temporary condensing units will

interface with the permanent Control Building Heating Ventilation and

Air Conditioning system in a similar manner to the permanent system.

The piping connections to the permanent system will be the same, and

the controls interface will be the same. No new cross-ties will be

created and no new piping will be run though the habitability boundary.

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

different kind of accident from any accident previously analyzed.

3. The proposed license amendments do not involve a significant

reduction in a margin of safety.

The proposed changes do not represent a significant change in the

assumptions and inputs to the analyses for Control Room operator doses.

No increase in the doses to the Control Room operators is expected

after a seismic event or tornado, since the integrity of existing

barriers to release of radioactive material are not affected.

Therefore, this change does not result in a significant reduction in

the margin of safety for a radiological event.

The proposed change does not represent a change to the leakage

criteria for the Control Room, or the Control Room ventilation

ductwork, following either a toxic gas or external smoke event. The

bounding analysis remains valid, unless the failure is caused by a

tornado or seismic event. Due to the low probability of such an event

occurring during the short time frame involved in this modification,

the occurrence of such an event is not of significant concern.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: University of North Carolina

at Wilmington, William Madison Randall Library, 601 S. College Road,

Wilmington, North Carolina 28403-3297.

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

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

Raleigh, North Carolina 27602.

NRC Project Director: James E. Lyons.

Consumers Energy Company, Docket No. 50-155, Big Rock Point Nuclear

Plant, Charlevoix County, Michigan

Date of amendment request: September 19, 1997 (Accession No.

9709240373).

Description of amendment request: The amendment request propose

changes to the Facility Operating License and technical specifications

(TS) to reflect the permanent cessation of power operations and

permanent transfer of nuclear reactor fuel to the spent fuel pool

(SFP). In particular, Consumers Energy requests to change: safety

limits; limiting safety system settings; limiting control system

settings; limiting conditions for operation; surveillance requirements;

design features; and administrative controls. On November 12, 1997,

Consumers Energy provided supplemental information regarding their no

significant hazards determination, as requested by NRC request for

additional information letter dated October 12, 1997. By letters dated

June 26 and September 23, 1997, the licensee certified permanent

cessation of power operations and permanent removal of all fuel from

the reactor, respectively.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

The proposed change provides the applicable requirements to assure

safe storage of spent nuclear fuel during decommissioning following the

permanent cessation of power operations at the Big Rock Point Nuclear

Plant (BRP) on August 30, 1997 [see Consumers Energy letter to NRC

dated June 26, 1997] and permanent removal of all fuel from the reactor

vessel on September 20, 1997 [see Consumer Energy letter to NRC dated

September 23, 1997]. Decommissioning activities conducted using these

controls do not present undue risk to the public, and do not impact

common defense and security. As such, these changes will not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

No accidents previously evaluated in the Updated Final Hazards

Summary Report (UFHSR) will have their probability of occurrence

increased because the proposed controls effectively preclude the

occurrence of criticality, fuel temperature exceeding limits, or fuel

handling accidents. The probability of plant accidents associated with

power operations have been significantly reduced. Accidents associated

with spent fuel handling, including cask and single bundle drop and

spent fuel cooling capability loss events, are still pertinent and were

reviewed using new data on pool inventory and revised 10 CFR 20

radiological limit determinations. The probability of occurrence of

accidents associated with storing 441 spent fuel assemblies in the SFP

(current license limit) have not been affected by the changes in the

proposed TSs.

The consequences of a fuel handling and cask drop accidents were

evaluated based on the removal of all fuel from the reactor and loading

spent nuclear fuel in the SFP. The removal of all fuel from the reactor

vessel to storage in the SFP and the subsequent decay of the fuel in

the pool result in no increase in the probability of these accidents

and continuously reduced consequences from these accidents.

Analyses using the techniques in Branch Technical Position APCSB 9-

2 provide the heat rate from a freshly-removed full core off-load in

the SFP whose racks are filled with a total of 441 fuel assemblies as

the most limiting cooling condition. Existing cooling equipment under

the current TSs provide sufficient cooling to preclude spent fuel pool

temperatures reaching 150 degrees-Fahrenheit with a complete loss of

spent fuel cooling for 72 hours. This precludes entry into an

unanalyzed condition for the SFP and provides 3 days to recover cooling

flow of ``approximately 30'' gallons per minute. Since this

specification change is intended for implementation following 93 days

after shutdown (approximately November 30, 1997), this analysis

justifies the allowance of 24 hours to re-establish cooling flow

provided in specification 3.1.2.

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

from any accident previously evaluated.

The permanent cessation of power operation and removal of fuel from

the

[[Page 63975]]

reactor eliminates the possibility of the following categories of

accidents and transients to create a hazard to the health and safety of

the public: increase in heat removal by the secondary system; increase

in reactor coolant inventory; decrease in heat removal by the secondary

system; decrease in reactor coolant inventory; reactivity and power

distribution anomalies; anticipated transient without scram; and,

single loop operation. These revised TSs, in combination with

requirements in the UFHSR, provide assurance that fuel handling and

spent fuel cask drop accident, which represent the remaining specific

pertinent accidents analyzed in the ``radioactive release from a

subsystem of component'' category, will not occur. Because the revised

TSs related to fuel handling, spent nuclear fuel storage, and handling

of the spent fuel cask satisfy current license and UFHSR requirements,

no new accidents are created.

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

The safety margins for analyzed accidents are maintained because

the containment structures and redundant control established by the

plant remain in place until the decay of spent fuel has reduced the

source term to levels that analysis confirms do not require the

containment features. ninety three days after permanent cessation of

operations, the spent nuclear fuel at BRP will have decayed to the

point where the added margin from this decay more than compensates for

the removal of the containment as a safety feature, and allows relaxed

controls for the cooling of the SFP.

The Big Rock Point Plant Safety Committee has reviewed this

Facility Operating License and TS change request and has determined

this change does not involve an unreviewed safety question and,

therefore, involves no significant hazards consideration. The proposed

change has been reviewed by the BRP Nuclear Performance Assessment

Department.

The NRC staff has reviewed the licensee's analysis, as provided by

licensee letters dated September 19 and November 12, 1997, and, based

on this review, it appears that the three standards or 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: North Central Michigan College, 1515

Howard Street, Petosky, MI 49770.

Attorney for licensee: Judd L. Bacon, Esquire, Consumers Energy

Company, 212 West Michigan Avenue, Jackson, Michigan 49201.

NRC Project Director: Seymour H. Weiss.

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: March 11, 1993; supplemented August 26,

November 29, December 6, 1993, October 3, 1995, February 27, and

September 3, 1997 (TSC 93-03).

Description of amendment request: The proposed changes would

replace the present Electrical Power Systems section of the Technical

Specifications, Sections 3.7 and 4.6, by consolidating and rearranging

the present specifications, incorporating new specifications, and

formating the section similar to the Babcock and Wilcox Standard

Technical Specifications. The proposed changes would address such

concerns as Keowee hydro station operability, Lee gas turbine

operability, overhead and underground emergency power path operability,

Keowee and Keowee main step-up transformer outage requirements,

surveillance requirements of various components and systems, Oconee

distribution system requirements, protective instrumentation system

requirements, operability of 125 VDC Vital Instrument and Control power

and limiting condition for operation, inverter requirements, Oconee

shutdown requirements related to various components, Keowee unit

extended outage, dc power operability requirements, battery cell

parameter requirements, and various editorial and related Bases

changes.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below.

Duke Power Company (Duke) [currently Duke Energy Corporation] has

made the determination that this amendment request involves a No

Significant Hazards Consideration by applying the standards established

by NRC regulations in 10 CFR 50.92. This ensures 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:

Each accident analysis addressed within the Oconee Final Safety

Analysis Report (FSAR) has been examined with respect to the changes

proposed within this amendment request. Changes included in this

amendment request are provided to assure availability of electrical

power systems for mitigation of Design Basis Accidents (DBAs). As

described within the technical justification, the following types of

changes are included:

(1) Editorial and administrative changes associated with

reformatting the Technical Specification requirements;

(2) Additional restrictions not presently included in the Technical

Specifications such as the addition of requirements for electrical

power systems during cold shutdown and refueling, for the 230 kV

switchyard degraded grid protection system and to delete the special

inoperability period for the Keowee CX transformer;

(3) Technical changes to current requirements to provide clarity

and operational flexibility. These changes maintain the ability of the

electrical power systems to mitigate the consequences of DBAs without a

significant reduction in availability. These changes include the

definition of emergency power paths to include the associated DC

sources and auxiliary transformers, the combination of special

inoperability periods for ``planned'' and ``unplanned'' reasons, and

the ability to use the Keowee special inoperability period more than

once in a three year period; and

(4) Relocation of requirements which are unnecessary for the

mitigation of DBAs to licensee controlled documents. Relocated

requirements include surveillance requirements for the External Grid

Trouble Protection system.

Based on the above and the technical justification * * *, there is

no significant increase in the probability of DBA as a result of this

change, nor is there a significant increase in the consequences of a

DBA as a result of this change since the proposed amendment assures

availability of electrical power systems.

(2) Create the possibility of a new or different kind of accident

from any kind of accident previously evaluated:

The proposed changes make no physical changes to the plant

configuration and do not adversely affect the performance of any

equipment. Operation of ONS [Oconee Nuclear Station] in accordance with

these Technical Specifications will not create any failure modes not

bounded by previously evaluated accidents. Consequently, this change

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

from any kind of accident previously evaluated.

[[Page 63976]]

(3) Involve a significant reduction in a margin of safety:

Margins of safety associated with these Technical Specifications

have been evaluated. These changes include editorial and administrative

changes associated with reformatting the Technical Specification

requirements, additional restrictions not presently included in the

Technical Specifications, technical changes to current requirements

which maintain the ability of the electrical power systems to mitigate

the consequences of DBAs, and relocation of requirements which are

unnecessary for the mitigation of DBAs to licensee controlled

documents. The design basis of auxiliary electrical systems is to

supply the required ES [emergency system] loads of one Unit and safe

shutdown loads of the other two units. The proposed amendment does not

affect any safety limits, setpoints, or design parameters and assures

the continued availability of electrical power systems; thus preserving

the existing margin of safety. Therefore, there will be no significant

reduction in any margin of safety.

Duke has concluded based on the above, and the technical

justification * * * that there are no significant hazards

considerations involved in this amendment request.

The NRC 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: J. Michael McGarry, III, Winston and Strawn,

1200 17th Street, NW., Washington, DC.

NRC Project Director: Herbert N. Berkow.

Duquesne Light Company, et al., Docket No. 50-334, Beaver Valley Power

Station, Unit No. 1, Shippingport, Pennsylvania

Date of amendment request: November 4, 1997.

Description of amendment request: The proposed amendment would

change Item 6.a.2, ``4.16 Emergency Bus (Start Diesel),'' of Table 3.3-

4 of Technical Specification 3.3.2.1. The proposed change would reduce

the trip setpoint for starting the emergency diesel generators on

emergency bus undervoltage from a trip setpoint of greater than or

equal to 83 percent with a 12-cycle delay time to greater than or equal

to 75 percent of nominal bus voltage with a time delay of less than 0.9

second including auxiliary relay times. The proposed change would also

reduce the allowable value from greater than or equal to 81 percent of

nominal bus voltage to greater than or equal to 74 percent of nominal

bus voltage with a time delay of less than 0.9 second including

auxiliary relay times.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The proposed change replaces the current Engineered Safety Feature

setpoint, allowable value and delay time for the diesel generator start

on loss of power function. An analysis has been performed to develop

the new values to minimize the diesel generator starts when a Reactor

Coolant Pump (RCP) is being started or a fast bus transfer occurs. The

heat generated by an increase in motor current, in response to reduced

voltage, will be less than the heat generated during motor starting.

The analysis results show that bus voltages may dip below the allowable

setpoint value and then recover to the pick-up setpoint within the

proposed delay time without stalling motors.

The proposed change does not affect the design and reliability of

any plant equipment; therefore, the probability of occurrence of a

previously evaluated accident is not increased. The operation of the

plant will not be changed as a result of this proposed amendment,

except that fewer diesel generator starts will be initiated.

This function anticipates the loss of voltage to protect equipment

connected to the 4.16 Kv emergency bus. The UFSAR [Updated Final Safety

Analysis Report] accident analyses do not take credit for this

function; therefore, the consequences of an accident previously

evaluated is not increased.

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

kind of accident from any accident previously evaluated?

The proposed change to the trip setpoint, allowable value and delay

time will continue to ensure that the safety-related equipment

connected to the emergency bus is adequately protected from a low

voltage condition. These setting changes will minimize the diesel

generator starts due to voltage drops when an RCP is started or a fast

bus transfer occurs.

The new setpoint and time delay allow normal voltage drops to occur

during expected plant operations without causing any thermal damage to

safety-related equipment. The performance of the safety system will

remain unchanged and will not alter any plant equipment, performance

requirements or safety analysis. Therefore, the proposed change does

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

any accident previously evaluated.

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

safety?

The proposed change does not involve a significant reduction in a

margin of safety since an analysis has been performed to verify that

safety-related equipment connected to the emergency bus is adequately

protected from a low voltage condition with the proposed settings. The

proposed changes do not affect the UFSAR design bases, accident

assumptions, or technical specification bases. In addition, the

proposed changes do not affect release limits, monitoring equipment or

plant operating practices. 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: B. F. Jones Memorial Library,

663 Franklin Avenue, Aliquippa, PA 15001.

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

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

NRC Project Director: John F. Stolz.

Entergy Operations, Inc., et al., Docket No. 50-416, Grand Gulf Nuclear

Station, Unit 1, Claiborne County, Mississippi

Date of amendment request: October 28, 1997.

Description of amendment request: The amendment would (1) revise

the frequency of conducting five Surveillance Requirements (SRs) and

(2) add a 10 CFR Part 50, Appendix J Testing Program for Primary

Containment Systems in the Technical Specifications (TSs) for Grand

Gulf Nuclear Station, Unit 1 (GGNS). The five SRs are the following: SR

3.6.1.1.1 for primary containment, SR 3.6.1.2.1

[[Page 63977]]

for primary containment air locks, and SRs 3.6.1.3.5, 3.6.1.3.8, and

3.6.1.3.9 for primary containment isolation valves. The proposed

revisions for each of the five SRs are to delete the references to SR

3.0.2 not being applicable and change the surveillance frequency from

being ``in accordance with 10 CFR 50, Appendix J, as modified by

approved exemptions'' to ``in accordance with 10 CFR Part 50, Appendix

J, testing program.'' The testing program would be added to Section

5.0, Administrative Controls, of the TSs. Changes to the Bases of the

TSs were also provided in the submittal.

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.

[On April 26, 1995, the licensee was granted an exemption to Appendix J

of 10 CFR Part 50 that allowed performance-based containment leak rate

testing. This exemption will expire on the startup from Refueling

Outage 9, currently scheduled for the spring of 1998. The licensee's

proposed changes to the TSs are to adopt Option B, Performance-Based

Requirements, that is now in Appendix J, but was not in Appendix J in

1995 when the exemption was granted. The technical findings that

support the rulemaking for Option B are in NUREG-1493, ``Performance-

Based Containment Leak Rate Test Program,'' dated September 1995. The

licensee stated in its submittal that its current containment leak rate

testing program meets the requirements of Option B.]

Two initiating events were identified which could be affected by

the proposed changes [in the submittal of October 28, 1997]. An

interfacing system LOCA [(loss-of-coolant accident)] could be caused by

significant leakage of both normally closed isolation valves in systems

with high pressure/low pressure interfaces. Interfacing systems LOCAs

were considered for the LPCI, LPCS, HPCS, and RCIC systems [(i.e., low

pressure coolant injection, low pressure core spray, high pressure core

spray, and reactor core isolation cooling)]. Because the frequency for

testing of these valves will not be changed under this proposal, there

is no increase in the probability or consequences of an accident

[previously evaluated].

The second event evaluated was a LOCA outside containment. In this

case the probability for failure of the MSIVs [(main steam isolation

valves)] and the feedwater isolation valves were calculated and

combined with the frequency of a pipe break outside containment and the

conditional probability of a core melt given a LOCA. The increase in

core damage is extremely small and therefore does not significantly

increase the probability of any previously evaluated accident. Further,

because the testing frequency for MSIVs and feedwater isolation valves

are not being changed, the LOCA outside containment events can be

discounted.

Failure of, or leakage through[,] a containment barrier can[,]

however, increase the consequences of those accidents previously

evaluated. Because the leakage probability for two valves in series to

fail is very small and because all lines isolated by a single

containment isolation valve always have a water seal and cannot act as

a release pathway unless the integrity of the connected system is

compromised, there is no significant increase in the consequences of

any previously evaluated accident.

Containment bypass can also increase the consequences of

[previously] evaluated accidents. Accident sequences involving

containment have been shown to be relatively insignificant by the GGNS

IPE [(Individual Plant Examination]). The potential for [containment]

bypass was analyzed. The analysis showed that the probabilities for

bypass were dominated by failure to close scenarios. Many programs are

in place at GGNS to monitor containment component performance[,] and to

ensure that proper maintenance and repairs are made during the service

life of the containment. Other routine surveillances are performed

periodically to ensure that the valves will close on demand. In fact,

all valves that are required to close for containment isolation and

that are not maintained closed at all times during power operations are

stroke tested quarterly or[,] at a minimum, during each refueling

outage in accordance with ASME [(American Society of Mechanical

Engineers) Boiler and Pressure Vessel Code,] Section XI, Subsection

IWV.

[Based on the above, the proposed changes do not significantly increase

the probability or consequences of an accident previously evaluated.]

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

different kind of accident from any accident previously evaluated.

The request involves the reduction in the local leak rate and the

integrated leak rate testing frequencies [in accordance with Option B

of Appendix J to 10 CFR Part 50]. Extending the test frequencies has no

influence on, nor does it contribute in any way to, the possibility of

a new or different kind of accident or malfunction from those

previously analyzed. The method of performing the test is not changed.

No new accident modes are created by extending the testing intervals.

No safety-related equipment or safety functions are altered as a result

of this change.

[Therefore, the proposed changes do not create the possibility of a new

or different kind of accident from any accident previously evaluated.]

III. The proposed change does not involve a significant reduction

in a margin of safety.

The only margin of safety that has the potential of being impacted

by the proposed changes involves the offsite dose consequences of

postulated accidents which are directly related to containment leakage

rate. The containment isolation system is designed to limit leakage to

La which is defined by the GGNS TSs to be 0.437 percent by weight of

the containment air [volume] per 24 hours at [the containment pressure

of] 11.5 psig (Pa). The limitation on containment leakage rate is

designed to ensure that total leakage volume will not exceed the value

assumed in the accident analyses at the peak accident pressure (11.5

psig, Pa).

To provide additional conservatism, the measured overall integrated

leakage rate is further limited to less than or equal to 0.75 La during

performance of the periodic integrated leakage rate test and to less

than or equal to 0.60 La for type B and C leakage rate tests [of

Appendix J]. This is done to account for the possible degradation of

the containment leakage barriers between [the Appendix J] tests. This

acceptance criteria ensures that an acceptable margin of safety is

being maintained and will not be altered by the proposed changes. The

preservation of this margin will continue to provide for potential

degradation of the leakage barriers between tests.

No change in the method of testing is being proposed. The tests

will continue to be done at full pressure (Pa) or greater [pressure].

The test pressure for primary containment isolation valves will

continue to be applied in the same direction as would be required for

the valve to perform its safety function (unless a different direction

can be shown to be equivalent or conservative). Primary containment

penetrations

[[Page 63978]]

which require Type B leakage rate tests will be performed in the same

manner as before. The Type A test [of Appendix J] will continue to be

performed at full pressure (Pa). Other programs are in place to ensure

that proper maintenance and repairs are performed during the service

life of the primary containment[,] and systems and components

penetrating the primary containment.

No change in the owners allowable leakage rate is being proposed.

These conservative leakage rates ensure that[,] if every penetration

were at its maximum allowable leakage rate, the total containment

leakage would still be below 0.60 La. The effect of multiple

penetration barriers is not considered which provides further

conservatism.

The assessment of risk analysis for the proposed changes concluded

that the overall risk impact of the changes are neutral and essentially

negligible. Any containment isolation barrier allowed to be tested at

less frequent intervals [through performance-based testing of proposed

Option B of Appendix J] will have demonstrated enhanced performance

which minimizes the potential for increased leakage. The assessment

further shows that there is reasonable assurance that an acceptable

level of performance for the containment isolation function can be

maintained. The overall risk impact for the proposed changes are small

enough to be almost indeterminate. No change to the leakage rate

specified in the TSs is being proposed.

[The proposed changes to the TSs are in accordance with Option B of

Appendix J of 10 CFR Part 50.]

[Therefore, the proposed changes do not involve a significant reduction

in a margin of safety.]

Based on the above evaluation, operation in accordance with the

proposed amendment involves no significant hazards consideration.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room Location: Judge George W. Armstrong

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

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

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

NRC Project Director: David A. Wigginton, Acting.

Maine Yankee Atomic Power Company, Docket No. 50-309, Maine Yankee,

Atomic Power Station, Lincoln County, Maine

Date of amendment request: September 30, 1997.

Description of amendment request: The proposed amendment would

eliminate certain license conditions of the Maine Yankee operating

license that are no longer appropriate in the permanently defueled

condition of the plant. These conditions include restrictions on the

Fire Protection Program and implementation of leakage reduction,

airborne iodine monitoring, secondary water chemistry, and cooling

water discharge monitoring programs. By letter dated August 7, 1997,

the licensee certified permanent cessation of power operations and

permanent removal of fuel from the reactor vessel. Most of the

provisions of the Maine Yankee operating license were established to

ensure protection of the public health and safety during power

operations. Maine Yankee has proposed to eliminate those license

requirements that are not relevant to the permanently defueled plant

condition to allow the Maine Yankee staff to focus on those provisions

which are still appropriate during decommissioning.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

The proposed change does not:

1. Involve a significant increase in the probability or consequence

of an accident previously evaluated.

The purpose of the proposed change is to eliminate requirements

which are not appropriate in the permanently defueled plant condition.

Since the plant has permanently ceased operation and will be maintained

in a defueled condition, many provisions of the license related to

operation of the plant are no longer appropriate. Elimination of these

unnecessary requirements allows the plant staff to focus on those

requirements which continue to be appropriate to the existing plant

condition. The proposed change does not affect those Chapter 14

accidents which are appropriate to the current plant conditions: fuel

handling accident, spent fuel cask drop, and radioactive liquid waste

system leaks and failures, and therefore, does not involve a

significant increase in the probability or consequences of an accident

previously evaluated.

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

from any accident previously evaluated.

The purpose of this proposed change is to eliminate requirements

which are not appropriate in the permanently defueled plant condition.

Since the plant has permanently ceased operation and will be maintained

in a defueled condition, many provisions of the license related to

operation of the plant are no longer appropriate. Elimination of these

unnecessary requirements allows the plant staff to focus on those

requirements which continue to be appropriate to the existing plant

conditions. This proposed change does not affect storage of spent fuel

and, therefore, does not create the possibility of a new or different

accident from any accident previously evaluated.