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 February 2, 1998, through February 12, 1998.

The last biweekly notice was published on February 11, 1998 (63 FR

6968).

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

[[Page 9590]]

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 March 27, 1998, the licensee may file a request for a hearing

with respect to issuance of the amendment to the subject facility

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

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

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

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

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

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

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

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

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

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

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

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

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

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

Secretary or the designated Atomic Safety and Licensing Board will

issue a notice of a hearing or an appropriate order.

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

shall set forth with particularity the interest of the petitioner in

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

the proceeding. The petition should specifically explain the reasons

why intervention should be permitted with particular reference to the

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

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

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

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

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

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

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

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

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

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

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

the specificity requirements described above.

Not later than 15 days prior to the first prehearing conference

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

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

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

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

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

brief explanation of the bases of the contention and a concise

statement of the alleged facts or expert opinion which support the

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

contention at the hearing. The petitioner must also provide references

to those specific sources and documents of which the petitioner is

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

facts or expert opinion. Petitioner must provide sufficient information

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

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

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

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

petitioner who fails to file such a supplement which satisfies these

requirements with respect to at least one contention will not be

permitted to participate as a party.

Those permitted to intervene become parties to the proceeding,

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

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

hearing, including the opportunity to present evidence and cross-

examine witnesses.

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

determination on the issue of no significant hazards consideration. The

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

If the final determination is that the amendment request involves

no significant hazards consideration, the Commission may issue the

amendment and make it immediately effective, notwithstanding the

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

of the amendment.

If the final determination is that the amendment request involves a

significant hazards consideration, any hearing held would take place

before the issuance of any amendment.

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

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

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

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

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

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

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

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

Nontimely filings of petitions for leave to intervene, amended

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

not be entertained absent a determination by the Commission, the

presiding officer or the Atomic Safety and Licensing Board that the

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

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

For further details with respect to this action, see the

application for amendment which is available for public inspection at

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

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

the particular facility involved.

[[Page 9591]]

Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-

529, and STN 50-530, Palo Verde Nuclear Generating Station, Units Nos.

1, 2, and 3, Maricopa County, Arizona

Date of amendment request: October 4, 1996, as supplemented by

letters dated June 6, September 19, November 7, and December 16, 1997.

Description of amendment request: The proposed amendment for each

unit identified above would change the distance criterion in Action b

to Limiting Condition for Operation (LCO) 3/4.1.3, ``Movable Control

Assemblies,'' by which more than one full-length or part-length control

element assembly (CEA) is misaligned from any other CEA in its group.

Action b states, in part, that if the misalignment is greater than the

specified distance criterion, the reactor core is to be placed in at

least hot standby within 6 hours. The proposed amendment would reduce

the distance criterion from 19 inches to 9.9 inches, and replace hot

standby in 6 hours by ``open the reactor trip breakers.''

This proposed amendment is included as a ``more restrictive''

change in the conversion of the current Technical Specifications (CTS)

to the Improved Technical Specifications, which was noticed in the

Federal Register (62 FR 18153) on April 14, 1997. The proposed

amendment would be included in Action F to LCO 3.1.5, ``Movable Control

Assemblies,'' of the Improved Technical Specifications. This proposed

amendment is a change to the current Technical Specifications and is in

addition to the six proposed changes to the CTS or proposed deviations

to the Improved Standard Technical Specifications (NUREG-1432) which

were identified in the notice of April 14, 1997.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase in

the probability or consequences of an accident previously evaluated.

The proposed changes provide more stringent requirements than

previously existed in the CTS. The more stringent requirements will not

result in operation that will increase the probability of initiating an

analyzed event. If anything, the new requirements may decrease the

probability or consequences of an analyzed event by incorporating the

more restrictive changes discussed in the specific Discussion of

Changes [for specification 3.1.5]. These changes will not alter

assumptions relative to mitigation of an accident or transient event.

The more restrictive requirements will not alter the operation and will

continue to ensure process variables, structures, systems, or

components are maintained consistent with safety analyses and licensing

basis [for the plant]. These changes have been reviewed to ensure that

no previously evaluated accident has been adversely affected.

Therefore, these changes will not involve a significant increase in the

probability or consequences of an accident evaluated.

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

different kind of accident from any accident previously evaluated.

Making existing requirements more restrictive and adding more

restrictive requirements to the CTS will not alter the plant

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

or change the methods governing normal plant operation. These changes

do impose different requirements. However, they are consistent with the

assumptions made in the safety analyses, licensing basis, and NUREG-

1432 [for the plant]. Therefore, these changes 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 changes provide more stringent requirements than

previously existed in the CTS. An evaluation of these changes concluded

that adding these more restrictive requirements either increases or has

no impact on the margin of safety. The changes provide additional

restrictions which may enhance plant safety. These changes maintain

requirements of the safety analysis, licensing basis, and NUREG-1432

[for the plant]. As such, no question of safety is involved. Therefore,

these changes 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: Phoenix Public Library, 1221

N. Central Avenue, Phoenix, Arizona 85004.

Attorney for licensee: Nancy C. Loftin, Esq., Corporate Secretary

and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail

Station 9068, Phoenix, Arizona 85072-3999.

NRC Project Director: William H. Bateman.

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

Station, Plymouth County, Massachusetts

Date of amendment request: September 19, 1997.

Description of amendment request: The proposed amendment would

relocate the Radioactive Effluent Technical Specifications (RETS) and

the Radiological Environmental Monitoring Program to the Offsite Dose

Calculation Manual (ODCM), in accordance with the recommendations of

Generic Letter 89-01 and NUREG-1433. In addition, changes to other

sections of the TSs are being proposed to align the current TSs with

NUREG-1433.

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?

Operation of PNPS in accordance with the proposed change will not

involve a significant increase in the probability or consequences of an

accident previously evaluated because of the following:

Definitions

Definitions perform a supporting function for other sections of the

TS. The proposed change to incorporate the definition for the Offsite

Dose Calculations Manual (ODCM) into Section 5.0, ``Programs and

Manuals'', subsection 5.5.1 of the proposed TS will carry forward the

requirements contained in the DEFINITION, with minor editorial

rewording to be consistent with NUREG 1433, and result in no technical

changes. Since the requirements will remain, the impact on initiators

of analyzed events or the assumptions assumed in the mitigation of

accidents or transient events will not change. Editorial rewording

(either adding or deleting) and reformatting is proposed to provide

clarity and does not change any technical requirements.

The definitions being proposed for relocation do not impact reactor

operation, identify a parameter which is an initial condition

assumption for a DBA or transient, identify a significant abnormal

degradation of the reactor coolant pressure boundary, and do not

[[Page 9592]]

provide any mitigation of a design basis event.

RAD Effluents

All editorial rewording (either adding or deleting) and renumbering

is made to restructure the section accounting for the requirements

relocated in accordance with Generic Letter 89-01. During the editorial

rewording and renumbering of the Improved Technical Specifications, no

technical changes (either actual or interpretational) to the TS were

made unless they were identified and justified.

Adding a note to clearly indicate that the first sample for noble

gas activity is not required for 31 days after SJAE is placed in

operation has always been considered the intent of this surveillance

requirement. This allowance is consistent with the frequency for the

required surveillance and allows time for concentrations of longer

lived isotopes to reach equilibrium. In addition, other instrumentation

continuously monitors the offgas to alert operators of significant

increases in radioactivity.

The proposed change provides more stringent requirements than

previously existed in the Technical Specifications. The more stringent

requirements will not result in operation that will increase the

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

requirements may decrease the probability or consequences of an

analyzed event by incorporating the more restrictive changes discussed

above. The change will not alter assumptions relative to mitigation of

an accident or transient event. The more restrictive requirements will

not alter the operation of process variables, structures, systems, or

components as described in the safety analyses.

These proposed changes relocate requirements from the Technical

Specifications to the T. S. BASES, FSAR, or ODCM. The licensee

controlled document containing the relocated requirements will be

maintained using the provisions of 10 CFR 50.59 or a change control

process in the Administrative Controls Section of the Technical

Specifications. Since any changes to these licensee controlled

documents will be evaluated per an NRC approved change control process,

no increase in the probability or consequences of an accident

previously evaluated will be allowed.

Basing the potential fission product release rate on gross gamma

activity rate is more representative of the whole body dose that would

be received by an individual at the site boundary should a release

occur. Therefore, reasonable assurance that the potential whole body

accident dose to an individual at the exclusion area boundary will not

exceed a small fraction of the limits specified in 10 CFR Part 100 is

maintained.

Allowing the sample to be taken from either pretreatment monitor

station will have no effect on the objective of assuring that the

potential whole body accident dose to an individual at the exclusion

area boundary will not exceed a small fraction of the limits specified

in 10 CFR Part 100, because both monitor stations are prior to

treatment, adsorption, or delay of the noble gases.

RAD Material Source

The requirements for miscellaneous radioactive materials do not

impact reactor operation, identify a parameter which is an initial

condition assumption for a DBA or transient, identify a significant

abnormal degradation of the reactor coolant pressure boundary, and do

not provide any mitigation of a design basis event.

Major Design Features

The reformatting, renumbering, and rewording along with the other

changes listed involve no technical changes to existing Technical

Specifications. The proposed changes are administrative in nature and

do not impact initiators or assumptions of analyzed accidents or

transient events.

The proposed change provides more stringent requirements than

previously existed in the Technical Specifications. The more stringent

requirements will not result in operation that will increase the

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

requirements may decrease the probability or consequences of an

analyzed event by incorporating the more restrictive changes discussed

above. The change will not alter assumptions relative to mitigation of

an accident or transient event. The more restrictive requirements will

not alter the operation of process variables, structures, systems, or

components as described in the safety analyses.

These proposed changes relocate requirements from the Technical

Specifications to the FSAR. Since any changes to the FSAR must be

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

in the probability or consequences of an accident previously evaluated

will be allowed.

Administrative Controls

The reformatting, renumbering, and rewording along with the other

changes listed involves no technical changes to existing Technical

Specifications. The change to the existing Technical Specifications was

done in order to be consistent with the NUREG-1433. During development

of NUREG-1433, certain wording preferences or English language

conventions were adopted. The proposed change to this section is

administrative in nature and does not impact initiators of analyzed

events. It also does not impact the assumed mitigation of accidents or

transient events.

The proposed change provides more stringent requirements than

previously existed in the Technical Specifications. These more

stringent requirements are administrative in nature (e.g., specifying

additional responsibilities for plant personnel, ensuring overtime

control, incorporating program and manual requirements already in

place, and adding details to reports). These additional requirements

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

equipment will be installed) or changes in methods governing normal

plant operation, not alter assumptions relative to the mitigation of an

accident or transient event, or alter the operation of process

variables, structures, systems, or components as described in the

safety analyses.

This proposed change relocates requirements from the Technical

Specifications to licensee controlled documents. The licensee

controlled documents containing the relocated requirements are required

to meet the applicable regulation and any change process invoked by the

regulation. Since any changes to the licensee controlled document must

continue to meet the regulation, no increase (significant or

insignificant) in the probability or consequences of an accident

previously evaluated will be allowed.

This change proposes to provide flexibility in meeting the minimum

shift staffing for up to two hours in order to provide for unexpected

absence. The proposed change does not affect the probability of an

accident. The actions of an individual are not assumed to be an

initiator of any analyzed event. Also, the change does not negate the

requirement to have licensed individuals in the control room. This

proposed change does not impact the assumptions of any design basis

accident. This change will not alter assumptions relative to the

mitigation of an accident or transient event.

This change proposes to relax the requirement to have an individual

qualified in radiation protection procedures to be onsite when fuel is

in the reactor. The proposed change will allow the position to be

vacant for up

[[Page 9593]]

to two hours in order to provide for unexpected absence.

The proposed change does not affect the probability of an accident.

The actions of an individual qualified in radiation protection

procedures are not assumed to be an initiator of any analyzed event.

Also, the consequences of an accident are not affected by the presence

of an individual qualified in radiation protection. This proposed

change does not impact the assumptions of any design basis accident.

This change will not alter assumptions relative to the mitigation of an

accident or transient event. This change will not have any impact on

the plant safety because the presence of a person qualified in

radiation protection is not required for the mitigation of any

accident.

This change proposes to relax the requirement for submitting the

Radioactive Effluent Release Report and to relocate the report details

outside the TS. The current TS require the report to be submitted semi-

annually. This proposed change will allow the report to be submitted

annually as required by 10 CFR 50.36a. The proposed change does not

affect the probability of an accident. Neither the submittal

requirements nor the contents of the Radioactive Effluent Release

Report is assumed to be an initiator of any analyzed event. Also, the

consequences of an accident are not affected by submittal requirements

nor the contents of the Radioactive Effluent Release Report. This

proposed change does not impact the assumptions of any design basis

accident. This change will not alter assumptions relative to the

mitigation of an accident or transient event. This change has no impact

on the safe operation of the plant. The report will still be required

to be submitted and does not affect any plant equipment or requirements

for maintaining plant equipment. The submittal of this report is not

required for the mitigation of any accident.

The proposed alternatives for control of access to high radiation

areas are consistent with the intent of 10 CFR 20.1601(a) and (b). The

proposed changes do not affect the probability of an accident. The

controls used for access to high radiation areas are not assumed in the

initiation of any analyzed event. Also, the consequences of an accident

are not affected by these changes. These changes are both consistent

with good radiological safety practice and will provide an adequate

level of radiation protection. These proposed changes do not impact the

assumptions of any design basis accident. These changes will not alter

assumptions relative to the mitigation of an accident or transient

event. These changes have no impact on safe operation of the plant.

Radiological Environmental Monitoring

The proposed changes only alter the format and location of

procedural details and administrative controls of the radioactive

effluents, radiological environmental monitoring, and solid radioactive

waste programs. The changes are administrative in nature and do not

involve any change to the configuration or operation of plant

equipment. The Radiological Effluent Technical Specifications (RETS)

procedural details are being moved to the Offsite Dose Calculation

manual (ODCM). In addition, new administrative controls have been added

to the Technical Specifications which will provide an equivalent level

of assurance that activities involving radioactive effluents, solid

radioactive waste, and radiological environmental monitoring are

conducted in full compliance with regulatory requirements. Since any

changes to these requirements will require NRC approval, no increase in

the probability or consequences of an accident previously evaluated

will be allowed.

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

kind of accident from any accident previously evaluated?

Operation of PNPS in accordance with the proposed change will not

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

accident previously evaluated because of the following:

Definitions

These proposed changes do not involve a physical alteration of the

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

changes in methods governing normal plant operation. The proposed

change will not impose any new or different requirements or eliminate

any existing requirements.

Relocating these definitions will not alter the plant configuration

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

methods governing normal plant operation. Relocating requirements will

not impose different requirements and adequate control of information

will be maintained. Relocating these definitions will not alter

assumptions made in the safety analysis and licensing basis.

RAD Effluents

The proposed change does not involve a physical alteration of the

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

changes in methods governing normal plant operation. The proposed

change will not impose any new or different requirements or eliminate

any existing requirements.

Making existing requirements more restrictive and adding more

restrictive requirements to the Technical Specifications will not alter

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

installed) or change methods governing normal plant operation. These

changes are consistent with current design bases, licensing bases or

assumptions made in the safety analysis.

These changes do not alter the plant configuration (no new or

different type of equipment will be installed) or methods governing

normal plant operation. These changes will not impose different

requirements and adequate control of information will be maintained.

These changes do not alter assumptions made in the safety analysis and

licensing basis.

The proposed change will not involve a physical alteration of the

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

changes in methods governing normal plant operation. Operation of the

plant will not be altered by this change. This change will not place

the plant in any new condition or introduce any mode of operation not

previously analyzed.

The proposed change will not involve a physical alteration of the

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

changes in methods governing normal plant operation. Operation of the

plant will not be altered by this change. This change will not place

the plant in any new condition or introduce any mode of operation not

previously analyzed.

RAD Material Source

Relocating these requirements will not alter the plant

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

or change methods governing normal plant operation. Relocating

requirements will not impose different requirements and adequate

control of information will be maintained. Relocating requirements does

not alter assumptions made in the safety analysis and licensing basis.

Major Design Features

The proposed change does not involve a physical alteration of the

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

changes in methods governing normal plant operation. The proposed

change will not impose any new or different requirements or eliminate

any existing requirements.

[[Page 9594]]

Making existing requirements more restrictive and adding more

restrictive requirements to the Technical Specifications will not alter

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

installed) or changes in methods governing normal plant operation. The

change does impose different requirements. However, the change is

consistent with assumptions made in the safety analyses.

These changes relocate requirements to the FSAR. These changes do

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

equipment will be installed) or the methods governing normal plant

operation. These changes do not impose different requirements and

adequate control of information will be maintained. This change will

not alter assumptions made in the safety analysis and licensing basis.

Administrative Controls

The proposed change does not involve a physical alteration of the

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

changes in methods governing normal plant operation. The proposed

change will not impose any new or different requirements or eliminate

any existing requirements.

Making existing requirements more restrictive and adding new

requirements to the Technical Specifications will not alter the plant

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

or changes in the methods governing normal plant operation.

This change relocates requirements to a licensee controlled

document. This change will not alter the plant configuration (no new or

different type of equipment will be installed) or changes in methods

governing normal plant operation. This change will not impose different

requirements and adequate control of information will be maintained.

This change will not alter assumptions made in the safety analysis and

licensing basis.

This change proposes to provide flexibility in meeting the minimum

shift staffing for up to two hours in order to provide for an

unexpected absence. The proposed change will not create the possibility

of an accident. This change will not physically alter the plant (no new

or different type of equipment will be installed).

This change proposes to relax the requirement to have an individual

qualified in radiation protection procedures to be onsite when fuel is

in the reactor. The proposed change will allow the position to be

vacant for up to two hours in order to provide for unexpected absence.

The proposed change will not create the possibility of an accident.

This change will not physically alter the plant (no new or different

type of equipment will be installed) or the methods of operation.

This change will not physically alter the plant (no new or

different type of equipment will be installed). The changes in methods

governing normal plant operation are consistent with the current safety

analysis assumptions.

The proposed change will not create the possibility of an accident.

This change will not physically alter the plant (no new or different

type of equipment will be installed). The changes in methods governing

normal plant operation are consistent with the current safety analysis

assumptions.

Radiological Environmental Monitoring

The procedural requirements of the RETS will be maintained in the

ODCM. Operation of the plant will not be altered by the changes

proposed to the administration of the RETS. This change will not place

the plant in any new condition or introduce any mode of operation not

previously analyzed.

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

safety?

Operation of PNPS in accordance with the proposed change will not

involve a significant reduction in a margin of safety because of the

following:

Definitions

Definitions perform a supporting function for other sections of the

TS and the proposed editing, omission or relocation of definitions

associated with this change will not, by itself, reduce existing

restrictions on plant operations.

The definitions to be transposed from the Technical Specifications

to the ODCM are the same as the existing Technical Specifications.

Future changes to the ODCM will be controlled in accordance with

proposed technical specification 5.5.1 ``Offsite Dose Calculation

Manual (ODCM)''.

RAD Effluents

The change is administrative in nature and does not involve any

technical changes. The proposed change will not reduce a margin of

safety because it has no impact on any safety analysis assumptions.

Also, because the change is administrative in nature, no question of

safety is involved.

Adding these new requirements and making existing ones more

restrictive does not affect any safety analysis assumptions. As such,

no question of safety is involved.

The requirements to be relocated from the Technical Specifications

to the FSAR T.S. BASES, or ODCM are the same as the existing Technical

Specifications and any future changes to this licensee controlled

document will be evaluated per an NRC approved change control process.

Specifying a release rate based only on gamma activity is more

representative of the whole body dose that would be received by an

individual at the site boundary should a release occur. The actual

margin of safety could be increased because potential errors in

converting beta activity to whole body exposures are eliminated

The sample used to determine the gaseous activity rate will

continue to be taken prior to treatment, adsorption, or delay of the

noble gases.

RAD Material Source

This change relocates requirements from the Technical

Specifications to a licensee controlled document. This change will not

reduce a margin of safety since it has no impact on any safety analysis

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

Technical Specifications to the licensee controlled documents are the

same as the existing Technical Specifications. Since any future changes

to these licensee controlled documents must be evaluated per the cited

regulations or requirements of 10 CFR 50.59, no reduction (significant

or insignificant) in a margin of safety will be allowed.

Major Design Features

The changes are administrative in nature and do not involve any

technical changes. The proposed changes do not impact initiators or

assumptions of analyzed accidents or transient events.

These new or more restrictive requirements are consistent with the

current design and licensing bases; therefore, a margin of safety is

not affected.

These changes relocate requirements from the Technical

Specifications to the FSAR. The requirements to be are the same as the

existing Technical Specifications. Since any future changes to the FSAR

must be evaluated per the requirements of 10 CFR 50.59, no reduction

(significant or insignificant) in a margin of safety will be allowed.

Administrative Controls

The change is administrative in nature and will not involve any

technical changes. The proposed change will not reduce a margin of

safety because it has no impact on any safety analysis assumptions.

[[Page 9595]]

Adding these new requirements and making existing ones more

restrictive does not introduce any new tests or changes in methods

governing normal plant operation. Therefore, the changes do not impact

any safety analysis assumptions.

This change relocates requirements from the Technical

Specifications to a licensee controlled document. The licensee

controlled documents containing the relocated requirements are required

to meet the applicable regulation and any change process invoked by the

regulation. Since any changes to a licensee controlled document must

continue to meet the regulation, no increase (significant or

insignificant) in the probability or consequences of an accident

previously evaluated will be allowed.

This change proposes to provide flexibility in meeting the minimum

shift staffing for up to two hours in order to provide for unexpected

absence. This proposed change has no effect on the assumptions of a

design basis accident. The safety analysis assumptions will still be

maintained; thus, no question of safety exists.

This change proposes to relax the requirement to have an individual

qualified in radiation protection procedures to be onsite when fuel is

in the reactor. The proposed change will allow the position to be

vacant for up to two hours in order to provide for unexpected absence.

The margin of safety is not affected by the presence or absence on site

of an individual qualified in radiation protection procedures. This

proposed change has no effect on the assumptions of the design basis

accident. This change will not have any impact on the plant safety

because the presence of a person qualified in radiation protection is

not required for the mitigation of any accident. The safety analysis

assumptions will still be maintained; thus, no question of safety

exists.

This proposed change has no effect on the assumptions of the design

basis accident. This change has no impact on the safe operation of the

plant. The report will still be required to be submitted and does not

affect any plant equipment or requirements for maintaining plant

equipment. The safety analysis assumptions will still be maintained;

thus, no question of safety exists.

The proposed alternatives for control of access to high radiation

areas are consistent with the intent of 10 CFR 20.1601(a) and (b). The

margin of safety is not reduced due to these proposed changes. These

changes are both consistent with good radiological safety practices and

have been found to provide an adequate level of radiation protection.

In addition, these changes provide the benefit of ensuring radiation

dose to all workers is minimized by providing the flexibility to select

the best means of providing a barrier and access control to a high

radiation area given the plant location and radiological conditions.

These proposed changes have no impact on the safe operation of the

plant. The safety analysis assumptions will still be maintained; thus,

no question of safety exists.

Radiological Environmental Monitoring

The proposed changes relocate the procedural details and Bases for

RETS from the Technical Specifications to the ODCM. The RETS procedural

details and Bases will be maintained by these programs. In addition,

new administrative controls have been added to the Technical

Specifications which assure the proper control and maintenance of these

documents and provide an equivalent level of assurance that activities

involving radioactive effluents, solid radioactive waste, and

radiological environmental monitoring are conducted in full compliance

with regulatory requirements.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Plymouth Public Library, 11

North Street, Plymouth, Massachusetts 0236.

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

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

NRC Project Director: Cecil O. Thomas.

Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455,

Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois, Docket Nos.

STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, Will

County, Illinois

Date of amendment request: November 7, 1997.

Description of amendment request: The proposed amendment would

revise the technical specifications and associated bases to allow the

licensee to perform 10 CFR Part 50, Appendix J, Type A testing on

Byron, Unit 2, and Braidwood, Unit 2, containments at least once per 10

years based on a single successful Type A test, rather than two

successful Type A tests.

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.

Performance of Type A tests at a different interval does not

involve a change to any structures, systems, or components, does not

affect reactor operations, is not an accident initiator, and does not

change any existing safety analysis previously evaluated in the UFSAR

[Updated Final Safety Analysis Report]. Therefore, there is no

significant increase in the probability of an accident previously

evaluated.

Several tables of UFSAR Chapter 15, ``Accident Analyses,'' provide

containment leak rate values used in assessing the consequences of

accidents discussed in this chapter. Although decreasing the test

frequency can increase the probability that an increase in containment

leakage could go undetected for an extended period of time, the risk

resulting from this proposed change is inconsequential as documented in

NUREG-1493, ``Performance-Based Containment Leakage Test Program''.

This document indicated that given the insensitivity of reactor risk to

containment leakage rate and a small fraction of leakage paths are

detected solely by Type A testing, increasing the interval between

integrated leak rate tests is possible with minimal impact on public

risk. Further, industry experience presented in this document indicated

that Type A testing has had insignificant impact on uncertainties

involved with containment leak rates.

Based on risk information presented in NUREG-1493, the proposed

change does not increase the probability or consequences of an accident

previously evaluated.

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

different kind of accident from any accident previously evaluated.

The proposed change does not alter the plant design, systems,

components, or reactor operations, only the frequency of test

performance. New conditions or parameters that contribute to the

initiation of accidents would not be created as a result of this

proposed change. The change does not involve new equipment and existing

equipment does not have to be operated in a

[[Page 9596]]

different manner, therefore there are no new failure modes to consider.

Changing test intervals as shown in NUREG-1493 has no impact on,

nor contributes to the possibility of a new or different kind of

accident as evaluated in the UFSAR. Therefore, the proposed change does

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

any accident previously evaluated.

3. The proposed change does not involve a significant reduction in

a margin of safety.

With the exception of the test frequency, the actual tests will not

change. Quantitative risk studies documented in NUREG-1493 regarding

extended testing intervals demonstrated that there was minimal impact

on the public health and safety. Reducing the frequency, as stated in

the NUREG resulted in an ``imperceptible'' increase in risk to public

safety. Further, a table in this NUREG regarding risk impacts due to a

reduction in testing frequency suggested that there was also minimal

difference in risk to the public safety when the test frequency was

relaxed.

The proposed change will not reduce the availability of systems and

components associated with containment integrity that would be required

to mitigate accident conditions nor are any containment leakage rates,

parameters or accident assumptions affected by the proposed change.

The proposed change does not involve a significant reduction in a

margin of safety, based on the above information.

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: For Byron, the Byron Public

Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010;

for Braidwood, the Wilmington Public Library, 201 S. Kankakee Street,

Wilmington, Illinois 60481.

Attorney for licensee: Michael I. Miller, Esquire; Sidley and

Austin, One First National Plaza, Chicago, Illinois 60603.

NRC Project Director: Robert A. Capra.

Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455,

Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois, Docket Nos.

STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, Will

County, Illinois

Date of amendment request: December 30, 1997.

Description of amendment request: The proposed amendment would

revise the Technical Specification (TS) 3.7.1.3, ``Condensate Storage

Tank,'' (CST) and its associated Bases for Byron and Braidwood to raise

the minimum allowable CST level to ensure that a sufficient volume of

water is available to meet the design basis requirements for the

auxiliary feedwater (AFW) system supply. The proposed amendment would

also revise the AFW system transfer to essential service water (SX)

trip setpoint and allowable value in Table 3.3-4 to ensure that the

design basis requirements for the AFW system are accurately reflected

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. The change does not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The amount of water in the CST [Condensate Storage Tank] at the

beginning of an accident and the setpoint for AF [auxiliary feedwater]

pump suction pressure-low trip have no impact on the probability of

occurrence of any accident analyzed in the UFSAR [Updated Final Safety

Analysis Report]. This is due to the availability of the safety-related

SX [essential service water] water supply as a backup system.

Therefore, the probability of an accident previously evaluated is

unchanged.

The loss of the Safety Category II CST under accident conditions

has already been evaluated in the UFSAR. The SX system is the emergency

source of water supply to the AF system under accident conditions. The

design basis analysis for the essential service water (SX) system and

the Limiting Condition for Operation requirements for the ultimate heat

sink ensure that a sufficient supply of water is available to plant

operators to mitigate the consequences of all analyzed accidents. None

of the proposed changes to the CST minimum level or the setpoints

documented in TS Table 3.3-4, functional unit 6.g. has any negative

impact on the assumptions or results of these analyzed accidents. To

the contrary, the proposed changes will ensure that the CST remains

available as the primary supply of water to the AF system and that

automatic suction transfer will occur for circumstances where the

Safety Category II CST becomes unavailable (e.g., seismic event or

tornado).

The level in the CST and the associated instrumentation and

setpoints help ensure that sufficient water is available to plant

operators to mitigate the consequences of accidents that are analyzed

in the UFSAR. The SX system is the emergency source of water credited

in the UFSAR. However, the proposed Technical Specification Bases

require that sufficient water be maintained in the CST to respond to

postulated events where the CST remains available (e.g., non-seismic

related events and events with no tornado assumed). The proposed CST

levels ensure that this requirement is met. The water level requirement

for the CST provides additional assurance that plant operators remain

capable of responding to postulated events as described in the UFSAR.

Therefore, the proposed changes do not increase the consequences of an

accident previously evaluated.

Therefore this proposed amendment does not involve a significant

increase in the probability or consequences of an accident previously

evaluated.

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

kind of accident from any accident previously evaluated.

The proposed changes are being implemented to account for

instrument accuracy and AF system suction requirements that affect the

volume of useable water in the CST. The amendment request incorporates

the full design requirements of the AF System and components to ensure

that sufficient water is maintained in the CST. The changes reduce the

probability of an undesirable introduction of lower quality essential

service (SX) system water into the steam generators unless required due

to the unavailability of the CST during emergency conditions (e.g.,

seismic event or tornado). Although the SX system is the safety-related

water supply to AF, the water contains high levels of impurities and

sediment that could eventually degrade the steam generators. The CST

contains demineralized water. Therefore, the long term reliability and

availability of the steam generators is enhanced by precluding

introduction of SX water into the steam generators unless required

under emergency conditions. The proposed CST levels account for the

incremental increase in CST water

[[Page 9597]]

volume required due to the larger metal mass and primary volume of the

replacement steam generators for Byron Unit 1 and Braidwood Unit 1.

Finally, the trip setpoint and allowable values in Table 3.3-4 of the

TS are being updated to reflect the current design basis of the AF

system. The required CST level changes when plant modifications are

completed. Each configuration has been evaluated and the associated CST

level maintains a sufficient water volume to perform its design

function.

The modification to the suction pressure circuitry involves the

addition of an electronic ``lead-lag'' circuit card for the motor-

driven AF pump, which experiences the most severe startup suction

pressure transients. This circuit card will be set up for ``lag'' only

operation and will filter the suction pressure signal during transients

associated with pump startup or other sudden changes in flow or

pressure. This will prevent an inadvertent trip during transient

conditions when the CST is available. In situations where the CST is

unavailable, the suction pressure will decrease with no recovery until

switchover. Under this condition, the output of the lead-lag card will

continue to decrease as well until the switchover setpoint is reached.

The time constant of the lead-lag card was selected such that the

resulting time delays in actuating SX switchover and pump trip are

consistent with pump protection requirements.

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

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

conclusion is also valid when considering the planned modifications to

the AF suction pressure transient circuitry.

3. The change does not involve a significant reduction in a margin

of safety.

The proposed change is made in the conservative direction with

respect to the current TS requirements for minimum CST level and AF

pump CST to SX switchover setpoints. Increasing the volume of water

contained in the CST level provides redundancy to the safety-related

source of water to the AF supply, which is the SX system. In

combination, the CST and the SX system ensure that sufficient water is

available to feed the steam generators under all anticipated normal and

emergency conditions to cool a unit from full power conditions down to

350 degrees Fahrenheit, when the residual heat removal system can be

placed into service. The proposed changes ensure the CST will have

sufficient water to meet all normal operating conditions and mitigate

the consequences of all analyzed accidents except those that result in

CST unavailability. In addition, automatic switchover of the AF water

supply from the CSTs to SX will occur as assumed in the current safety

analyses for events where the CST becomes unavailable. The SX system

remains capable of supplying the emergency source of water to the AF

supply.

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: For Byron, the Byron Public

Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010;

for Braidwood, the Wilmington Public Library, 201 S. Kankakee Street,

Wilmington, Illinois 60481.

Attorney for licensee: Michael I. Miller, Esquire; Sidley and

Austin, One First National Plaza, Chicago, Illinois 60603.

NRC Project Director: Robert A. Capra.

Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County,

Michigan

Date of amendment request: January 28, 1998 (NRC-98-0002).

Description of amendment request: The proposed amendment would

revise technical specification (TS) surveillance requirements

4.8.2.1.a.2, 4.8.2.1.b, and 4.8.2.1.c.4 to accommodate differences in

the monitored parameters between the existing batteries and the

batteries that will be installed for Division II during the sixth

refueling outage.

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 changes do not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The proposed changes do not involve a change in the manner in which

the plant is operated. TS Section 4.8.2.1 is being revised to reflect

the new Division II battery cell/system characteristics and associated

requirements. The new battery will have an increased capacity over the

present battery, while maintaining the existing battery system voltage

requirements. This is possible because the present and new battery

specific gravity (1.215) and type (lead calcium) are the same. Also,

the end of battery system discharge voltage remains the same as 210

VDC. The Division II batteries will continue to furnish power to

redundant essential loads as required and as designed. The new

surveillance requirement voltages are based on the same volts/cell

criteria used for the existing batteries. Furthermore, failure or

malfunction of the station batteries does not initiate any of the

analyzed accidents previously evaluated in the UFSAR [updated final

safety analysis report]. The changes described will therefore not

involve an increase in the probability or consequences of an accident

previously evaluated.

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

kind of accident from any accident previously evaluated.

The new battery is Class 1E qualified equipment and is being

maintained within the same overall design parameters as the existing

battery. That is, the battery terminal voltage on float voltage

conditions (2.167 volt[s]/cell), overvoltage conditions (2.5 volts/

cell) and charger capability (2.15 volts/cell) are the same as the

original design. Furthermore, the end of system discharge voltage of

the battery system is maintained the same; therefore, there is no

negative impact to plant loads supplied by the batteries. Failures of

the batteries and chargers have been considered in both the existing

and modified configurations. The proposed changes will not change

performance or reliability nor introduce any new or different failure

modes or common mode failure and will therefore not create the

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

previously evaluated.

3. The changes do not involve a significant reduction in the margin

of safety.

The changes act to increase overall battery capacity from 560

ampere-hours to 1200 ampere-hours with the minimum battery discharge

voltage remaining at 210 VDC (or 105 VDC per battery). The battery

terminal voltage on float voltage conditions (2.167 volt[s]/cell),

overvoltage conditions (2.5 volts/cell) and charger capability (2.15

volts/cell) are the same as the original design. The new surveillance

requirement voltages are based on the same volts/cell criteria used for

the existing batteries. The batteries' ability to satisfy the design

requirements (battery duty cycle) of the dc system will not be reduced

from original plant design and will therefore not have any negative

impact to plant loads the battery supplies. The

[[Page 9598]]

proposed changes therefore do 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: Monroe County Library System,

3700 South Custer Road, Monroe, Michigan 48161.

Attorney for licensee: John Flynn, Esq., Detroit Edison Company,

2000 Second Avenue, Detroit, Michigan 48226.

NRC Project Director: Cynthia A. Carpenter.

Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County,

Michigan

Date of amendment request: January 28, 1998 (NRC-98-0003).

Description of amendment request: The proposed amendment would

revise technical specification (TS) 3.4.10, TS Figure 3.4.10-1 and the

associated bases by changing the prohibited and restricted operating

regions associated with core thermal-hydraulic stability. TS 3.4.1.4,

TS Figure 3.4.1.4-1, and the associated bases would also be revised to

reflect stability-related improvements in operating restrictions for

idle recirculation loop startup. Finally, in an unrelated change, TS

Tables 3.3.7.5-1 and 4.3.7.5-1 would be revised to delete neutron flux

from the parameters the licensee is required to monitor by TS 3.3.7.5,

Accident Monitoring Instrumentation.

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:

Thermal Hydraulic Stability and Idle Recirculation Loop Startup

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

the probability or consequences of an accident previously evaluated.

These changes act to prohibit operations which have been found to

carry a significant potential for the formation of core thermal-

hydraulic instabilities and eliminates inappropriate technical

specifications for maintaining eff to greater than 0.95

is not a credible event. Neither implementation of credit for the

reactivity of fuel pool soluble boron nor the proposed increase in the

fuel pool storage capacity will create the possibility of a new or

different type of accident at St. Lucie Unit 2.

An examination of the limiting fuel assembly misload has determined

that this would not represent a new or different type of accident. None

of the other accidents examined as a part of this license submittal

represent a new or different type of accident; each of these situations

has been previously analyzed and determined to produce acceptable

results.

The proposed license amendment will not result in any other changes

in the mode of spent fuel pool operation at St. Lucie Unit 2 or in the

method of handling irradiated nuclear fuel. The spatial relationship

between the fuel storage racks and the cask crane range of motion is

not affected by the proposed change.

As a result of the evaluation and supporting analyses, FPL has

determined that the proposed fuel pool capacity increase does not

create the possibility of a new or different type of accident from any

accident previously evaluated.

3. The proposed amendment will not involve a significant reduction

in the margin of safety.

FPL has determined, based on the nature of the proposed license

amendment that the issue of margin of safety, when applied to this fuel

pool capacity increase, should address the following areas:

(1) Fuel Pool reactivity considerations

(2) Fuel Pool boron dilution considerations

(3) Thermal-Hydraulic considerations

(4) Structural loading and seismic considerations

The Technical Specification changes proposed by this license

amendment, the proposed spent fuel pool storage configuration and the

existing Technical Specification limits on fuel pool soluble boron

concentration provide sufficient safety margin to ensure that the array

of fuel assemblies stored in the spent fuel pool will always remain

subcritical. The revised spent fuel storage configuration is based on a

Unit 2 specific criticality analysis performed using methodology

consistent with that approved by the NRC. Additionally, the soluble

boron concentration required by current Technical Specifications

ensures that the fuel pool keff will always be maintained

substantially less than 0.95.

The Unit 2 criticality analysis established that the

keff of the spent fuel pool storage racks will be less than

1.0 with no soluble boron in the fuel pool water, including the effect

of all uncertainties and tolerances. Credit for the soluble boron

actually present is used to offset uncertainties, tolerances, off-

normal conditions and to provide margin such that the spent fuel pool

keff is maintained less than or equal to 0.95. FPL has also

demonstrated that a decrease in the fuel pool boron concentration such

that keff exceeds 0.95 is not a credible event.

Current Technical Specifications require that the fuel pool boron

concentration be maintained greater than or equal to 1720 ppm. This

boron value is substantially in excess of the 520 ppm required by the

uncertainty and reactivity equivalencing analyses discussed in this

evaluation and the 1266 ppm value required to maintain keff

less than or equal to 0.95 in the presence of the most adverse

mispositioned fuel assembly.

The St. Lucie Unit 2 fuel pool boron concentration will continue to

be maintained significantly in excess of 1266 ppm; the proposed license

amendment will not result in changes in the mode of operation of the

refueling water tank (RWT) or in its use for makeup to the fuel pool.

Thus, operation of the spent fuel pool following the proposed change,

combined with the existing fuel pool boron concentration Technical

Specification limit of 1720 ppm, will continue to ensure that

keff of the fuel pool will be substantially less than 0.95.

[[Page 9604]]

Even if this not-credible dilution event was to occur, no radiation

would be released; the only consequence would be a reduction of

shutdown margin in the fuel pool. The volume of unborated water

required to dilute the fuel pool to a keff of 0.95 is so

large (in excess of 358,900 gallons to dilute the fuel pool to 520 ppm

boron) that only a limited number of water sources could be considered

potential dilution sources. The likelihood that this level of water use

could remain undetected by plant personnel is extremely remote.

In meeting the acceptance criteria for fuel pool reactivity, the

proposed amendment to increase the storage capacity of the existing

fuel pool racks does not involve a significant reduction in the margin

of safety for nuclear criticality.

Calculations of the spent fuel pool heat load with an increased

fuel pool inventory were performed using ANSI/ANS-5.1-1979 methodology.

This method was demonstrated to produce conservative results through

benchmarking to actual St. Lucie Unit 2 fuel pool conditions and by

comparison of its results to those generated by a calculation using

Auxiliary Systems Branch Technical Position 9-2 methodology.

Conservative methods were also used to demonstrate fuel cladding

integrity is maintained in the absence of cooling system forced flow.

The results of these calculations demonstrate that, for the limiting

case, the existing fuel pool cooling system can maintain fuel pool

conditions within acceptable limits with the increased inventory of

discharged assemblies. Therefore, the proposed change does not result

in a significant reduction in the margin of safety with respect to

thermal-hydraulic or spent fuel cooling considerations.

The primary safety function of the spent fuel pool and the fuel

storage racks is to maintain discharged fuel assemblies in a safe

configuration for all environments and abnormal loadings, such as an

earthquake, a loss of pool cooling or a drop of a spent fuel assembly

during routine spent fuel handling. The proposed increase in spent fuel

inventory on the fuel pool and the existing storage racks have been

evaluated and show that relevant criteria for fuel rack stresses and

floor loadings have been met and that there has been no significant

reduction in the margin of safety for these criteria.

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 Project Director: Frederick J. Hebdon.

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

Point Plant Units 3 and 4, Dade County, Florida

Date of amendment request: November 22, 1996, as revised and

replaced on February 2, 1998.

Description of amendment request: The licensee proposed to change

the Technical Specifications (TS) to allow the use of a temporary fuel

oil storage system for up to 10 days in order to perform a surveillance

requirement on the Unit 3 fuel oil storage tank with Unit 3 in Modes 5,

6, or defueled.

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:

Question 1 Does the proposed license amendment involve a

significant increase in the probability or consequences of an accident

previously evaluated?

The proposed amendment will allow the installation of a temporary

fuel oil storage and transfer system for up to 10 days, once every 10

years. EDGs [emergency diesel generators] are designed as backup AC

power sources for essential safety systems in the event of a loss of

offsite power. Since the EDGs are not accident initiators, the

probability of occurrence of accidents previously analyzed has not been

increased.

The temporary fuel oil storage tanks will be located greater than

fifty (50) feet from safety related or safe shutdown components or

circuits. This does not produce any threat to fire protection or safe

shutdown capability and therefore represents a configuration that is

bounded by existing fire hazards analysis.

The proposed amendment will not change the condition or minimum

amount of operating equipment assumed in the plant safety analyses for

accident mitigation. The temporary fuel storage and transfer system

provides a reliable means of performing the required delivery support

function for the Unit 3 EDGs.

An insignificant increase in the consequences of an accident

previously evaluated is possible since the temporary storage and

transfer system will not meet requirements for Seismic Category I or

Class 1E. However, the probability of a seismic event will be very low

due to the limited time that the temporary storage system will be in

use.

The increase in the consequences of an accident previously

evaluated is insignificant due to the following:

Manual actions required to provide a 7 day supply of fuel to the

EDGs can easily be accomplished in the 17 hours of EDG operation

provided by the 3880 gallon capacity of a single EDG day and skid tank.

The location of the temporary fuel oil supply inside the protected area

security fence by the Central Receiving Facility provides multiple

access routes to transfer fuel to the Unit 3 EDGs and is in close

proximity to a severe weather shelter for the mobile tanker.

Additionally, more than 17 hours will be available to manually

transfer fuel from the temporary fuel storage tanks located inside the

protected area, by filling the Unit 4 EDG storage tanks with

approximately 8600 gallons of fuel oil above that required for Unit 4

EDG operability. This extra capacity will be available to the Unit 3

EDGs prior to taking the permanent Unit 3 storage tank out of service.

This will be done by filling the Unit 4 fuel tanks to 39,000 gallons,

which is just below the high level alarm. This gives a capacity of 4300

gallons in each tank above the Unit 4 Technical Specification minimum

required volume of 34,700 gallons. The Unit 4 tanks are contained

within a Seismic Class 1 structure and protected by installed fire

protection equipment.

Combining the excess available fuel from the Unit 4 storage tanks

and the nominal volume of the Unit 3 day and skid tanks gives a total

of 12,480 gallons (4300 x 2+3880) of available fuel to either of the

Unit 3 EDGs. This allows a run time for a Unit 3 EDG of 55 hours

(assuming fuel oil transfer from Unit 4) prior to reaching the

Technical Specification minimum volume for the Unit 4 fuel oil storage

tanks. Manual actions to replenish the Unit 4 or Unit 3 fuel oil

storage tanks from the temporary storage tanks, via the mobile tanker,

can easily be accomplished within the 55 hours. Procedures currently

exist for the transfer of fuel from (1) the mobile tanker to the

auxiliary fill station at the Unit 3 EDGs, and (2) from the Unit 4 EDG

storage tanks to the Unit 3 day tanks by using either of the Unit 4

transfer pumps. The

[[Page 9605]]

Unit 4 transfer pumps are powered from redundant Class 1E power

supplies.

The temporary storage tanks will be located inside the protected

area in the vicinity of the Nuclear Plant Central Receiving Facility.

The temporary tanks will be located greater than fifty (50) feet from

safety related or safe shutdown components or circuits. This does not

produce any threat to fire protection or safe shutdown capability and

therefore represents a configuration that is bounded by existing fire

hazards analysis.

A dedicated mobile tanker staged inside the protected area to

transfer fuel from the temporary storage tanks to the permanent day/

skid tank system. The mobile tanker will have an integral transfer pump

to facilitate movement of fuel to either of the two truck fills at the

Unit 4 EDG building or day tank truck fills (auxiliary fill station) at

the Unit 3 EDGs. One truck fill at the Unit 4 EDG building supplies

fuel to the 4A and 4B storage tanks, the other truck fill at the Unit 4

EDG building can provide fuel directly to the Unit 3 day tanks. This

fuel supply will provide continued operation for 7 days. The temporary

storage and transfer system will not meet requirements for Seismic

Category I or Class 1E.

The capability to operate an Unit 3 EDG for 7 days during the tank

cleaning evolution will be assured by an approved plant procedure that

controls the following:

A minimum fuel supply of 3880 gallons from the Unit 3 day and

skid tank. This provides 17 hours of operation.

The extra fuel supply of 8600 gallons in the Unit 4 EDG tanks

which will be transferred by using one of the installed Unit 4 transfer

pumps. This provides an additional 38 hours of operation.

Three temporary tanks containing a minimum fuel supply of

38,000 gallons. This fuel supply will provide continued operation for 7

days.

Consequently, 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.

Question 2 Does the proposed license amendment 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 modes

of plant operation defined in the Turkey Point Units 3 and 4 operating

license. The change will not involve addition or modification of

equipment for Unit 3 EDG fuel storage and transfer. The temporary fuel

supply system provides a reliable means of performing the required fuel

delivery support function for the Unit 3 EDGs.

Consequently, operation of either unit in accordance with the

proposed amendment would not create the possibility of a new or

different kind of accident from any accident previously evaluated.

Question 3 Does the proposed amendment involve a significant

reduction in the margin of safety?

The proposed amendment is designed to provide flexibility to

schedule and perform required surveillance activities. Surveillance

intervals or operating requirements are not changed by the proposal;

only the method of fuel oil storage on a temporary basis for a single

operable EDG is addressed. The proposed change will not alter the basis

for any Technical Specification that is related to the establishment

of, or maintenance of, a nuclear safety margin.

Consequently, operation of Turkey Point Units 3 and 4 in accordance

with this 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: 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 Project Director: Frederick J. Hebdon.

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

Point Plant Units 3 and 4, Dade County, Florida

Date of amendment request: January 9, 1998.

Description of amendment request: The licensee proposed to change

the Technical Specifications (TS) to allow the use of

ZIRLOtm fuel rod clad material.

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:

Question 1 Does the proposed license amendment involve a

significant increase in the probability or consequences of an accident

previously evaluated?

Implementation of ZIRLOtm fuel rod cladding will have no

impact on the probability or consequences of any Design Basis Event

occurrences which were previously evaluated. The determination that

fuel design limits are met will continue to be performed using NRC

approved fuel performance analysis methodology. Changing to

ZIRLOtm fuel rod cladding poses no significant increase in

the probability or consequences of any accident previously evaluated.

No new performance requirements are being imposed on any system or

component in order to support implementation of ZIRLOtm fuel

rod cladding. Since the LOCA and Non-LOCA analysis results will remain

within design limits, the inputs to the radiation dose analysis do not

change. Therefore, the consequences to the public resulting from any

accident previously evaluated in the Updated Final Safety Analysis

Report (UFSAR) is not increased.

Fuel rod design criteria will be evaluated every cycle to ensure

proper compliance with fuel rod design limits and therefore the UFSAR.

The evaluation of the fuel design against fuel design limits will be

performed in accordance with 10 CFR 50.59, which ensures that the

reload will not involve an increase in the probability or consequence

of an accident previously evaluated.

Question 2 Does the proposed license amendment create the

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

previously evaluated?

Implementation of ZIRLOtm fuel rod cladding will have no

impact, nor does it contribute in any way to the probability or

consequences of an accident.

No new accident scenarios, failure mechanisms or limiting single

failures are introduced as a result of using ZIRLOtm fuel

rod cladding. The institution of ZIRLOtm fuel rod cladding

will have no adverse effect on, and does not challenge the performance

of, any safety related system.

The determination that the fuel rod design limits are met will be

performed using NRC approved methodology. Therefore, the proposed

amendment does not in any way create the possibility of a new or

different kind of accident from any accident previously evaluated.

[[Page 9606]]

Question 3 Does the proposed amendment involve a significant

reduction in the margin of safety?

The margin of safety is not affected by the implementation of

ZIRLOtm fuel rod cladding. Use of ZIRLOtm fuel

rod cladding has been approved by the NRC and does not constitute a

significant reduction in the margin of safety.

The margin of safety provided in the fuel design limits is

acceptable and will be maintained and not reduced.

In addition, each future reload will involve a 10 CFR 50.59 review

to assure that operation of the units within the cycle specific limits

will not involve a reduction in the margin of safety. Therefore, the

proposed amendment does not significantly reduce the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: 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 Project Director: Frederick J. Hebdon.

Northeast Nuclear Energy Company (NNECO), et al., Docket No. 50-423,

Millstone Nuclear Power Station, Unit No. 3, New London County,

Connecticut

Date of amendment request: January 22, 1998.

Description of amendment request: The amendment would incorporate

the proposed revision into Chapter 9 of the Millstone Unit 3 Final

Safety Analysis Report. The proposed revision to the Millstone Unit 3

licensing basis would accept the existing use of epoxy coatings on

safety-related components.

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:

NNECO has reviewed the proposed revision in accordance with

10CFR50.92 and has concluded that the revision does not involve a

significant hazards consideration (SHC). The basis for this conclusion

is that the three criteria of 10CFR50.92(c) are not satisfied. The

proposed revision does not involve [an] SHC because the revision would

not:

1. Involve a significant increase in the probability or consequence

of an accident previously evaluated.

Past experience indicates that failure of previous ARCOR

applications may have degraded the performance of SWS [service water

system] heat exchangers within one train, but there is no indication

that failure of multiple heat exchangers on both trains is feasible.

Furthermore, the likelihood of ARCOR material being released has been

reduced by improving the application procedure and performing

destructive testing to detect disbondment. In addition, the completion

of normal heat exchanger performance surveillance's and periodic visual

inspections minimizes the potential for disbonded ARCOR to degrade SWS

components.

Therefore, the presence of ARCOR coating material within the SWS

does not involve a significant increase in the probability or

consequence of an accident previously evaluated.

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

from any accident previously evaluated.

The application of ARCOR material may lead to the degradation of

SWS heat exchangers. However, multiple ARCOR application failures

occurring simultaneously either instantaneously or gradually resulting

in failure of all SWS heat exchangers in both trains is not considered

feasible. An instantaneous failure is discounted by analysis which

concludes that normal system operations are more likely to cause the

release of degraded ARCOR than what might be expected following a

seismic event. Gradual degradation is not expected since normal SWS

heat exchanger performance surveillance's will identify heat exchanger

tubesheet fouling and thus, provide early detection of coating

failures. Therefore, the use of ARCOR coating material within the SWS

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

from any accident previously evaluated.

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

Although the gradual release of ARCOR material creates the

potential to simultaneously degrade the performance of mitigating

equipment in both trains of safety systems, it is determined to be

unrealistic due to normal heat exchanger performance surveillance's.

These surveillance's are expected to identify heat exchanger tubesheet

fouling and provide early detection and mitigation of a problem with

the pipe coatings. Therefore, the application of ARCOR coating within

the SWS does not involve a significant reduction in the margin of

safety.

In conclusion, based on the information provided, it is determined

that the proposed revision does not involve an SHC.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Learning Resources Center,

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

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

49 Rope Ferry Road, Waterford, Connecticut.

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

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

Connecticut.

NRC Deputy Director: Phillip F. McKee.

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

Generating Plant, Wright County, Minnesota

Date of amendment request: July 26, 1996, as supplemented September

5 and December 4, 1997.

Description of amendment request: The proposed amendment would, as

part of the licensee's power rerate program, increase the maximum power

level to 1775 megawatts thermal (MWt). This change is approximately 6.3

percent above the current maximum power level of 1670 MWt.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

A. The proposed amendment will not involve a significant increase

In the probability or consequences of an accident previously evaluated.

The probability of occurrence and consequences of an [accident]

previously evaluated have been evaluated for MNGP [Monticello Nuclear

Generating Plant] Power Rerate. This evaluation has concluded that MNGP

Power Rerate will not involve a significant increase in the probability

of occurrence or consequences of previously evaluated accidents.

1. Evaluation of Accident Consequences

(a) ECCS-LOCA Analysis

The Emergency Core Cooling System Loss of Coolant Accident (ECCS-

LOCA)

[[Page 9607]]

performance analysis has been evaluated for MNGP Power Rerate using

methodology which has been approved by the NRC for LOCA 10CFR50.46

analyses [requirements]. The current ECCS performance requirements were

used in the power rerate analysis; no further parameter relaxations

were included in the analysis. The ECCS-LOCA analysis was performed for

MNGP Power Rerate for the existing licensed rated thermal power and at

a bounding thermal power level of 1880 MWt that is approximately 6%

greater than the proposed power rerate to 1775 MWt [megawatts thermal].

In addition, the bounding thermal power level was increased by an

additional 2% in accordance with regulatory guidance. The licensing

peak clad temperature for the bounding analyzed thermal power level

remains below the 10CFR50.46 required limit of 2,200'F. Therefore the

analysis demonstrates that MNGP will continue to comply with 10CFR50.46

and 10CFR50, Appendix K at rerated conditions thus the consequences of

a LOCA is not significantly increased for the proposed power rerate.

(b) Abnormal Operating Transient Analysis

An evaluation of the Updated Safety Analysis Report (USAR) and

reload transients has been performed for MNGP Power Rerate to

demonstrate that the proposed power rerate has no adverse effect on

plant safety. This evaluation was performed for a power level of 1775

MWt, with the exception that certain event evaluations were performed

at 102% of the rerate power level. The transient analysis performed to

demonstrate the acceptability of MNGP Power Rerate used the NRC

approved methods identified in the MNGP Technical Specifications.

The limiting transient events at the power rerate conditions have

been analyzed. This includes all events that establish the core thermal

operating limits and the events that bound other transient acceptance

criteria. These limiting transients were benchmarked against the

existing rated thermal power level by performance of the event analysis

at both the proposed rerate power level and the existing rated power

level. In addition, an expanded group of transient events was evaluated

to confirm that these events were less severe with the power rerate

than the most limiting transients. The events included in the expanded

group of transient events were chosen based on those events which have

been demonstrated to be sensitive to initial power level. This

evaluation confirmed that the existing set of limiting transient events

remains valid for MNGP Power Rerate. The evaluation was performed for a

representative core and demonstrated the overall capability to meet all

transient safety criteria for the power rerate. Cycle specific analysis

will continue to be performed for each fuel reload to demonstrate

compliance with the applicable transient criteria and to establish

cycle specific operating limits.

The results of the evaluation of transients demonstrate that the

power rerate can be accomplished without a significant increase in the

consequences of the transients evaluated. The fuel thermal-mechanical

limits at the power rerate conditions are within the specific design

criteria for the GE [General Electric] fuels currently loaded in the

MNGP core. Also, the power-dependent and flow-dependent MCPR [minimum

critical power ratio] and Maximum Average Planar Linear Heat Generation

Rate (MAPLHGR) methods developed as part of the core performance

improvement program remain applicable to rerate conditions. The

transient event evaluation confirmed that MNGP Power Rerate has no

significant effect on the power-dependent and flow-dependent MCPR and

MAPLHGR limits. The peak reactor pressure vessel bottom head pressure

remains within the ASME [American Society of Mechanical Engineers]

requirement for reactor pressure vessel overpressure protection.

The effects of plant transients were evaluated by assessing a

number of disturbances of process variables and malfunctions or

failures of equipment consistent with USAR. The transient events were

evaluated against the Safety Limit Minimum Critical Power Ratio,

(SLMCPR). The SLMCPR is determined using NRC-approved methods. The

limiting transient events are slightly more severe when initiated from

the rerate power level. The power rerate transient evaluation results

show a slightly more limiting event initial CPR [critical power ratio]

(less than or equal to 0.02) than that initiated from the present rated

power level for the near limiting transients. However, for the most

limiting transient, the evaluation of a representative core showed that

no change is required to the Operating Limit MCPR for the power rerate

and that the integrity of the SLMCPR is maintained. The margin of

safety established by the SLMCPR is not affected and the event

consequences are not significantly affected by the proposed power

rerate to 1775 MWt. Cycle specific analysis will continue to be

performed for each fuel reload to demonstrate compliance with the

applicable transient criteria and to establish cycle specific operating

limits.

The results demonstrate that the MNGP core thermal power output can

be safely increased to the power rerate level without significant

effect on the consequences of previously evaluated postulated transient

events. The results of the rerate transient analysis are summarized as

follows.

(1) Events Resulting in a Nuclear System Pressure Increase

(a) Main Generator Load Rejection with No Steam Bypass

At rerated conditions, the fuel transient thermal and mechanical

overpower results remain below the NRC accepted design criteria.

(b) Main Turbine Trip with No Steam Bypass

At rerate conditions, the fuel transient thermal and mechanical

overpower results remain below the NRC accepted design criteria.

(c) Main Steam Isolation Valve Closure, Flux Scram

The peak reactor pressure vessel bottom head pressure for rerate

conditions is slightly higher than the reactor pressure vessel bottom

head pressure at current conditions. However, the resultant pressure is

still below the ASME overpressure limit of 1,375 psig [pounds per

square inch].

(d) Slow Closure of a Single Turbine Control Valve

The results of this transient for the power rerate remain non-

limiting as compared with other more severe pressurization events.

(2) Event Resulting in a Reactor Vessel Water Temperature Decrease

(a) Feedwater Controller Failure-Maximum Demand

The delta CPR calculated for this event at rerate conditions is

about 0.01 higher than the corresponding value for the current rated

power when the impact of the new condensate pumps is factored in. The

trend for the Feedwater Controller Failure-Maximum Demand event is

consistent with the analysis for the current rated power. The fuel

thermal margin results are within the acceptable limits for the fuel

types analyzed.

(b) Loss of Feedwater Heating

This event at the rerate conditions remains significantly less than

the cycle operating MCPR limit. The results at low core flow conditions

are actually slightly higher than for the high core

[[Page 9608]]

flow condition because of increased inlet coolant subcooling into the

reactor core. The calculated thermal and mechanical overpower limits at

the power rerate conditions for this event also meet the fuel design

criteria.

(c) Inadvertent HPCI [high-pressure coolant injection] Actuation

For the limiting condition analyzed, both the high water level

setpoint and the high reactor pressure vessel steam dome pressure scram

setpoints are not reached. Based on the peak average fuel surface heat

flux results, the HPCI actuation event will be bounded by the limiting

pressurization event with respect to delta Critical Power Ratio

([delta] CPR) considerations. In addition, the fuel transient thermal

and mechanical overpower limits remain within the NRC accepted design

values.

(3) Event Resulting in a Positive Reactivity Insertion

(a) Rod Withdrawal Error (RWE)

The current Rod Block Monitor (RBM) system for MNGP with power

dependent setpoints was analyzed for the rod withdrawal error event at

the power rerate conditions using a statistical approach consistent

with NRC approved methods. The analysis concluded that the transient is

slightly more severe with a greater delta Critical Power Ratio ([delta]

CPR) from the initial most limiting CPR. However, the fuel and

mechanical overpower results remain within the NRC accepted design

criteria.

(4) Event Resulting in a Reactor Vessel Coolant Inventory Decrease

(a) Pressure Regulator Failure to Full Open

The results of this transient for the power rerate remain non-

limiting as compared with other more severe pressurization events.

(b) Loss of Feedwater Flow

This transient event does not pose any direct threat to the fuel in

terms of a power increase from the initial conditions. Water level

declines rapidly and a low level causes a reactor scram. The closure of

the main steam isolation valves and the actuation of High Pressure

Coolant Injection and Reactor Core Isolation Cooling terminate the

event. This event was included in the power rerate evaluation to

provide assurance that sufficient water makeup capability is available

to keep the core covered when all normal feedwater is lost. The generic

analysis performed in support of the extended power uprate program

shows that at the power rerate conditions a large amount of water

remains above the top of the active fuel. These sequences of events do

not require any new operator actions or shorter operator response

times. Therefore, the operator actions for the event do not

significantly change for the power rerate.

(5) Event Resulting in a Core Coolant Flow Decrease

(a) Recirculation Pump Seizure

The recirculation pump seizure assumes instantaneous stoppage of

the pump motor shaft of one recirculation pump. As a result, the core

flow decreases rapidly. The heat flux decline lags core power and flow

and could result in a degradation of core heat transfer. At the power

rerate conditions, the transient results confirmed that the

consequences of the pump seizure event remain non-limiting.

(6) Event Resulting in a Core Coolant Flow Increase

(a) Recirculation Flow Controller Failure Increasing Flow

The results of this transient for the power rerate remain non-

limiting as compared with other more severe pressurization events.

(c) Design Basis Accident Challenges to the Containment

The primary containment response to the limiting design basis

accident was evaluated for a bounding reactor power level approximately

6% greater than the proposed power rerate to 1775 MWt. In addition, the

bounding reactor power level was increased by an additional 2% in

accordance with regulatory guidance. The effect of the power rerate on

the short term containment response (peak values) as well as the long

term containment response for containment pressure and temperature

confirms the suitability of the plant for operation at the bounding

power level, thus the proposed power rerate to 1775 MWt is acceptable.

Factors of safety provided in the ASME Code are maintained and safety

margin is not affected for the power rerate to 1775 MWt.

Short-term containment response analyses were performed for the

limiting design basis LOCA consisting of a double-ended guillotine

break of a recirculation suction line, to demonstrate that operation at

a bounding reactor power will not result in exceeding the containment

design limits. This limiting design basis LOCA event results in the

highest short-term containment pressures and dynamic loads. The

analysis determined that for a bounding reactor power the maximum

drywell pressure values are bounded by the current USAR analysis value

and by the containment design pressure. The power rerate to 1775 MWt

has no adverse effect on the containment structural design pressure.

Because there will be more residual heat with increased thermal

power, the containment long term response will have slightly higher

temperatures. Long term suppression chamber temperatures remain within

the design temperature of the structure, thus factors of safety

provided in the ASME code are maintained and safety margin is not

affected. Analysis confirmed that ECCS pump NPSH is adequate for this

temperature response. It was confirmed that the long term response does

not adversely affect the containment structure or the environmental

qualification (EQ) of equipment located in the drywell or suppression

chamber room. The drywell long term temperature response is not

adversely affected for a bounding reactor power. An analytical power

level of 1880 MWt bounds the decay heat associated with the 1775 MWt

power level with a one sided confidence interval of 95%. The

containment long term response is therefore acceptable for the power

rerate to 1775 MWt.

The impact of a reactor power increase on the containment dynamic

loads have been determined, evaluated and found to have no adverse

effects for conditions which well bound the proposed power rerate. Thus

the containment dynamic loads were found to be acceptable for the power

rerate to 1775 MWt.

The MNGP Power Rerate evaluation of the primary containment

response to the design basis accident confirmed that the power rerate

does not result in a significant increase in consequences for a

bounding reactor power approximately 6% greater than the proposed power

rerate to 1775 MWt.

(d) Radiological Consequences of Design Basis Accidents

For MNGP Power Rerate, the radiological consequences of the

limiting design basis accidents were re-evaluated. These evaluations

included the effect of the power rerate on the radiological

consequences of accidents presented in USAR Section 14.7.

This evaluation was performed using inputs and evaluation

techniques consistent with the current regulatory guidance, the current

GE analysis methods, and the appropriate plant design basis. The inputs

and analysis methods used for MNGP Power Rerate differ from those

utilized in the current licensing basis evaluation presented in

[[Page 9609]]

the USAR and the AEC [Atomic Energy Commission] safety evaluation

supporting plant initial licensing. The MNGP Power Rerate evaluations

used the more contemporary staff approved methods. The inputs used in

the MNGP Power Rerate evaluation provide a conservative assessment of

the potential radiological consequences. The conclusions of these

evaluations are consistent with the original licensing basis

evaluations. The radiological consequences of the limiting design basis

accidents remain well within 10CFR100 guidelines for a bounding thermal

power approximately 6% greater than the proposed power rerate of 1775

MWt. In addition the bounding thermal power level was increased by an

additional 2% in accordance with regulatory guidance.

To conservatively analyze the change in consequences, the

evaluation of radiological consequences using the analysis inputs and

methods was performed for the existing licensed rated thermal power and

a thermal power bounding the proposed power rerate. This provides a

conservative bounding change in consequences for the requested power

rerate to 1775 MWt.

The MNGP Power Rerate evaluation of the radiological consequences

of design basis accidents confirmed that the power rerate does not

result in a significant increase in consequences for a bounding power

level approximately 6% greater than the proposed power rerate. The

results remain below the 10CFR100 guideline values as well as the

licensing basis established in the March 18, 1970 AEC safety

evaluation. Therefore, the postulated radiological consequences do not

represent a significant change in accident consequences and are clearly

within the regulatory guidelines for the proposed power rerate to 1775

MWt.

(e) Other Evaluations

(1) Performance Improvements

The MNGP Power Rerate safety analysis has been performed taking

into account the implementation of the following previously approved

special operational features.

(a) Maximum Extended Load Line Limit/Increase Core Flow (MELLL/ICF)

The safety analysis for rerate conditions shows that the extended

operating domain as analyzed by MELLL/ICF remains valid for the power

rerate conditions.

(b) Average Power Range Monitor/Rod Block Monitor Technical

Specification (ARTS) Improvements

The safety analysis for rerate conditions shows that the ARTS

improvements remain valid for the power rerate conditions.

(c) Single Loop Operation (SLO)

The safety analysis for rerate conditions shows that the single

loop operating mode remains valid for the power rerate conditions. The

MELLLA trip setpoints determined for two-loop operation were confirmed

to be acceptable for single loop operation with a correction applied to

account for the actual effective drive flow applied when operating in

single loop. The single loop settings have been conservatively

established to be consistent with the two loop settings while ensuring

the appropriate corrections are applied to the MAPLHGR and the

operating limit MCPR to account for single loop operation.

(2) Effect of Power Rerate on Support Systems

An evaluation was performed to address the effect of MNGP Power

Rerate on accident mitigation features, structures, systems, and

components within the balance of plant. The results are as follows:

Auxiliary systems such as, building heating, Ventilation and Air

Conditioning (HVAC) systems, reactor building closed cooling water,

service water and emergency service water, spent fuel pool cooling,

process auxiliaries such as instrument air and makeup water and the

post-accident sampling system were confirmed to operate acceptably

under normal and accident conditions at rerate conditions.

The secondary containment and standby gas treatment system were

confirmed to be able to adequately contain, process, and control the

release of normal and post-accident levels of radioactivity at rerate

conditions.

Instrumentation was reviewed and confirmed to be capable of

performing its control and monitoring functions under rerate

conditions. As required, analyses were performed to determine the need

for setpoint changes for various functions (e.g., APRM [average power

range monitor] neutron flux scram setpoints). In general, setpoints are

to be changed only to maintain adequate difference between plant

operating parameters and trip setpoints, while ensuring safety

performance is demonstrated. The revised setpoints have been

established using the NRC reviewed methodology as guidance.

Electric power systems including the turbine generator and

switchgear components were verified as being capable of providing the

electrical load as a result of the rerate power levels. An evaluation

of the auxiliary power system for the power rerate conditions confirmed

that the system has sufficient capacity with the changes identified in

Exhibit I [of the 12/4/97 submittal] to support all required loads for

safe shutdown, to maintain a safe shutdown condition, and to operate

the required engineered safeguards equipment following postulated

accidents. No safety-related electrical loads were affected which would

adversely impact the emergency diesel generators.

Piping systems were evaluated for the effect of operation at higher

power levels, including transient loading. The evaluation confirmed

that, with few exceptions, piping and supports are adequate to

accommodate the increased loading resulting from operation at rerate

power conditions. In a few cases, piping supports will be modified to

accept higher forces due to rerate conditions.

The effect of rerate conditions on high energy line break (HELB)

was evaluated. The evaluation confirmed structures, systems, and

components important to safety are capable of accommodating the effects

of jet impingement and blowdown forces and the environmental effects

resulting from HELB events at rerate conditions.

Control room habitability was evaluated. With the implementation of

minor hardware and non-hardware changes to the control room ventilation

system, Post-accident Control Room and Technical Support Center doses

at rerate conditions were confirmed to be within the guidelines of

General Design Criterion 19 of 10CFR50, Appendix A.

The environmental qualification of equipment important to safety

was evaluated for the effect on normal and accident operating

conditions at rerate power levels. The equipment remains qualified for

the new conditions. Minor adjustments will reflect some changes to

maintenance frequencies. The preventative maintenance program will

continue to provide for equipment maintenance or replacement to ensure

equipment environmental qualification at rerate power conditions.

(3) Effect on Special Events

The consequences of special events (i.e., ATWS [anticipated

transient without scram], 10CFR50, Appendix R, and Station Blackout)

remain within NRC accepted criteria for rerate conditions. Concurrent

malfunctions assumed to occur during accidents have

[[Page 9610]]

been accounted for in the safety analyses for rerate conditions. The

consequences of these equipment malfunctions does not change with

implementation of the MNGP Power Rerate program. The generic ATWS

analysis for operation at rerate conditions is being revised. The

revision is not expected to affect MNGP compliance with NRC acceptance

criteria.

(f) Conclusion

The evaluation of the Emergency Core Cooling System performance has

demonstrated the criteria of 10CFR50.46 are satisfied, thus the margin

of safety established by the criteria is maintained. The analysis

demonstrated that the ECCS will function with the most limiting single

failure to mitigate the consequences of the accidents and maintain fuel

integrity. The system will continue to perform as required under rerate

conditions to mitigate the consequences of accidents and thus the power

rerate does not adversely affect ECCS performance in a manner to

increase the severity of consequences. Challenges to the containment

have been evaluated and the integrity of the fission product barrier

has been confirmed. The radiological consequences of design basis

accidents have been evaluated and it was found that the effect of the

proposed power rerate on postulated radiological consequences does not

result in a significant increase in accident consequences. These

evaluations have been performed for a bounding reactor power

approximately 6% greater than the proposed power rerate. In addition

the bounding reactor power level was increased by an additional 2% in

accordance with regulatory guidance. Thus the evaluations provide

conservative bounding results for the proposed power rerate to 1775 MWt

and demonstrate that the proposed power rerate does not result in

significant increase in accident consequences.

The abnormal transients have been analyzed under the power rerate

conditions, and the analysis has confirmed that the power rerate to

1775 MWt has only a minor effect on the minimum critical power ratio

and that no change to the safety limit critical power ratio results,

thus the margin of safety as assured by the safety limit critical power

ratio is maintained. The effect of the power rerate on the consequences

of abnormal transients which result from potential component

malfunctions has been shown to be acceptable, thus the power rerate

does not result in a significant increase in transient event

consequences.

The spectrum of analyzed postulated accidents and transients has

been investigated, and has been determined to meet the current

regulatory criteria for the MNGP at rerate conditions. In the area of

core design, the fuel operating limits will still be met at the rerate

power level, and fuel reload analyses will show plant transients meet

the criteria accepted by the NRC as specified in the plant Technical

Specifications. The evaluation of transient and accident consequences

was performed consistent with the proposed changes to the plant

Technical Specifications. Therefore, the proposed Operating License and

Technical Specification changes will not cause a significant increase

in the consequences of an accident previously evaluated for the

Monticello plant.

2. Evaluation of the Probability of Previously Evaluated Accidents

The proposed power rerate imposes only minor increases in the plant

operating conditions. No changes are required to the rated core flow,

rated reactor pressure, or turbine throttle pressure. The power rerate

will result in moderate flow increases in those system[s] associated

with the turbine cycle (i.e., condensate, feedwater, main steam, etc.).

For MNGP Power Rerate, the small increase in operating temperatures for

balance of plant support systems has no significant effect on LOCA or

other accident probabilities.

The increase in flow rates in balance of plant systems is

addressed by compliance with NRC Generic Letter 89-08, ``Erosion/

Corrosion in Piping.'' The MNGP Power Rerate evaluations have confirmed

that the power rerate has no significant effect on flow induced

erosion/corrosion. The worst case limiting feedwater and main steam

piping flow increases were evaluated to be approximately proportional

to the power increase. The affected systems are currently monitored by

the MNGP Erosion/Corrosion program. Continued monitoring of the systems

provides a high level of confidence in the integrity of potentially

susceptible high energy piping systems.

The occurrence frequency of accident precursors and transients

[has] been addressed when required by applying the guidance of NRC

reviewed setpoint methodology to insure that acceptable trip avoidance

is provided during operational transients subsequent to implementation

of rerate. The setpoint evaluation has confirmed that MNGP Power Rerate

does not result in any increase in challenges to the plant protective

instrumentation.

Plant systems, components, and structures have been verified to be

capable of performing their intended functions under rerate conditions

with a few minor exceptions. Where necessary, some components will be

modified prior to implementation of the MNGP Power Rerate Program to

accommodate the revised operating conditions (e.g., a limited number of

pipe supports changes, instrumentation setpoint changes, control room

habitability improvements). MNGP Power Rerate does not significantly

affect the reliability of plant equipment. Where reliability effects

have been identified, modifications and administrative controls will be

implemented prior to the power rerate to adequately compensate. No new

components or system interactions that could lead to an increase in

accident probability are created due to the power rerate.

The probability (i.e., frequency of occurrence) of design basis

accidents occurring is not affected by the increased power level, as

the applicable criteria established for plant equipment (e.g., ANSI

Standard B31.1, ASME Code,) will still be followed as the plant is

operated at the rerate power level. The MNGP Power Rerate analysis

basis assures that the power dependent margin prescribed by the Code of

Federal Regulations (CFR) will be maintained by meeting the appropriate

regulatory criteria. Similarly, factors of safety specified by

application of the Code design rules have been demonstrated to be

maintained, as have other margin-assuring acceptance criteria used to

judge the acceptability of the plant. Reactor scram setpoints as

established are such that there is no significant increase in scram

frequency due to rerate conditions. No new challenges to safety-related

equipment will result from the power rerate. Therefore, the proposed

Operating License and Technical Specifications changes do not involve a

significant increase in the probability of an accident previously

evaluated.

B. The proposed Operating License changes do not create the

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

previously evaluated.

The basic Boiling Water Reactor configuration, operation and event

response is unchanged by the power rerate. Analysis of transient events

has confirmed that the same transients remain limiting and that no

transient events result in a new sequence of events which could lead to

a new accident scenario. The MNGP Power Rerate analyses confirmed that

the accident progression is basically unchanged by the power rerate.

[[Page 9611]]

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 same fission product barriers such as the fuel cladding, the

reactor coolant pressure boundary and the reactor containment, remain

in place. Fuel rod cladding integrity is ensured by operating within

thermal, mechanical, and exposure design limits and is demonstrated by

the MNGP Power Rerate transient analysis and accident analysis.

Similarly, analysis of the reactor coolant pressure boundary and

primary containment have demonstrated that the power rerate has no

adverse effect on these fission product barriers. The proposed changes

to the plant Technical Specifications to support the power rerate

implementation are consistent with the MNGP Power Rerate analyses and

assure transient and accident mitigation capability in compliance with

regulatory requirements.

The effect of MNGP Power Rerate on plant equipment has been

evaluated. No new operating mode, safety-related equipment lineup,

accident scenario, or equipment failure mode resulting from the power

rerate was identified. The full spectrum of accident considerations

defined in the USAR have been evaluated and no new or different kind of

accident resulting from the power rerate has been identified. MNGP

Power Rerate uses already developed technology and applies it within

the capabilities of already existing plant equipment in accordance with

presently existing regulatory criteria which includes accepted codes,

standards, and methods. GE has designed BWRs of higher power levels

than the rerate power of any of the currently operating BWR fleet and

no new power dependent accidents have been identified. In addition,

MNGP Power Rerate does not create any new sequence of events or failure

modes that lead to a new type of accident.

All actions to ensure that safety-related structures, systems, and

components will remain within their design allowable values and ensure

they can perform their intended functions under rerate conditions will

be taken prior to implementation of the power rerate. MNGP Power Rerate

does not increase challenges to or create any new challenge to safety-

related equipment or other equipment whose failure could cause an

accident. Plant modifications required to support implementation of

MNGP Power Rerate will be made to existing systems (e.g., a limited

number of pipe supports, instrumentation setpoints, control room

habitability improvements), rather than by adding new systems of a

different design which might introduce new failure modes or accident

sequences. The Technical Specification changes required to implement

the power rerate require little change to the plant's configuration,

and all changes have been evaluated and are acceptable.

Therefore, the proposed Operating License and Technical

Specification changes do not create the possibility of a new or

different kind of accident from any previously evaluated.

C. The proposed Operating License changes do not involve a

significant reduction in a margin of safety.

The accident analysis, as well as a majority of the plant specific

evaluations performed in support of MNGP Power Rerate have been

performed assuming a bounding steady state power level 112.6% of the

existing licensed limit of 1670 MWt, and approximately 6% above the

licensed maximum thermal power level of 1775 MWt proposed by MNGP Power

Rerate. In addition, the bounding reactor power level was increased by

an additional 2% in accordance with regulatory guidance when applicable

for the evaluation of accidents and transients. For plant conditions

associated with a bounding analysis power level, the analyses

demonstrated operating margin to criteria establishing margins of

safety, thus additional operating margin is demonstrated and assured

for the proposed power rerate to 1775 MWt and added confidence is

established in the integrity of criteria establishing margin to safety.

The cycle specific transient analysis, as well as the analysis to

establish plant instrumentation set points have been performed assuming

a plant steady state power level of 1775 MWt. This analysis approach

was taken in order to demonstrate safety and equipment margins while

ensuring appropriate cycle specific operating limits. The evaluation of

transient events and instrument setpoints demonstrated operating margin

to criteria establishing margins of safety for the proposed power

rerate conditions.

The MNGP Power Rerate analysis basis assures that the power

dependent safety margin assuring criteria prescribed by the Code of

Federal Regulations (CFR) will be maintained by meeting the appropriate

regulatory criteria. Similarly, factors of safety specified by

application of the code design rules have been maintained, as have

other margin-assuring acceptance criteria used to judge the

acceptability of the plant.

1. Fuel Thermal Limits

No change is required in the basic fuel design to achieve the

rerate power levels or to maintain the margins as discussed above. No

increase in the allowable peak bundle power is requested for the power

rerate. The abnormal transients have been evaluated under the power

rerate conditions for a representative core configuration. The analysis

has confirmed that the power rerate has no adverse effect on the

operating limit Minimum Critical Power Ratio (MCPR) and that 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 such

as Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) and the

operating limit MCPR will still be met at the rerate power level. The

MNGP Power Rerate analyses have confirmed the acceptability of these

operating limits for the power rerate without an adverse effect on

margins to safety. Cycle specific analysis will continue to be

performed for each fuel reload to demonstrate compliance with the

applicable transient criteria and to establish cycle specific operating

limits.

2. Design Basis Accidents Challenges to Fuel

The evaluation of the Emergency Core Cooling System performance has

demonstrated the criteria of 10CFR50.46 are satisfied, thus the margin

of safety established by the criteria is maintained. This evaluation

was performed for a bounding reactor power level approximately 6%

greater than the proposed power rerate. In addition the bounding

reactor power level was increased by an additional 2% in accordance

with regulatory guidance. The analysis demonstrates that MNGP will

continue to comply [with] the 10 CFR 50.46 at the rerate conditions and

that the margin of safety established by the regulation is maintained

for the proposed power rerate.

3. Design Basis Accident Challenges to Containment

The primary containment response to the limiting design basis

accident was evaluated for a bounding reactor power level approximately

6% greater than the proposed power rerate to 1775 MWt. In addition, the

bounding reactor power level was increased by an additional 2% in

accordance with regulatory guidance. The effect of the power rerate on

the short term containment response (peak values) as well as the long

term containment response for containment pressure and temperature

confirms the

[[Page 9612]]

suitability of the plant for operation at the bounding power level,

thus the proposed power rerate to 1775 MWt is acceptable. Factors of

safety provided in the ASME Code are maintained and safety margin is

not affected for the power rerate to 1775 MWt.

Short-term containment response analyses were performed for the

limiting design basis LOCA consisting of a double-ended guillotine

break of a recirculation suction line, to demonstrate that operation at

a bounding reactor power will not result in exceeding the containment

design limits. The analysis determined that for a bounding reactor

power the maximum drywell pressure values are bounded by the current

USAR analysis value and by the containment design pressure. The power

rerate to 1775 MWt has no adverse effect on the containment structural

design pressure.

Long term suppression chamber temperatures remain within the design

temperature of the structure, thus factors of safety provided in the

ASME code are maintained and safety margin is not affected. An

analytical power level of 1880 MWt bounds the decay heat associated

with the 1775 MWt power level with a one sided confidence interval of

95%. Analysis confirmed that ECCS pump NPSH is not adversely affected

with this temperature response. It was confirmed that the long term

response does not significantly affect the containment structure or the

environmental qualification (EQ) of equipment located in the drywell or

suppression chamber room.

The impact of a reactor power increase on the containment dynamic

loads [has] been determined, evaluated and found to have no adverse

effects for conditions which well bound the proposed power rerate. Thus

the containment dynamic loads were found to be acceptable for the power

rerate to 1775 MWt.

The MNGP Power Rerate evaluation of the primary containment

response to the design basis accident confirmed that the power rerate

does not result in a reduction in margins of safety for a bounding

reactor power approximately 6% greater than the proposed power rerate

to 1775 MWt.

4. Design Basis Accident Radiological Consequences

The Updated Safety Analysis Report (USAR) provides the radiological

consequences for each of the design basis accidents. The magnitude of

the potential consequences is dependent upon the quantity of fission

products released to the environment, the atmospheric dispersion

factors and the dose exposure pathways. For power rerate, the

atmospheric dispersion factors and the dose exposure pathways do not

change. Therefore, the only factor which will influence the magnitude

of the consequences is the quantity of activity released to the

environment. This quantity is a product of the activity released from

the core and the transport mechanisms between the core and the effluent

release point.

The radiological consequences of design basis accidents have been

evaluated, and it was found that the consequences did not result in a

significant increase in consequences for a bounding reactor power level

approximately 6% greater than the proposed power rerate. In addition,

the bounding reactor power level was increased by an additional 2% in

accordance with regulatory guidance. The results remain below the

10CFR100 guideline values as well as the licensing basis established in

the March 18, 1970 AEC safety evaluation. Therefore, the postulated

radiological consequences are clearly within the regulatory guidelines

and all radiological safety margins are maintained for the power rerate

to 1775 MWt.

5. Transient Evaluations

The effects of plant transients were evaluated by assessing a

number of disturbances of process variables and malfunctions or

failures of equipment consistent with USAR. The transient events were

evaluated against the Safety Limit Minimum Critical Power Ratio,

(SLMCPR). The SLMCPR is determined using NRC-approved methods. The

Power Rerate transient analyses were performed using the approved

methodology specified in the plant Technical Specifications. The

limiting transient events are slightly more severe when initiated from

the rerate power level. The power rerate transient evaluation results

show a slightly more limiting transient initial CPR (less than or equal

to 0.02) than that initiated from the present rated power level for the

near limiting transients. However, for the most limiting transient, the

evaluation of a representative core showed that no change is required

to the Operating Limit MCPR for the power rerate and that the integrity

of the SLMCPR is maintained. Cycle specific analysis will continue to

be performed for each fuel reload to demonstrate compliance with the

applicable transient criteria and to establish cycle specific operating

limits.

The fuel thermal-mechanical limits at the power rerate conditions

are within the specific design criteria for the GE fuels currently

loaded in the MNGP core. Also, the power-dependent and flow-dependent

MCPR and Maximum Average Planar Linear Heat Generation Rate (MAPLHGR)

methods developed as part of the core performance improvement program

remain applicable to rerate conditions. The transient event evaluation

confirmed that MNGP Power Rerate has no significant effect on the

power-dependent and flow-dependent MCPR and MAPLHGR limits. The peak

reactor pressure vessel bottom head pressure remains within the ASME

requirement for reactor pressure vessel over pressure protection.

The margin of safety established by the SLMCPR is not affected by

the proposed power rerate to 1775 MWt.

6. Technical Specification Changes

The Technical Specifications ensure that the plant and system

performance parameters are maintained at the values assumed in the

safety analysis. The Technical Specification (setpoints, trip settings,

etc.) are selected such that the actual equipment is maintained equal

to or conservative with respect to the inputs used in the safety

analysis. Proper account is taken of inaccuracies introduced by

instrument drift, instrument accuracy, and calibration accuracy. The

Technical Specifications address equipment availability and limit

equipment out-of-service to assure that the plant can be expected to

have at least the complement of equipment available to deal with plant

transients as that assumed in the safety analysis. The evaluations and

analyses performed to demonstrate the acceptability of MNGP Power

Rerate were performed using inputs consistent with the proposed changes

to the plant Technical Specifications.

The events that form the Technical Specification Bases were

evaluated for the power rerate conditions using inputs and initial

conditions consistent with the proposed Technical Specification

changes. Although some changes to the Technical Specifications are

required for the power rerate, no NRC acceptance limit will be

exceeded. Therefore, the margins of safety assured by safety limits and

other Technical Specification limits will be maintained. The changes to

the Technical Specification Bases proposed by this submittal are

consistent with the evaluations which demonstrated acceptability of the

power rerate.

7. Conclusion

The spectrum of postulated accidents, transients, and special

events has been investigated and [has] been determined to meet the

current regulatory criteria

[[Page 9613]]

for the MNGP at the power rerate conditions. In the area of core

design, the fuel operating limits will still be met at the rerate power

level, and fuel reload analyses will show plant transients meet the

criteria accepted by the NRC as specified in the plant Technical

Specifications. Challenges to fuel or ECCS performance were evaluated

and shown to meet the criteria of 10 CFR 50.46 and 10 CFR 50, Appendix

K. Challenges to the containment have been evaluated and the integrity

of the fission product barrier has been confirmed. Radiological release

events have been evaluated and shown to meet the guidelines of 10 CFR

100. The proposed Operating License and Technical Specification changes

are consistent with the MNGP Power Rerate evaluation performed. The

evaluations demonstrated compliance with the margin assuring acceptance

criteria contained in applicable codes and regulations. Therefore, the

proposed Operating License and Technical Specifications changes 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 10 CFR 50.92(c) are

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Minneapolis Public Library,

Technology and Science Department, 300 Nicollet Mall, Minneapolis,

Minnesota 55401

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

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

NRC Project Director: Cynthia A. Carpenter

Philadelphia Electric Company, Docket No. 50-352, Limerick Generating

Station (LGS), Unit 1, Montgomery County, Pennsylvania

Date of amendment request: February 9, 1998.

Description of amendment request: The amendment request proposes to

revise the LGS, Unit 1 Technical Specifications (TS) Section 2.1 and

its associated TS Basis to reflect the change in the minimum critical

power ratio (MCPR) safety limit due to the plant-specific evaluation

performed by General Electric Company (GE) for LGS, Unit 1, Cycle 8.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed TS change does not involve a significant increase

in the probability or consequences of an accident previously evaluated.

The revised MCPR Safety Limits for LGS Unit 1 Technical

Specifications, and their use to determine cycle-specific thermal

limits, have been calculated using NRC-approved methods (i.e., GESTAR-

II, Rev. 13) and are based on LGS Unit 1 Cycle 8 specific inputs. The

use of these methods assures that the [safety limit for minimum

critical power ratio] SLMCPR value is within the existing design and

licensing basis, and cannot increase the probability or severity of an

accident.

The basis of the MCPR Safety Limit calculation is to ensure that

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

boiling if the limit is not violated. The MCPR Safety limit preserves

the existing margin to transition boiling and fuel damage in the event

of a postulated accident. The probability of fuel damage is not

increased.

Therefore, the proposed TS change does not involve an increase in

the probability or consequences of an accident previously evaluated.

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

or different kind of accident from any accident previously evaluated.

The MCPR Safety Limit is a Technical Specification numerical value

designed to ensure that fuel damage from transition boiling does not

occur as a result of the limiting postulated accident. The MCPR Safety

Limit is not an accident initiator; therefore, it cannot create the

possibility of any new type of accident. The new MCPR Safety Limits are

calculated using NRC-approved methods (i.e., GESTAR-II, Rev. 13) and

are based on LGS Unit 1, Cycle 8 specific inputs.

Therefore, the proposed TS change does not create the possibility

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

evaluated.

3. The proposed TS change does not involve a significant reduction

in the margin of safety.

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

same. The new MCPR Safety Limits are calculated using NRC-approved

methods (i.e., GESTAR-II, Rev. 13), which are in accordance with the

current fuel design and licensing criteria, and are based on LGS Unit 1

Cycle 8 specific inputs. The MCPR Safety Limit remains high enough to

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

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

fuel cladding integrity.

Therefore, the proposed TS change does not involve a reduction in

the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Pottstown Public Library, 500

High Street, Pottstown, PA 19464.

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

General Counsel, Philadelphia Electric Company, 2301 Market Street,

Philadelphia, PA 19101.

NRC Project Director: John F. Stolz.

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

FitzPatrick Nuclear Power Plant, Oswego County, New York

Date of amendment request: October 14, 1997.

Description of amendment request: The proposed changes would

correct the maximum exposure dependent, infinite lattice multiplication

factor for fuel bundles and provide for installation of additional

storage racks to increase spent fuel capacity.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

Operation of the FitzPatrick plant in accordance with the proposed

Amendment would not involve a significant hazards consideration as

defined in 10 CFR 50.92, since it would not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated because:

A change in the infinite lattice neutron multiplication factor for

a fuel bundle in the reactor core geometry which ensures the

criticality limit for fuel in the spent fuel pool [SFP] geometry is met

does not affect initiation of any accident.

Operation in accordance with the revised limit ensures the

consequences of previously analyzed accidents are not changed. Storage

of additional fuel assemblies in the pool does not affect the

probability or consequences of dropping a fuel assembly, since this

accident is localized to a small area of the storage array. Likewise,

addition of

[[Page 9614]]

specifications containing details presently in plant design documents

and editorial changes do not change the probability or consequences of

a previously analyzed accident.

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

for any accident previously evaluated because:

A change in the infinite lattice neutron multiplication factor for

a fuel bundle in the reactor core geometry which ensures the

criticality limit for fuel in the spent fuel pool geometry is met does

not affect the types of reactivity accidents which may occur. Therefore

changing the limit will not [create the possibility of] a new or

different type of accident. Maintenance of available decay heat removal

systems ensures that no new type of loss of cooling accident associated

with the SFP will occur as a result of storing additional irradiated

fuel assemblies. Likewise, addition of specifications containing

details presently in plant design documents and editorial changes do

not create the possibility of a new or different type of accident.

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

The revised limit on infinite lattice neutron multiplication factor

for a fuel bundle in the reactor core geometry ensures maintenance of

the same margin of safety with respect to criticality as presently

exists for storage of fuel in the SFP. Storing additional irradiated

fuel assemblies in the pool does not affect the margin of safety with

regard to pool cooling since sufficient heat removal systems will be

maintained available to ensure maintenance of acceptable pool

temperatures. Addition of specifications containing details presently

in other design documents and editorial changes have no effect on the

margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Reference and Documents

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

York 13126.

Attorney for licensee: Mr. David E. Blabey, 1633 Broadway, New

York, New York 10019.

NRC Project Director: S. Singh Bajwa.

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

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

Unit No. 1, Fairfield County, South Carolina

Date of amendment request: February 9, 1998.

Description of amendment request: The proposed amendment would

revise the Virgil C. Summer Nuclear Station Technical Specifications

(TS) to remove emergency diesel generator (1) accelerated testing

requirements (TS 3/4.8.1, Table 4.8-1), and (2) special reporting

requirements (TS Surveillance Requirement 4.8.1.1.3) in accordance with

NRC Generic Letter (GL) 94-01, ``Removal of Accelerated Testing and

Special Reporting Requirements for Emergency Diesel Generators.''

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. This request does not involve a significant increase in the

probability or consequences of an accident previously evaluated.

This change will provide flexibility to structure the emergency

diesel generator maintenance program based on the risk significance of

the structures, systems, and components that are within the scope of

the maintenance rule. The removal of the diesel generator accelerated

testing is acceptable as the maintenance rule applies system and train

specific performance criteria to monitor diesel generator performance.

These criteria include a running availability and reliability measure.

The performance criteria for the diesel generator reliability and

unavailability established by the maintenance rule, and the causal

determinations and corrective actions required for functional failures

and/or exceeding performance criteria, is considered to be an

acceptable method for monitoring diesel generator performance.

As the diesel generator performance will [continue] to be assured

by the maintenance rule, the proposed changes do not affect any of the

initiators for an accident previously evaluated. The changes do not

impact the diesel's design sources, operating characteristics, system

functions, or system interrelationships. The failure mechanisms for the

accidents previously analyzed are not affected, and no additional

failure modes are created that could cause an accident previously

evaluated. Since the changes are administrative in nature, and the

diesel generator performance and reliability will continue to be

assured by the maintenance rule, the proposed changes cannot involve a

significant increase in the probability or consequences of an accident

previously evaluated.

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

different kind of accident from any accident previously evaluated.

This proposed change does not involve a change to the plant design

or operation. As a result, the proposed change does not affect any of

the parameters or conditions that could contribute to the initiation of

any accidents. The proposed changes only affect the methods used to

monitor and assure diesel generator performance. The performance

criteria for both the diesel generator reliability and unavailability

established by the maintenance rule, and the causal determinations and

corrective actions required for functional failures and/or exceeding

performance criteria, is considered by GL 94-01 to be an acceptable

method for monitoring diesel generator performance.

No SSC [structure, system, or component], method of operating, or

system interface is altered by this change. The changes do not impact

the diesel's design sources, operating characteristics, system

functions, or system interrelationships. The failure mechanisms for the

accidents are not affected, and no additional failure modes are

created. Because the proposed changes are administrative in nature, and

the diesel generator performance and reliability will continue to be

assured by the maintenance rule, the proposed changes cannot 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 [of] safety.

The proposed changes only affect the methods used to monitor and

assure diesel generator performance. The performance criteria for both

the diesel generator reliability and unavailability established by the

maintenance rule, and the causal determinations and corrective actions

required for functional failures and/or exceeding performance criteria,

is considered by GL 94-01 to be an acceptable method for monitoring

diesel generator performance. No margin [of] safety as defined in the

basis for any technical specification is impacted by these changes.

This change does not impact any uncertainty in the design,

construction, or operation of any SSC.

[[Page 9615]]

Diesel generator response to accident initiators is unchanged. No SSC,

method of operating, or system interface is altered by this change. The

changes do not impact the diesel's design sources, operating

characteristics, system functions, or system interrelationships.

Because the proposed changes are administrative in nature, and the

diesel generator performance and reliability will continue to be

assured by the maintenance rule, the proposed changes cannot 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: Fairfield County Library, 300

Washington Street, Winnsboro, SC 29180.

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

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

NRC Project Director: William M. Dean.

Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek

Generating Station, Coffey County, Kansas

Date of amendment request: January 28, 1998.

Description of amendment request: The proposed amendment would

revise Technical Specification (TS) Sections 6.3 and 6.12 to reflect

the merger of the positions of Superintendent Radiation Protection and

Superintendent Chemistry into one new position, Manager Chemistry/

Radiation Protection.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase in

the probability or consequences of an accident previously evaluated.

The proposed change does not involve a significant increase in the

probability of consequences of an accident previously evaluated. These

changes involve administrative changes to the WCNOC organization.

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 does not create the possibility of a new or

different kind of accident from any accident previously evaluated. This

change is administrative in nature and does not involve a change to the

installed plant systems or the overall operating philosophy of Wolf

Creek Generating Station.

3. The proposed change does not involve a significant reduction in

a margin of safety.

The proposed change does not involve a significant reduction in a

margin of safety. This change does not involve any changes in overall

organizational commitments and will not affect qualification

requirements of any unit staff personnel. A position and title change

alone does not reduce the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room locations: Emporia State University,

William Allen White Library, 1200 Commercial Street, Emporia, Kansas

66801 and Washburn University School of Law Library, Topeka, Kansas

66621.

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

Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037.

NRC Project Director: William H. Bateman.

Previously Published Notices of Consideration of Issuance of Amendments

to Facility Operating Licenses, Proposed No Significant Hazards

Consideration Determination, and Opportunity for a Hearing

The following notices were previously published as separate

individual notices. The notice content was the same as above. They were

published as individual notices either because time did not allow the

Commission to wait for this biweekly notice or because the action

involved exigent circumstances. They are repeated here because the

biweekly notice lists all amendments issued or proposed to be issued

involving no significant hazards consideration.

For details, see the individual notice in the Federal Register on

the day and page cited. This notice does not extend the notice period

of the original notice.

Northern States Power Company, Docket No. 50-282, Prairie Island

Nuclear Generating Plant, Unit 1, Goodhue County, Minnesota

Date of amendment request: January 15, 1998.

Description of amendment request: The proposed amendment would

initiate a one-time only change for Prairie Island Unit 1 Cycle 19 that

would allow the use of the moveable incore detector system for

measurement of the core peaking factors with less than 75% and greater

than or equal to 50% of the detector thimbles available.

Date of individual notice in the Federal Register: January 30, 1998

(63 FR 4676).

Expiration date of individual notice: March 2, 1998.

Local Public Document Room location: Minneapolis Public Library,

Technology and Science Department, 300 Nicollet Mall, Minneapolis,

Minnesota 55401.

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

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

NRC Project Director: Cynthia A. Carpenter.

Notice of Issuance of Amendments to Facility Operating Licenses

During the period since publication of the last biweekly notice,

the Commission has issued the following amendments. The Commission has

determined for each of these amendments that the application complies

with the standards and requirements of the Atomic Energy Act of 1954,

as amended (the Act), and the Commission's rules and regulations. The

Commission has made appropriate findings as required by the Act and the

Commission's rules and regulations in 10 CFR Chapter I, which are set

forth in the license amendment.

Notice of Consideration of Issuance of Amendment to Facility

Operating License, Proposed No Significant Hazards Consideration

Determination, and Opportunity for A Hearing in connection with these

actions was published in the Federal Register as indicated.

Unless otherwise indicated, the Commission has determined that

these amendments satisfy the criteria for categorical exclusion in

accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),

no environmental impact statement or environmental assessment need be

prepared for these amendments. If the Commission has prepared an

environmental assessment under the special circumstances

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provision in 10 CFR 51.12(b) and has made a determination based on that

assessment, it is so indicated.

For further details with respect to the action see (1) the

applications for amendment, (2) the amendment, and (3) the Commission's

related letter, Safety Evaluation and/or Environmental Assessment as

indicated. All of these items are available for public inspection at

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

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

the particular facilities involved.

Baltimore Gas and Electric Company, Docket No. 50-317, Calvert Cliffs

Nuclear Power Plant, Unit No. 1, Calvert County, Maryland

Date of application for amendment: May 16, 1997, as supplemented

November 14, 1997.

Brief description of amendment: The amendment involves replacing

the service water (SRW) heater exchangers with new plate and frame heat

exchangers (PHEs), having increased thermal performance capability. The

Saltwater (SW) and SRW piping configuration will be modified as

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

scheme to throttle saltwater flow to the heat exchangers and the

associated bypass lines will be added. Saltwater strainers with an

automatic flushing arrangement will be added upstream of each heat

exchanger. The majority of the physical work associated with this

modification is restricted to the SRW pump room. The amendment is

partially denied to the extent that the licensee is not authorized to

operate with one PHE secured, and removing one containment air cooler

from service to enable the affected subsystem to remain operable while

the one PHE is secured.

Date of issuance: February 10, 1998.

Effective date: As of the date of issuance to be implemented within

30 days.

Amendment No.: 225.

Facility Operating License No. DPR-53: Amendment revised the

Updated Final Safety Analysis Report.

Date of initial notice in Federal Register: June 18, 1997 (62 FR

33118).

The November 14, 1997, letter provided clarifying information that

did not change the initial proposed no significant hazards

consideration.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated February 10, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Calvert County Library, Prince

Frederick, Maryland 20678.

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 application for amendments: November 6, 1997, as

supplemented by letter dated January 28, 1998.

Brief Description of amendments: The 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 will

be limited in duration (approximately 3 months) and will allow

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

building ventilation system, including new air conditioning (AC) units

and improved ductwork supports, 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.

Date of issuance: February 6, 1998.

Effective date: February 6, 1998.

Amendment Nos.: 191 and 222.

Facility Operating License Nos. DPR-71 and DPR-62: Amendments

authorize changes to the facility's Technical Specifications.

Date of initial notice in Federal Register: December 3, 1997 (62 FR

63973).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated February 6, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: University of North Carolina

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

Wilmington, North Carolina 28403-3297.

Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455,

Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois, Docket Nos.

STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, Will

County, Illinois

Date of application for amendments: February 28, 1997. Information

related to the proposed restoration of the primary coolant dose

equivalent iodine-131 (DEI) to their original licensing basis had been

previously submitted in Commonwealth Edison Company's (ComEd) letter

dated November 13, 1996, which was supplemented in subsequent letters

dated March 20, June 24, August 19 and November 3, 1997.

Brief description of amendments: The amendments revise the

technical specifications (TS) to reflect the forthcoming replacement of

the original steam generators (OSG) in Byron, Unit 1, and Braidwood,

Unit 1, which are Westinghouse Model D4 steam generators (SG), with the

replacement steam generators (RSG) which are Babcock and Wilcox,

International (BWI) SG. The present revisions to the TS remove the

interim plugging criteria (IPC) related to outer diameter stress

corrosion cracking (ODSCC) in the OSG as well as the F* alternative

repair criteria and two separate SG tube sleeving methodologies which

are not needed for the RSG.

Date of issuance: February 3, 1998

Effective date: This license amendment is effective as of the date

of its issuance and shall be implemented in the first operating cycle

after installation of the BWI replacement steam generators

Amendment Nos.: 101, 101, 92 and 92.

Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77:

The amendments revised the Technical Specifications.

Date of initial notice in Federal Register: December 17, 1997 (62

FR 66134). The November 13, 1996, and March 20, June 24, August 19 and

November 3, 1997, submittals provided clarifying information that did

not change the initial proposed no significant hazards consideration

determination.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated February 3, 1998.

No significant hazards consideration comments received: No

Local Public Document Room location: For Byron, the Byron Public

Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010;

for

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Braidwood, the Wilmington Public Library, 201 S. Kankakee Street,

Wilmington, Illinois 60481.

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

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

Date of application for amendments: November 6, 1995, and March 11,

1996, as supplemented June 5, 1997. The June 5, 1997, letter provided

clarifying information that did not change the initial proposed no

significant hazards consideration determination or expand the amendment

request beyond the scope of the December 20, 1995, and April 10, 1996,

Federal Register notices.

Brief description of amendments: These amendments revise the alarm

setpoints for the effluent radiation and in-containment area radiation

monitors listed in Technical Specification (TS) Table 3.3-6. These

revisions make these alarm setpoints consistent with criteria for the

Emergency Action Levels (EALs) approved by the Nuclear Regulatory

Commission in August 1994. The EALs use these monitors as an indication

of fission product barrier challenges or failures. These amendments

also revise Action Statement 36 of TS Table 3.3-6 to reflect a

previously approved change (License Amendment Nos. 188 and 70) in

reporting frequency (change from semi-annual to annual) for effluent

releases. The revision to Action Statement 36 makes it consistent with

the previously approved change. These amendments include sever

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