Biweekly Notice

Federal RegisterOct 22, 1997

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NUCLEAR REGULATORY COMMISSION

Biweekly Notice

Applications and Amendments to Facility Operating Licenses

Involving No Significant Hazards Considerations

I. Background

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

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

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

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

publish notice of any amendments issued, or proposed to be

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issued, under a new provision of section 189 of the Act. This provision

grants the Commission the authority to issue and make immediately

effective any amendment to an operating license upon a determination by

the Commission that such amendment involves no significant hazards

consideration, notwithstanding the pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all notices of amendments issued, or

proposed to be issued from September 29, 1997, through October 9, 1997.

The last biweekly notice was published on October 8, 1997 (62 FR

52578).

Notice Of Consideration Of Issuance Of Amendments To Facility

Operating Licenses, Proposed No Significant Hazards Consideration

Determination, And Opportunity For A Hearing

The Commission has made a proposed determination that the following

amendment requests involve no significant hazards consideration. Under

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

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

involve a significant increase in the probability or consequences of an

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

or different kind of accident from any accident previously evaluated;

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

for this proposed determination for each amendment request is shown

below.

The Commission is seeking public comments on this proposed

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

publication of this notice will be considered in making any final

determination.

Normally, the Commission will not issue the amendment until the

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

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

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

the Commission may issue the license amendment before the expiration of

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

the amendment involves no significant hazards consideration. The final

determination will consider all public and State comments received

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

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

opportunity for a hearing after issuance. The Commission expects that

the need to take this action will occur very infrequently.

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

Directives Branch, Division of Freedom of Information and Publications

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

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

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

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

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

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

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

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

intervene is discussed below.

By November 21, 1997, 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

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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.

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

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

Date of amendment request: April 7, 1997, as supplemented on August

7, 1997.

Description of amendment request: The proposed amendment would

revise the plants' technical specifications to permit replacement of

the 125 volt dc Gould batteries with new C&D Charter Power Systems,

Inc., batteries.

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 change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The replacement C&D battery has been selected to meet or exceed

the design, functional, and operational requirements of those of the

present Gould battery, including crosstie load limitations. The C&D

batteries are similar in design to the installed Gould batteries

(e.g., electrolyte specific gravity and construction of the plates)

except for capacity. The replacement C&D batteries have a

significantly larger capacity than the Gould batteries, which can

provide additional margin for future use. Also, the C&D batteries

are qualified for a 20 year life and meet the latest applicable

standards. The short circuit current provided by the C&D batteries

is well within the interrupting capability of the existing DC system

circuit breakers.

Additionally, the crosstie limit is increased to take advantage

of the larger C&D battery capacity. The C&D batteries were sized

based on having sufficient capacity to energize the design basis DC

loads for an operating unit with the IEEE-485 design margin while

maintaining the desired limited DC load of 200 amps for a shutdown

unit. This proposed change allows use of the C&D batteries' larger

capacity. The overall design, function, and operation of the DC

system and equipment has not been altered by these changes. The

proposed changes do not affect any accident initiators or precursors

and do not alter the design assumptions for the systems or

components used to mitigate the consequences of an accident as

analyzed in UFSAR Chapter 15. Therefore, there is no increase in the

probability or consequences of an accident previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

The replacement C&D batteries will provide the same functions as

those of the installed Gould batteries and will be operated with the

same types of operational controls. These limits include battery

float terminal voltage, individual cell voltage and electrolyte

specific gravity, and crosstie loading. Crosstie conditions are

allowed under the present Technical Specifications. The crosstie

limit is increased to take advantage of the larger C&D battery

capacity. The remaining changes are administrative in nature or

provide clarification to maintain consistency with other Technical

Specifications.

The DC system and its equipment will continue to perform the

same functions and be operated in the same fashion. The proposed

change does not create any new or common failure modes. The proposed

changes do not introduce any new accident initiators or precursors,

or any new design assumptions for the systems or components used to

mitigate the consequences of an accident. Therefore, the possibility

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

evaluated has not been created.

C. The proposed change does not involve a significant reduction

in a margin of safety.

The replacement C&D batteries will meet or exceed the design,

functional, and qualification requirements [of] those of the

installed Gould batteries. The proposed Technical Specification

limitations for the C&D batteries are derived from the same

methodology as the Gould batteries with applied margins in

accordance with IEEE-485. Increasing the crosstie loading limit

takes advantage of the larger C&D battery capacity with its

increased design margin. The proposed change to the crosstie loading

limit will continue to conservatively envelope the postulated design

requirements. The remaining changes are administrative in nature or

provide clarification to maintain consistency with other Technical

Specifications.

The inherent design conservatism of the DC system and its

equipment has not been altered. The DC system and its equipment will

continue to be operated with the same degree of conservatism.

Therefore, there is no reduction in the margin of safety.

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

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

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

requested amendments involve no significant hazards consideration.

Local Public Document Room location: Byron Public Library District,

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

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: June 30, 1997, as supplemented on

September 25, 1997.

Description of amendment request: The proposed amendment would

revise the plants' technical specifications to permit the licensee to

take credit for soluble boron in spent fuel storage pool water to

maintain an acceptable margin of subcriticality.

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 following accidents have been specifically evaluated

relative to the SFP [spent fuel pool]: fuel assembly drop,

accidental misloading of spent fuel

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assemblies into the SFP racks, and loss of normal cooling.

There is no increase in the probability of a fuel assembly drop

accident in the SFP when considering the presence of soluble boron

in the SFP water for criticality control. The handling of the fuel

assemblies in the SFP has previously been performed in borated

water. The criticality analysis shows the consequences of a fuel

assembly drop accident in the SFP are not affected when considering

the presence of soluble boron.

There is no increase in the probability of the accidental

misloading of spent fuel assemblies into the SFP racks when

considering the presence of soluble boron in the pool water for

criticality control. Fuel assembly placement will continue to be

controlled in accordance with approved fuel handling procedures and

the spent fuel storage configuration limitations. Periodic

surveillances of the SFP inventory (physical inventory and piece

counts) are performed in accordance with station procedures. These

surveillances ensure physical SFP inventory verification is

performed at least once per year and in a timely manner upon

completion of fuel movement in the SFP. The addition of credit for

decay time in the spent fuel pool in determining allowable storage

requirements is an extension of the reactivity equivalencing

methodologies used for burnup credit in WCAP-14416-NP-A,

``Westinghouse Spent Fuel Rack Criticality Analysis Methodology,''

Revision 1, November 1996.

There is no increase in the consequences of the accidental

misloading of spent fuel assemblies into the SFP racks because

criticality analyses demonstrate that the pool will remain

subcritical following an accidental misloading if the pool contains

an adequate boron concentration. The proposed TS limitations and

surveillance frequency will ensure that an adequate SFP boron

concentration is maintained.

There is no increase in the probability of the loss of normal

cooling to the SFP water when considering the presence of soluble

boron in the pool water for subcriticality control since a high

concentration of soluble boron has previously been maintained in the

SFP water. A loss of normal cooling to the SFP water causes an

increase in the temperature of the water passing through the stored

fuel assemblies. This causes a decrease in water density which would

result in a decrease in reactivity when Boraflex neutron absorber

panels are present in the racks. However, since the proposed change

does not consider Boraflex to be present in the racks, and the SFP

water has a high concentration of boron, a density decrease causes a

positive reactivity addition. [The] consequences of this accident

are bounded by the misloaded assembly analysis. Because adequate

soluble boron will be maintained in the SFP water, the consequences

of a loss of normal cooling to the SFP will not be increased.

The proposed 48 hour surveillance frequency will be used to

verify the boron concentration is within the initial assumptions of

the criticality analysis. The current frequency of 24 hours was

based on the sampling frequency for reactor coolant system (RCS)

shutdown margin in Mode 5. A dilution of the SFP to a

keff greater than 0.95 would take a much longer time than

an RCS dilution resulting in loss of shutdown margin. This is due to

the larger SFP volume compared to the RCS volume, and the turnover

rate of water in the SFP is much less due to the lack of large

dilution sources for the SFP. The 48 hour sampling frequency is

sufficient based on operating experience, and based on the fact that

significant changes in the boron concentration in the spent SFP are

difficult to produce without detection, due to the large inventory

of water. Soluble boron concentration reduction requires the inflow

and outflow of large volumes of water which are readily detected by

SFP and fuel handling building sump high level alarms, flooding in

the fuel handling building or by normal operator rounds through the

SFP area (once every eight hours), allowing adequate time for

operator intervention prior to exceeding a keff of 0.95.

Therefore, consequences of an accident previously evaluated are not

increased by the change in surveillance frequency.

The format revisions to Specification 5.6.1.1 and reference to

the report containing the specific NRC-approved criticality

methodology in Specification 6.9.1.10 are administrative in nature

and will not result in an increase in the probability or

consequences of an accident previously evaluated.

Therefore, based on the above analysis, the proposed changes

will not involve a significant increase in the probability or

consequences of an accident previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

The results of criticality accident analyses in the SFP are

discussed in the UFSAR [Updated Final Safety Analysis Report] and in

Criticality Analysis Reports associated with previous licensing

activities. Specific accidents considered include fuel assembly

drop, accidental misloading of spent fuel assemblies into the SFP

racks, and loss of normal cooling.

LCO 3.9.1, ``BORON CONCENTRATION,'' contains limitations on the

boron concentration in the filled portions of the reactor coolant

system and the refueling canal during Mode 6. ComEd has maintained

soluble boron in the SFP at all times and has imposed administrative

limits on the SFP boron concentration, due in part to this

requirement. LCO 3.9.11 establishes specific boron concentration

requirements for the SFP water consistent with the results of the

new criticality analysis based on the NRC-approved methodology of

WCAP-14416-NP-A, ``Westinghouse Spent Fuel Rack Criticality Analysis

Methodology,'' Revision 1, November 1996. Credit is also taken for

radioactive decay time of the spent fuel.

Since soluble boron has always been maintained in the SFP water

and is currently controlled administratively, the implementation of

this requirement will have little effect on normal pool operations

and maintenance. The implementation of the proposed limitations on

the SFP boron concentration will only result in a requirement to

verify boron concentration of the SFP water every 48 hours rather

than every 24 hours. Sampling every 48 hours is sufficient to verify

the SFP boron concentration meets the assumptions of the criticality

analysis.

Because soluble boron has always been present in the SFP and has

been administratively controlled, a dilution of the SFP soluble

boron has always been a possibility. As shown in the SFP dilution

evaluation performed for Byron and Braidwood, a dilution of the SFP

which could increase the rack keff to greater than 0.95

(i.e., which could reduce the required margin to criticality) is not

a credible event.

Therefore, the implementation of the proposed limitations on the

SFP boron concentration and surveillance frequency will not result

in the possibility of a new kind of accident.

The proposed change to Specification 5.6.1.1 identifies the

requirements for the spent fuel rack storage configurations. The

proposed changes relate to the criteria for determining the storage

configuration. Since the proposed SFP storage configuration

limitations will be similar to those currently in the Byron and

Braidwood TS, these limitations will not have any significant effect

on normal SFP operations and maintenance and will not create any

possibility of a new or different kind of accident. Verifications

will continue to be performed to ensure that the SFP loading

configuration meets specified requirements.

The format revisions to Specification 5.6.1.1 and reference to

the report containing the specific NRC-approved criticality

methodology in Specification 6.9.1.10 are administrative in nature

and will not create the possibility of a new [or] different kind of

accident.

As discussed above, there is no significant change in plant

configuration or equipment and the proposed changes will not create

the possibility of a new or different kind of accident.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The proposed TS changes and the resulting spent fuel storage

operating limits will provide adequate safety margin to ensure that

the stored fuel assembly array will always remain subcritical. These

limits are based on a plant specific criticality analysis performed

in accordance with the NRC-approved Westinghouse spent fuel rack

criticality analysis methodology (WCAP-14416-NP-A). Credit is also

taken for radioactive decay time of the spent fuel.

Soluble boron credit provides significant negative reactivity in

the SFP such that the keff is maintained less than or

equal to 0.95. The proposed surveillance frequency will be used to

verify the boron concentration is within the initial assumptions of

the criticality analysis. A storage configuration has also been

defined, with a 95-percent probability at a 95-percent confidence

level, that ensures the spent fuel rack keff will be less

than 1.0 with no credit for soluble boron or Boraflex panels in the

racks. In addition to soluble boron credit, credit is taken for fuel

assembly burnup, decay time, and IFBAs [Integral Fuel Burnable

Absorber] when determining assembly storage requirements.

[[Page 54870]]

The loss of substantial amounts of soluble boron from the SFP

which could lead to exceeding a keff of 0.95 has been

evaluated and shown not to be credible. These evaluations show that

the dilution of the SFP boron concentration from 2000 ppm to 550 ppm

is not credible and that the spent fuel rack keff will

remain less than 1.0 when flooded with unborated water.

The format revisions to Specification 5.6.1.1 and reference to

the report containing the specific NRC-approved criticality

methodology in Specification 6.9.1.10 are administrative in nature

and will not result in a significant reduction in the plant's margin

of safety.

Therefore, the proposed changes in this license amendment will

not result in a significant reduction in the plant's margin of

safety.

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

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

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

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. 50-373 and 50-374, LaSalle

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

Date of amendment request: August 12, 1997

Description of amendment request: The proposed amendments would

remove a Technical Specification surveillance requirement to verify

that sediment deposition within the lake screenhouse is not greater

than one foot in thickness. Control of sediment accumulation in the

lake screenhouse would be accomplished through the Service Water

Performance Monitoring Program.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1) Involve a significant increase in the probability or

consequences of an accident previously identified because:

Surveillance's [sic] to fully verify [that] the Ultimate Heat

Sink contains enough water to perform its design function will

continue. All cleanliness issues associated with ensuring

operability of Core Standby Cooling System - Equipment Cooling Water

System (CSCS-ECWS) equipment will be performed under the Service

Water Performance Monitoring Program, which meets GL 89-13

[Service Water System Problems Affecting Safety-Related

Equipment] recommended actions. By performing these

inspections per GL 89-13, LaSalle will ensure that there is no build

up of sediment, which could hinder or impede the design operation of

any safety or non-safety related equipment which takes a suction

from the service water tunnel. Based on the nature of sediment,

where it collects, and system design, the CSCS-ECWS will be

available if called upon or started to respond in case of an

accident for equipment cooling and long term cooling.

At no time, during approximately fourteen years of LaSalle

operation, has sediment built up or accumulated either in front of

the inlet to the CSCS cooling water screen bypass supply line or the

six 36-inch normal tunnel supply lines in such a manner that the

flow of water through these lines could have been reduced or

blocked. Instead, loose sediment collects in quiescent areas near

the traveling screens, the north end of the Service Water Tunnel,

under the outlets of the 36-inch normal tunnel supply lines in the

service water tunnel, and downstream of the butterfly isolation

valve in the 54 inch CSCS cooling water screen bypass supply line.

The sediment that collects in the service water tunnel does not

build up in a manner such that CSCS-ECWS, non-essential station

service water, or fire pump suctions from the tunnel are affected,

based on inspections since 1992.

The CSCS equipment cooling bypass valve, OE12-F300, is the

manual butterfly valve in the CSCS cooling water screen bypass

supply line. The bypass valve is being added to the ASME Section XI

Inservice Testing Program to cycle the valve quarterly. This valve

cycling will help maintain sediment level in the bypass line at a

low level due to flow through the line while the valve is not fully

closed and thus assure the bypass line remains available. The flow

is created due to the differential pressure across the circulating

water traveling screens with circulating water pumps in operation.

Therefore, neither essential nor non-essential service water

will be lost due to sediment. Neither the probability nor the

consequences of an accident are increased by the deletion of SR

4.7.1.3.c.

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

from any accident previously evaluated because:

Inspections for sedimentation will continue to be required by

LaSalle's Service Water System Performance Monitoring Program per GL

89-13, to ensure continued operability of Core Standby Cooling

System-Equipment Cooling Water System (CSCS-ECWS). The Ultimate Heat

Sink operability requires assurance of a specific volume of water to

provide cooling for at least 30 days for long term cooling following

an accident. The public will be protected by the safety analysis in

place by the fact that the safety and non-safety related equipment

which take a suction from the service water tunnel will not be

impaired by sediment. Therefore, there will be no possibility of a

new or different kind of accident from any accident previously

evaluated.

3) Involve a significant reduction in the margin of safety

because:

The Ultimate Heat Sink continues to be demonstrated Operable by

verifying a sufficient volume of water per TS SR 4.7.1.3.a and

4.7.1.3.b. Equipment operability will still be required per

Technical Specifications 3/4.7.1.1 and 3/4.7.1.2 for the CSCS-ECWS

systems. Sedimentation in the lake screenhouse is a maintenance/

cleanliness issue addressed by the LaSalle Service Water Performance

Monitoring Program. The program ensures equipment operability by

both inspection for and removal of sedimentation and chemical

control with a biocide to limit the growth of biological material

and silt dispersant to help keep silt in the flow stream from

coagulating. Therefore, there is minimal or no reduction in the

margin of safety due to the deletion of this surveillance

requirement.

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

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

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

requested amendments involve no significant hazards consideration.

Local Public Document Room location: Jacobs Memorial Library,

Illinois Valley Community College, Oglesby, Illinois 61348

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: September 29, 1997 (NRC-97-0089)

Description of amendment request: The proposed amendment would

relocate the requirements for selected instrumentation and the

associated Bases from the technical specifications (TS) to the updated

final safety analysis report. The affected instrumentation is seismic

monitoring (TS 3.7.2), meteorological monitoring (TS 3.7.3), the

traversing in-core probe system (TS 3.7.7), the chlorine detection

system (TS 3.7.8), and the loose parts detection system (TS 3.7.10).

Changes to the TS index and list of tables were also requested to

reflect the relocation of these TS and associated Bases. NRC Generic

Letter 95-10, ``Relocation of Selected Technical Specification

Requirements Related to Instrumentation,'' dated December 15, 1995,

provided information concerning relocation of the requirements for

these instruments.

[[Page 54871]]

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 would relocate TS 3/4.3.7.2 - Seismic

Monitoring Instrumentation, TS 3/4. 3.7.3 - Meteorological

Monitoring Instrumentation, TS 3/4.3.7.7 - Traversing In-Core Probe

System, TS 3/4.3.7.8 - Chlorine Detection System, and TS 3/4.3.7.10

- Loose-Part Detection System and their associated Bases to the

Fermi 2 Updated Final Safety Analysis Report (UFSAR). They would

also delete the special reporting requirements from the

aforementioned TS which contain such requirements. The proposed

changes would revise the TS Index and List of Tables to reflect the

relocation of these TS and associated Bases. The relocated TS

changes would be controlled in accordance with the requirements of

10 CFR 50.59.

The proposed changes affect TS that do not meet the NRC's

``Final Policy Statement on Technical Specification Improvements for

Nuclear Power Reactors'' or 10 CFR 50.36(c)(2)(ii) criteria for

inclusion in TS. These TS relocations are consistent with NUREG-

1433, ``Standard Technical Specifications, General Electric Plants,

BWR/4,'' Revision 1, April 1995. Furthermore, these five TS are

specifically identified in NRC Generic Letter 95-10, ``Relocation of

Selected Technical Specifications Requirements Related to

Instrumentation,'' dated December 15, 1995, as suitable for

relocation to licensee-controlled documents.

The Special Report requirements of TS 3/4.3.7.2, TS 3/4.3.7.3,

and TS 3/4.3.7.10 would be deleted as part of their relocation to

the UFSAR. The NRC reporting criteria of 10 CFR 50.72, ``Immediate

Notification Requirements for Operating Nuclear Reactors,'' and 10

CFR 50.73, ``Licensee Event Report Systems'' provide appropriate

requirements for reporting degraded and non-conforming conditions to

the NRC.

These proposed TS changes do not involve a significant increase

in the probability of an accident previously evaluated because no

changes are being made to any accident initiator. No previously

analyzed accident scenario is changed, and initiating conditions and

assumptions remain as previously analyzed.

These proposed TS changes do not involve a significant increase

in the consequences of an accident previously evaluated because the

proposed changes do not affect accident sequences or assumptions

used in evaluating the radiological consequences of an accident. The

proposed changes do not alter the source term, containment isolation

or allowable radiological releases.

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

different kind of accident from any accident previously evaluated.

The proposed changes do not change the way in which the plant is

operated and no new or different failure modes have been defined for

any plant system or component. No limiting single failure has been

identified as a result of the proposed changes. No new or different

types of failures or accident initiators are introduced by the

proposed changes.

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

margin of safety.

The proposed changes involve instrumentation and systems which

are not inputs in the calculation of any safety margin with regard

to Technical Specification Safety Limits, Limiting Safety System

Settings, Limiting Control Settings or Limiting Conditions for

Operation, or other previously defined margins for any structure,

system, or component.

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: John N. Hannon

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

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

Pennsylvania

Date of amendment request: September 11, 1997

Description of amendment request: The proposed amendments would

relocate the reactor trip system and engineered safety feature

actuation system response times from technical specification (TS)

tables 3.3-2 and 3.3-5 to Section 3 of the licensee's Licensing

Requirements Manual (LRM) in accordance with the guidance provided in

NRC Generic Letter 93-08. Subsequent changes to the LRM would be

controlled in accordance with the requirements of 10 CFR 50.59. The

proposed amendments would also make several editorial changes in TSs

3.3.1.1 and 3.3.1.2, as well as making conforming changes to the Bases

for these TSs.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The proposed amendment relocates the instrument response time

limits for the reactor trip system (RTS) and engineered safety

feature actuation system (ESFAS) from the technical specifications

to the Licensing Requirements Manual (LRM). The Core Operating

Limits Report (COLR) and containment penetrations table (containment

isolation valves) are controlled and maintained in the LRM. The LRM

was developed to control and maintain those items removed from the

technical specifications. The proposed amendment conforms to the

guidance given in Enclosures 1 and 2 of Generic Letter 93-08.

Neither the response time limits nor the surveillance requirements

for performing response time testing will be altered by this

submittal. The overall RTS and ESFAS functional capabilities will

not be changed and assurance that action requirements of the

protective and engineered safety features systems are completed

within the time limits assumed in the accident analyses is

unaffected by the proposed amendment. Therefore, operation of the

facility in accordance with the proposed amendment will not involve

a significant increase in the probability or consequences of an

accident previously evaluated.

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

kind of accident from any accident previously evaluated?

The proposed amendment will not change the physical plant or the

modes of plant operation defined in the operating license. The

change does not involve the addition or modification of equipment

nor does it alter the design or operation of plant systems.

Therefore, operation of the facility in accordance with the proposed

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

of accident from any accident previously evaluated.

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

of safety?

The measurement of instrumentation response times at the

frequencies specified in the technical specification provides

assurance that actions associated with the protective and engineered

safety features systems are accomplished within the time limits

assumed in the accident analyses. The response time limits, and the

measurement frequencies remain unchanged by the proposed amendment.

The proposed changes do not alter the basis for any other technical

specification that is related to the establishment of or maintenance

of a nuclear safety margin. Therefore, operation of the facility in

accordance with the proposed amendment will not involve a

significant reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

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

663 Franklin Avenue, Aliquippa, PA 15001

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

Potts &

[[Page 54872]]

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

NRC Project Director: John F. Stolz

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

Nuclear Station, Unit 1, Claiborne County, Mississippi

Date of amendment request: September 18, 1997

Description of amendment request: The amendment would decrease the

safety limit for the minimum critical power ratio (MCPR) from 1.12 to

1.11 for two recirculation loop operation and from 1.14 to 1.12 for

single recirculation loop operation in Technical Specification (TS)

2.1.1.2. Because the proposed amendment is for Cycle 10 operation, the

amendment would also revise the footnotes to TSs 2.1.1.2 and 5.6.5 to

state that the MCPR values and the items 19 and 20 are ``applicable

only for Cycle 10 operation.'' Cycle 10 operation is after the next

(i.e., 9th) 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:

I. The proposed change does not significantly increase the

probability or consequences of an accident previously evaluated.

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

in the Bases to Technical Specification [TS] 2.1.1 as that limit

which ``ensures that during normal operation and during Anticipated

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

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

limit is re-evaluated for each reload and, for GGNS [Grand Gulf

Nuclear Station, Unit 1] Cycle 10, the analyses have concluded that

a two-loop MCPR safety limit of 1.11 based on the application of

GE's [General Electric Company's] cycle-specific MCPR safety limit

methodology is necessary to ensure that this acceptance criterion is

satisfied. For single-loop operation, a MCPR safety limit of 1.12

based on GE's cycle-specific MCPR safety limit methodology was

determined to be necessary. Core MCPR operating limits are developed

to support the Technical Specification [TS] 3.2 requirements and

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

case transient. Since the MCPR safety limit will be maintained at

all times, operation under the proposed changes will ensure [that]

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

transition boiling. Therefore, these changes to the [MCPR] safety

limit do not affect the probability or consequences of an accident

[previously evaluated].

GE's GESTAR-II approved methodology will continue to be

implemented and has no effect on the probability or consequences of

any accidents previously evaluated. One exception to GESTAR is that

the mis-oriented and mis-located bundle events will continue to be

analyzed as accidents subject to the acceptance criteria in the

current licensing basis [for GGNS]. The design of the GE11 fuel

bundles[, to be added to the core to replace Siemens fuel bundles,]

is such that the bundles are not likely to be mis-oriented or mis-

located and the normal administrative controls will be in effect for

assuring proper orientation and location. Therefore, the probability

of a fuel loading error is not increased. This analysis ensures that

postulated dose releases will not exceed a small fraction (10

percent) of 10CFR100 [10 CFR Part 100] limits. Therefore, the

probability or consequences of accidents previously evaluated are

unchanged.

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

or different kind of accident from any accident previously

evaluated.

The GE 11 fuel to be [added to the core and] used in Cycle 10

[operation] is of a design compatible with fuel present in the core

and used in the [current 9th] cycle. [The current core is a mixture

of GE11 and Siemens fuel bundles. The addition of GE11 to the core

for the 9th cycle is addressed in Amendment 131 to the license dated

November 21, 1996.] Therefore, the GE11 fuel will not create the

possibility of a new or different kind of accident. The proposed

changes do not involve any new modes of operation, any changes to

setpoints, or any plant modifications.

They introduce revised MCPR safety limits that have been proven

to be acceptable for Cycle 10 operation. Compliance with the

applicable criterion for incipient boiling transition continues to

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

creation of any new precursors to an accident.

Therefore, the proposed changes do not create the possibility of

a new or different kind of accident from any accident previously

evaluated.

III. The proposed change does not involve a significant

reduction in a margin of safety.

The MCPR safety limits have been evaluated in accordance with

GE's current cycle-specific methodology to ensure that during normal

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

core are not expected to experience transition boiling. Unless

otherwise approved, GGNS will implement only the NRC-approved

revisions to GE's GESTAR methodology. This GE methodology is similar

to those SPC [(Siemens Power Corporation)] reports current listed in

TS 5.6.5 and it will be applied in a similar, conservative fashion.

[TS 5.6.5, Core Operating Limits Report, lists the analytical

methods which are approved by NRC and are used to determine the core

operating limits for the GGNS core, including the MCPR.] One

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

events will continue to be analyzed as accidents subject to the

acceptance criteria in the current [GGNS] licensing basis. This

analysis ensures that postulated dose releases will not exceed a

small fraction (10 percent) of 10CFR100 limits. [The proposed

changes are to maintain the margin of safety for transition boiling

in the core.] On this basis, the implementation of this GE

methodology does not involve a significant reduction in a margin of

safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room Location: Judge George W. Armstrong

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

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

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

NRC Project Director: James W. Clifford, Acting

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

Electric Station, Unit 3, St. Charles Parish, Louisiana

Date of amendment request: September 25, 1997

Description of amendment request: The proposed change modifies

Limiting Condition for Operation (LCO) 3.6.1.2 (Containment Leakage),

the associated Action, and Surveillance Requirement (SR) 4.6.1.2 in

Technical Specification (TS) for Waterford Steam Electric Station, Unit

3 (Waterford 3). The air lock door seal leakage rate acceptance

criteria in TS 6.15 is being changed from 0.01La to 0.005La. TS 6.15 is

also being modified to make the terms used in the Containment Leakage

Rate Testing Program consistent with terms used in the TS. This change

corrects an error that inadvertently decreased the allowed outage time

from 24 hours to 1 hour when the containment purge valve or containment

air lock leakage rates are not within limits. This error was made in

the Waterford 3 TS change request that was approved in Amendment 124

for Waterford 3 on April 10, 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. Will operation of the facility in accordance with this

proposed change involve a significant increase in the probability or

consequences of an accident previously evaluated?

Response: No

The proposed change adds the specific type of containment

leakage to the Limiting Condition for Operation (LCO), Action, and

Surveillance Requirement (SR) in the Containment Leakage Technical

Specification (TS) which results in increasing

[[Page 54873]]

the allowed outage time from 1 hour to 24 hours when the containment

purge valve or containment air lock leakage rates are not within

limits. The proposed change revises the air lock door seal leakage

rate acceptance criteria. Also, the proposed change revises the

Actions in the Containment Leakage TS to be consistent with the

Applicability, and revises terms in the Containment Section and

Administrative Controls Section of the TS to be consistent with the

Containment Leakage Rate Testing Program. This change will not

affect the probability of an accident. The containment purge valve

and air lock leakage rates are not an initiator of any analyzed

event. This change corrects two errors that were made in the

Waterford 3 10CFR50 Appendix J, Option B, TS change request that was

approved in TS Amendment 124. The first error inadvertently

decreased the allowed outage time from 24 hours to 1 hour when

either the containment purge valve or containment air lock leakage

rate acceptance criteria is not met. The second error inadvertently

increased the acceptance criteria for the air lock door seal

leakage. The revised air lock door seal leakage rate acceptance

criteria was never used at Waterford 3. This change also

administratively changes the Containment Leakage TS Action and terms

in the TS for consistency.

The proposed change will not affect the consequences of an

accident. The amount of leakage from the containment purge valve and

from the containment air lock will still be included in the overall

combined containment leak rate. Neither the overall containment

leakage rate limit nor the Action required to be taken if the

overall containment leakage rate were exceeded is being changed. The

Containment Leakage TS Action will be consistent with the

Applicability and TS 3.0.4 will prohibit entry into Mode 4 (RCS

[Reactor Coolant System] temperature 200 deg.F), unless

the overall containment leakage rate is within limit. The revised

air lock acceptance criteria was never used. Waterford 3 will

continue using the more restrictive acceptance criteria which is

controlled administratively. This proposed change does not affect

the mitigation capabilities of any component or system, nor does it

affect the assumptions relative to the mitigation of accidents or

transients.

Therefore, the proposed change will not involve a significant

increase in the probability or consequences of any accident

previously evaluated.

2. Will operation of the facility in accordance with this

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

accident from any accident previously evaluated?

Response: No

The proposed change adds the specific type of containment

leakage to the LCO, Action, and SR in the Containment Leakage TS.

This results in increasing the allowed outage time from 1 hour to 24

hours when the containment purge valve or containment air lock

leakage rates are not within limits. The proposed change revises the

air lock door seal leakage rate acceptance criteria. Also, the

proposed change revises the Actions in the Containment Leakage TS to

be consistent with the Applicability, and revises terms in the

Containment Section and Administrative Controls Section of the TS to

be consistent with the Containment Leakage Rate Testing Program.

Neither the design nor configuration of the plant, or how the plant

is operated is being changed due to the addition of the specific

types of leakage from the Containment Leakage Rate Testing Program,

corrections made to the air lock door seal leakage rate acceptance

criteria, or the changes made to make the TS consistent. There has

been no physical change to plant systems, structures, or components

nor will these changes reduce the ability of any of the safety-

related equipment required to mitigate anticipated operational

occurrences or accidents. Therefore, the proposed change will not

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

any accident previously evaluated.

3. Will operation of the facility in accordance with this

proposed change involve a significant reduction in a margin of

safety?