Federal Property Suitable as Facilities to Assist the Homeless

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 January 5, 1995, through January 20, 1995.

The last biweekly notice was published on January 18, 1995 (60 FR

3669).

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 Rules Review and

Directives Branch, Division of Freedom of Information and Publications

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

Washington, DC 20555, 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. [[Page 6297]] 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 3, 1995, 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, Attention: Docketing and Services

Branch, or may be delivered to the Commission's Public Document Room,

the Gelman Building, 2120 L Street, NW., Washington DC, by the above

date. Where petitions are filed during the last 10 days of the notice

period, it is requested that the petitioner promptly so inform the

Commission by a toll-free telephone call to Western Union at 1-(800)

248-5100 (in Missouri 1-(800) 342-6700). The Western Union operator

should be given Datagram Identification Number N1023 and the following

message addressed to (Project Director): petitioner's name and

telephone number, date petition was mailed, plant name, and publication

date and page number of this Federal Register notice. A copy of the

petition should also be sent to the Office of the General Counsel, U.S.

Nuclear Regulatory Commission, Washington, DC 20555, 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.

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

Station, Plymouth County, Massachusetts

Date of amendment request: November 22, 1994.

Description of amendment request: The proposed amendment would

increase the current Emergency Diesel Generator (EDG) allowed out-of-

service time in Specification 3.5.F from 72 hours to 7 days, deletes

the daily testing of the operable diesel generator in Specification

4.5.F.1, when it is determined that the other diesel generator is

inoperable, and revises specification 3.9.B.1 and 2 for EDG

operability.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

[[Page 6298]] (1) The proposed amendment does not involve a

significant increase in the probability of consequences of an

accident previously evaluated.

Operation of PNPS [Pilgrim Nuclear Power Station] in accordance

with the proposed license amendment will not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

Implementation of the proposed change is expected to result in

an increase in the probability of core damage, from 5.85E-5/year

(this is the PNPS IPE [individual plant examination] core damage

frequency) to 5.88E-5/year. This increase is less than one percent

and is considered to be insignificant relative to the underlying

uncertainties involved with probabilistic risk assessments.

Deleting the testing requirement for an EDG when the other EDG

is in repair does not increase the probability or consequences of an

accident previously evaluated because the reliability program and

Technical Specification required surveillances continue to provide

the added assurance sought by the testing. The elimination of this

testing might improve the overall reliability of the EDGs.

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

new or different kind of accident from any accident previously

evaluated.

Operation of PNPS in accordance with the proposed license

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

of accident from any accident previously evaluated. No change is

being made in the manner in which the EDG's provide plant

protection. No new modes of plant operation are involved. Extending

the EDG OOS [out of service] and, deleting the testing requirement

for one EDG when the other EDG is in repair does not necessitate

physical alteration of the plant or changes in plant operational

limits.

3. The proposed amendment does not involve a significant

reduction in a margin of safety.

Operation of PNPS in accordance with the proposed license

amendment will not involve a significant reduction in a margin of

safety. [***], incorporation of the proposed change involves an

insignificant reduction in the margin of safety.

As previously stated, implementation of the proposed changes is

expected to result in an insignificant increase in: (1) power

unavailability to the emergency buses (given that a loss of offsite

power has occurred), and (2) core damage frequency. EDG reliability

improvement is expected due to increased quality and thoroughness of

EDG maintenance. Implementation of the proposed changes does not

increase the consequences of a previously analyzed accident nor

significantly reduce a margin of safety. Functioning of the EDGs and

the manner in which limiting condition of operability are

established are unaffected.

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 02360.

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

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

NRC Project Director: Walter R. Butler.

Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson

Steam Electric Plant, Unit No. 2, Darlington County, South Carolina

Date of amendment request: December 27, 1994.

Description of amendment request: The requested Technical

Specifications (TS) change relocates the turbine rotor inspection

requirement, TS 4.1-3, Item 13, to the Updated Final Safety Analysis

Report (UFSAR), Section 10.2. This TS requires a turbine inspection,

including visual, magnaflux, and dye petrant inspections on a frequency

of every five years with a maximum time between tests of six years.

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 requested change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated. The requested change relocates the turbine inspection

requirement from the TS to the UFSAR. Turbine inspections will

continue to be controlled and performed such that the low turbine

missile generation probability will be maintained. The consequences

of missile generation are unchanged since this change does not

involve the addition or modification of plant equipment, nor does it

alter the design or operation of plant systems. Therefore, there

would be no increase in the probability or consequences of an

accident previously evaluated.

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

or different kind of accident from any accident previously

evaluated. The requested change relocates the turbine inspection

requirement from the TS to the UFSAR. Turbine inspections will

continue to be controlled and performed such that the low turbine

missile generation probability will be maintained. This change does

not involve the addition or modification of plant equipment, nor

does it alter the design or operation of plant systems. Therefore,

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

different kind of accident from any accident previously evaluated.

3. The requested change does not involve a significant reduction

in the margin of safety. The requested change relocates the turbine

inspection requirement from the TS to the UFSAR. Turbine inspections

will continue to be controlled and performed such that the low

turbine missile generation probability will be maintained.

Therefore, the proposed changes do not involve a significant

reduction in a margin of safety.

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

this review, it appears that the three standards of 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: Hartsville Memorial Library,

147 West College Avenue, Hartsville, South Carolina 29550.

Attorney for licensee: R. E. Jones, General Counsel, Carolina Power

& Light Company, Post Office Box 1551, Raleigh, North Carolina 27602.

NRC Project Director: William H. Bateman.

Carolina Power & Light Company, et al., Docket No. 50-400, Shearon

Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties,

North Carolina

Date of amendment request: December 19, 1994.

Description of amendment request: The proposed one-time schedular

extension would allow the third test of the first 10-year service

period to be performed during refueling outage no. 7, at approximately

a 54 month interval instead of the current maximum Technical

Specification interval of 50 months, and coincident with the 10-year

service period to be performed during refueling outage no. 7 and the

10-year inservice inspection,

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

This [extension] request applies to the ILRT [integrated leak

rate testing] and does not affect the local leak rate testing of

containment penetrations and isolation valves where the majority of

the leakage occurs. The allowable containment leakage used in the

accident analysis for offsite doses, La, is 0.1 [weight percent

per day] and for conservatism the leakage is limited to 75 percent

La at startup to account for the possible degradation of

containment leakage barriers between two ILRT tests. Based on the

``as left'' leakage data for the past two ILRTs, the additional time

period added to the testing interval would not adversely impact the

containment leakage barriers to the extent [[Page 6299]] that

degradation would cause leakage to exceed that assumed in the

accident analysis.

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

new or different kind of accident from any accident previously

evaluated.

The change to the Surveillance Requirement is a one time

[extension] to extend the surveillance interval from the maximum of

50 months to approximately 54 months for performance of the third

ILRT in the first service period. There are no design changes being

made that would create a new type of accident or malfunction and the

method and manner of plant operation remain unchanged. Extension of

the surveillance interval for performing the ILRT does not adversely

impact the surveillances ability to show that containment integrity

is maintained.

3. The proposed amendment does not involve a significant

reduction in the margin of safety.

There are no changes being made to the safety limits or safety

system settings that would adversely impact plant safety. The change

is a one time [extension] to extend the time interval for performing

an ILRT approximately four months beyond the current maximum

interval. In addition to the indication of continued containment

integrity provided by the Local Leak Rate Testing program, the

surveillance test data from the first and second ILRTs illustrates

that there is sufficient leakage margin to remain well below the

allowable leakage rate of La. The as-left leakage rate for the

last ILRT was 0.0614 [weight percent per day], which is well below

the 0.075 [weight percent per day] allowed by the T.S., and

therefore provides margin for degradation that is greater than the

minimum provided by the Technical Specifications. Therefore, this

change does not significantly reduce the margin of safety for

Technical Specification 3.6.1.2.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Cameron Village Regional

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

Attorney for licensee: R. E. Jones, General Counsel, Carolina Power

& Light Company, Post Office Box 1551, Raleigh, North Carolina 27602.

NRC Project Director: William H. Bateman.

Consumers Power Company, Docket No. 50-255, Palisades Plant, Van

Buren County, Michigan

Date of amendment request: December 29, 1994.

Description of amendment request: The proposed amendment would

affect the method of controlling the pH of the post-LOCA containment

sump solution by allowing the replacement of the existing operator

actuated Iodine Removal System with a passive system of baskets of

Trisodium Phosphate (TSP) in the lower regions of the containment. The

current Iodine Removal System provides sodium hydroxide (NaOH) for

injection into the containment spray to maintain pH of the sump

solution.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

The following evaluation supports the finding that operation of the

facility in accordance with the proposed change from NaOH to TSP

requirements would not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The substitution of TSP baskets for the NaOH addition equipment

would not cause any changes to the capability, settings, or

operation of the plant systems (other than the Iodine Removal System

itself) and would not, therefore, have any effect on the probability

of occurrence of an accident.

The substitution of TSP baskets for the NaOH addition equipment

has the effect of providing more immediate control of post-LOCA sump

pH, thereby increasing the assurance that iodine will remain in

solution throughout a postulated event. The consequences of

accidents evaluated in the FSAR [Final Safety Analysis Report] will

not be increased by this increased assurance.

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

from any previously evaluated.

The TSP baskets are passive components which have no interaction

with plant equipment unless flooding occurs in the containment. They

are designed and located such that they will not interact with any

plant safety equipment during a seismic event. The NaOH equipment,

which will be replaced by the TSP baskets, has no function or effect

on other equipment except during accident conditions. Therefore, the

substitution of TSP baskets for NaOH addition equipment cannot

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

any previously evaluated.

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

The substitution of TSP baskets for the NaOH addition equipment

would assure that the sump pH at the initiation of RAS

[recirculation actuation signal] is between 7.0 and 8.0 as assumed

in the MHA [maximum hypothetical accident] analysis. Therefore, this

change 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 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: Van Wylen Library, Hope

College, Holland, Michigan 49423.

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

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

NRC Project Director: John N. Hannon.

Duke Power Company, Docket Nos. 50-369 and 50-370, McGuire Nuclear

Station, Units 1 and 2, Mecklenburg County, North Carolina

Date of amendment request: December 7, 1994.

Description of amendment request: The amendments revise the

Technical Specification action statement to allow the Control Room Air

Intake to remain open when radiation monitors (EMF-43A and EMF-43B) are

inoperable. Immediate action to return the monitors to service would be

required.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed amendment would not involve a significant

increase in the probability or consequences of any accident

previously evaluated in the FSAR [Final Safety Analysis Report].

The amendment change will ensure correct Control Room

Ventilation system alignment in order to mitigate the consequence of

a Design Basis LOCA as described in FSAR Section 15.6.5.3,

Environmental Consequences of a Loss of-Coolant Accidents, Control

Room Operator Dose.

The amendment change will permit the intake to remain open and

will specify that action to repair the affected monitor shall be

taken immediately. The change itself is not considered to be an

initiator of any previously evaluated accident. Maintaining the VC

intake open with an inoperable monitor will not result in any

accidents that have not been previously evaluated. The

implementation of immediate actions to repair the inoperable monitor

does not in itself represent any accidents that have not been

previously evaluated. Therefore, the proposed Technical

Specification change does not increase the occurrence probability of

previously evaluated accidents.

The change to permit maintenance of open intakes will not

increase the consequences of any previously evaluated accidents. The

proposed amendment change is consistent with the original Safety

Analysis concerning the Dose to the Operators.

The analysis determined that the Doses to the Operators were

within acceptable ranges given the assumptions that the intakes

would [[Page 6300]] remain open and the contaminated air was

processed through a Safety Related filter train prior to

introduction into the Control Room. The proposed change remains

consistent with this analysis and does not change the assumptions or

methodology utilized to assess the Doses to the Operators for a

hypothesized DBA; therefore, the proposed amendment change will not

increase the consequences of any previously evaluated accident.

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

new or different kind of accident not previously evaluated.

The proposed change will not modify, delete, or add any systems

or components; therefore, no new failure modes or accidents

scenarios will be created.

No test or experiments will be revised; therefore, no new

initiating events or unanalyzed condition will be created.

Administrative changes to surveillance procedures will be minor and

will not create a safety concern.

3. No significant reduction in a margin of safety will occur.

The proposed amendment change requiring immediate action to

initiate repairs to an inoperable monitor does not impact existing

Safety Margins. Since requirements for immediate corrective action

does not currently exist within the Specification, the changes will

enhance the availability of the subject monitors.

The proposed amendment does not change/impact any assumption or

methods utilized to assess the doses to the operators for a

hypothetical worst case DBA. Accordingly, the proposed amendment

does not reduce any safety margins.

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: Atkins Library, University of

North Carolina, Charlotte (UNCC Station), North Carolina 28223.

Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422

South Church Street, Charlotte, North Carolina 28242.

NRC Project Director: Herbert N. Berkow.

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

Station, Unit 3, St. Charles Parish, Louisiana

Date of amendment request: December 9, 1994.

Description of amendment request: The proposed amendment would

revise the technical specifications (TSs) by revising the allowable

opening tolerances on the Pressurizer Code Safety Valves and the Main

Steam Line Code Safety Valves from plus or minus 1% to plus or minus

3%. This request is submitted as a result of an effort to improve valve

performance and to ensure that the TS limits are consistent with

expected valve performance capabilities.

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 any change to the

physical characteristics of the PSVs [pressurizer safety valves] and

MSSVs [main steam safety valves] and will have no impact on the PSVs

and MSSVs as-left setting. This change only allows for a larger

(plus or minus 3% versus plus or minus 1%) as-found setpoint

tolerance. Therefore, this change has no impact on the probability

of occurrence of any accident previously evaluated. The impact of

this change on the FSAR [final safety analyses report] analyses has

been evaluated and the results of the impacted events have been

found to be within the acceptable limits.

Therefore, revising the PSV and MSSV as-found opening setpoint

tolerance from plus or minus 1% to plus or minus 3% does not

increase the probability or consequences of an accident previously

evaluated.

2. The proposed changes to the PSVs and MSSVs as-found opening

setpoint tolerance do not modify equipment or change the manner in

which the plant will be operated. The safety valves will continue to

function per their design. Since no hardware modifications or

changes in operation procedures will be made, the proposed changes

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

accident from any accident previously evaluated.

3. The impact of the proposed changes on the Waterford 3 FSAR

analyses have been evaluated. The evaluation demonstrates that the

results of the impacted events remained within the acceptable

limits. The system capabilities to mitigate and/or prevent accidents

will be the same as they were prior to these changes. Therefore, the

proposed changes do not involve a reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room Location: University of New Orleans

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

Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn 1400 L

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

NRC Project Director: William D. Beckner.

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

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

Date of amendment request: December 9, 1994.

Description of amendment request: The proposed amendment would

revise the technical specifications (TSs) by revising a plant

protection system (PPS) trip setpoint and several allowable values such

that they will be consistent with the current setpoint/uncertainty

methodology being implemented at Waterford 3.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

Implementing the proposed change will not affect any design

basis accident. The revised Trip Setpoint and Allowable Values are

based upon the same Analytical Limits that form the basis for the

current Trip Setpoints and Allowable Values. The design basis for

each Trip Setpoint was verified to be consistent with the

appropriate accident analyses as part of the process of revising the

PPS setpoint analysis. The proposed change would implement a new

Trip Setpoint for the Reactor Coolant (RC) System Low Flow Reactor

trip and new Allowable Values for RC Low Flow, HI Log Power, HI

Steam Generator Water Level, HI Containment Pressure, Low

Pressurizer Pressure, Low Steam Generator Pressure, Low Steam

Generator Water Level, and Low RWSP [refueling water storage pool]

Level, based on the results of calculation EC-I92-019. The revised

Low RC Flow Trip Setpoint is based on the same analytical limit as

the current setpoint. The revised calculation uses the same design

inputs with a similarly based methodology to calculate a smaller

loop uncertainty. This results in a revised RC Low Flow Trip

Setpoint that retains the original analysis limit. Therefore, the

proposed change will not involve a significant increase in the

probability or consequences of any previously analyzed accident.

Plant operation and the manner in which the plant is operated

will not be altered as a result of implementing the proposed change

since no new system or design change is being implemented. The

proposed Setpoint and Allowable Value changes do not create any new

system interactions or interfaces. All information used to calculate

the new Trip Setpoint is consistent with that of the existing

accident analyses, and no new system interfaces/interactions are

created. Therefore, the proposed change will not create the

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

previously evaluated.

The proposed setpoint change revised the point at which the RCS

Low Flow reactor trip initiates a reactor trip. The Trip Setpoint is

based on the same Analytical Limit used to determine the current

setpoint. In addition, the same basic setpoint determination

[[Page 6301]] methodology is employed. That is, the Trip Setpoint is

the Analytical Limit plus or minus the Total Loop Uncertainty [TLU].

The Allowable Value is the Trip Setpoint plus or minus the Periodic

Test Error [PTE]. The change in the setpoint and allowable values

are [sic] due to a change in calculated TLU and PTE. The proposed

Trip Setpoint and Allowable Values are based on the same Analytical

Limits for the affected parameters and are determined using approved

methodology. Therefore, the proposed change will not involve a

significant reduction in margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room Location: University of New Orleans

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

Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn 1400 L

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

NRC Project Director: William D. Beckner.

Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric

Authority of Georgia, City of Dalton, Georgia, Docket Nos. 50-424 and

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

Georgia

Date of amendment request: December 27, 1994.

Description of amendment request: The proposed amendments would

revise the period for conducting leak testing of containment purge

valves to every 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 proposed change to the Technical Specifications does not

involve a significant increase in the probability or consequences of

an accident previously evaluated because the [test] results have

demonstrated that the resilient seat material does not degrade and

cause containment isolation valves to leak. Therefore the valves

will perform as assumed in the accident analyses.

2. The proposed change to the Technical Specifications does not

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

any accident previously evaluated because it does not require the

valves to function in any manner other than that which is currently

required.

3. The proposed addition to the Technical Specifications does

not involve a significant reduction in a margin of safety because it

only affects the frequency of the test and does not change the

leakage acceptance criteria. Since sufficient data has been

collected to demonstrate that the resilient seals do not degrade,

testing at the same frequency as other containment isolation valves

will not reduce the margin of safety provided by the Technical

Specifications.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Burke County Public Library,

412 Fourth Street, Waynesboro, Georgia 30830.

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

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

Georgia 30308.

NRC Project Director: Herbert N. Berkow.

Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric

Authority of Georgia, City of Dalton, Georgia, Docket Nos. 50-424 and

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

Georgia

Date of amendment request: December 29, 1994.

Description of amendment request: This request withdraws a similar

request dated January 22, 1993, as supplemented August 8, 1993, and

submits a new one in its place. The proposed amendments would revise

the Technical Specifications (TS) to add the automatic load sequencer

specification to TS Section 3/4.3, Instrumentation, and associated

Bases, and TS Section 3/4.8, Electrical Power Systems.

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 to the Technical Specifications does not

involve a significant increase in the probability or consequences of

an accident previously evaluated because the action to be taken when

an automatic load sequencer is inoperable is consistent with that of

a more stringent condition already specified, namely, the loss of an

entire train of emergency power during Modes 1-4, and for Modes 5

and 6 adding specific actions which previously had never been

addressed in TS.

2. The proposed change to the Technical Specifications does not

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

any accident previously evaluated because it does not involve any

change to the design, operation, or performance of the automatic

load sequencer. It only serves to clearly identify the appropriate

conservative response to an inoperable automatic load sequencer

applicable to the plant mode of operation.

3. The proposed change to the Technical Specifications does not

involve a significant reduction in a margin of safety because the

proposed actions to take when an automatic load sequencer is

inoperable [are] the same as the action already required by the

Technical Specifications when no power is available to the entire

emergency bus during Modes 1-4 and by adding requirements during

Modes 5 and 6, which had previously never been addressed.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Burke County Public Library,

412 Fourth Street, Waynesboro, Georgia 30830.

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

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

Georgia 30308.

NRC Project Director: Herbert N. Berkow.

Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric

Authority of Georgia, City of Dalton, Georgia, Docket Nos. 50-424 and

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

Georgia

Date of amendment request: January 3, 1995.

Description of amendment request: The proposed amendments would

revise the Technical Specifications (TS) with editorial changes to the

Action Statements of TS Sections 3.8.1.1 and 3.8.1.2 in order to

reflect the availability of a third offsite ac electrical source.

Surveillance Requirement 4.8.1.1.1 is being clarified to distinguish

that the offsite ac circuits which are connected to the onsite Class 1E

distribution system are required to be verified OPERABLE. The

amendments also modify the Technical Specifications with the addition

of a footnote to TS Section 3.8.3.1, to allow the connection of the

third offsite ac source to the onsite busses.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below: [[Page 6302]]

Based on the considerations regarding the addition of a footnote

for proper bus alignment during operating conditions, the licensee

submitted the following analysis in accordance with 10 CFR 50.92.

1. The proposed change to the Technical Specifications does not

involve a significant increase in the probability or consequences of

an accident previously evaluated because the probability of an LOSP

or an SBO is not increased by the allowance of having both redundant

emergency busses of 4160 volt switchgear connected to one offsite

source (RAT). The probability of having an LOSP is not increased

since the TS currently allow for a 72 hour LCO for one offsite power

source and the time the two redundant 4160 volt safety busses will

be temporarily aligned to one RAT is well within this time frame.

During this time the busses are interconnected, each bus is provided

adequate protection and separation by having separate and redundant

Class 1E circuit breakers, one per bus. The probability of an SBO is

not increased since neither bus' EDG will be affected during this

operation, and since this is a proceduralized manual alignment, the

interconnection to one RAT will not be initiated if either EDG were

inoperable. Also, the addition of the new ``swing'' offsite power

source (SAT), increases availability and flexibility of the VEGP

response to either an LOSP or SBO.

2. The proposed change to the Technical Specifications does not

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

any accident previously evaluated because the only postulated

adverse consequences of tying both redundant 4160 volt safety busses

together to one RAT is an LOSP. An LOSP is a design basis event

which has already been analyzed for VEGP. In response to an LOSP,

both EDGs remain capable of carrying the required loads to mitigate

the consequences of any postulated design basis accident during or

coincident with an LOSP.

3. The proposed addition to the Technical Specifications does

not involve a significant reduction in a margin of safety because

the only accident mitigating equipment and/or power sources which

will be unavailable during the transfer of offsite power sources is

the offsite power source being removed from service, allowed by

existing TS LCO 3.8.1.1(a). The 13.8 kV loads associated with the

RAT being removed from service and all of the 4160 volt non-Class 1E

loads fed from either RAT will be unavailable during this temporary

alignment. All of these loads are nonsafety related and therefore

are enveloped by the existing LOSP analysis.

Based on the considerations regarding clarification of SAT Use and

Expanded Bases, the licensee submitted the following analysis in

accordance with 10 CFR 50.92.

1. The proposed change to the TS does not involve a significant

increase in the probability or consequences of an accident

previously evaluated because only clarifications to existing TS

action statements and an additional expanded bases are being made.

No changes to the existing TS requirements for A.C. sources are

being made. The safety function of the offsite power source is

unchanged by the addition of the SAT and the probability of an LOSP

or SBO is not increased. In actuality, the addition of the SAT

increases the availability and flexibility of VEGP responses to

either an LOSP or SBO.

2. The proposed change to the TS does not create the possibility

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

evaluated because the loss of the SAT while being utilized to meet

TS offsite power source requirements is enveloped by existing LOSP

analysis.

3. The proposed change does not involve a significant reduction

in a margin of safety because although the SAT has no 13.8 kV

secondary winding, nor the same capacity as a RAT for accepting 4.16

kV non Class 1E loads, these loads are nonsafety related and

therefore enveloped by existing analysis. If a unit trip were to

occur while one 4.16 kV safety bus is being powered from the SAT,

the effect is a loss of the 13.8 kV and non Class 1E 4.16 kV loads

associated with the out of service RAT. This scenario is enveloped

by existing LOSP analysis.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Burke County Public Library,

412 Fourth Street, Waynesboro, Georgia 30830.

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

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

Georgia 30308

NRC Project Director: Herbert N. Berkow.

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald

C. Cook Nuclear Plant, Unit Nos. 1 and 2, Berrien County, Michigan

Date of amendment requests: August 12, 1992 and supplemented April

12, 1993.

Description of amendment requests: The proposed amendments would

change the minimum channels operable for the pressurizer safety valve

position indicator acoustic monitor to two out of three total from one

per valve. The amendments also delete footnotes which are no longer

applicable.

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:

We [the licensee] have evaluated the proposed T/Ss exemption and

have determined that it should not require a significant hazards

consideration based on the criteria established in 10CFR50.92(c).

Operation of the Cook Nuclear Plant in accordance with the proposed

amendment will not:

(1) Involve a significant increase in the probability or

consequences of an accident previously evaluated.

Although the proposed exemption results in the operator having

one less source of information on plant status, it does not create a

significant increase in the probability or consequences of an

accident previously evaluated. The acoustic monitors do not perform

a function vital to safe shutdown or to the isolation of the

reactor, or the reactor coolant system pressure boundary, nor is

there a mechanism involving an operable or inoperable pressurizer

safety valve acoustic monitor which would initiate an accident.

These monitors were added to meet the requirements of NUREG-0578 and

NUREG-0737. During normal operations, other instrumentation exists

that provides the operator with indication of safety valve

actuation. The acoustic monitors are not necessary to and are not

used in the emergency operating procedures. In addition, the

acoustic monitors being inoperable will not result in an

uncontrolled release of radiation to the environment and will not

initiate an accident. Finally, although the operator may have one

less channel operable, the operator receives no less information

than if all three channels are operable because one valve opening

causes all operable channels to actuate. Therefore, we conclude that

the proposed T/Ss changes do not involve a significant increase in

the probability or consequences of an accident previously evaluated.

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

accident from any previously analyzed.

As previously stated, the purpose of the acoustic monitor is to

provide the operator with information regarding safety valve

position that may assist in the mitigation of the consequences of an

accident. Specifically, it provides information that a safety valve

has lifted. However, the operator has other mechanisms for obtaining

equivalent information. In addition, the signals generated by an

acoustic monitor do not initiate any other equipment actuation, nor

will the inoperability of an acoustic monitor initiate any accident.

Consequently, the proposed T/Ss changes do not create the

possibility of a new or different kind of accident from any

previously analyzed.

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

The proposed T/Ss changes result in the operator potentially

having one less source of information on plant status. However, we

believe the margin of safety is not reduced for several reasons.

First, the operator is provided with other viable flow detection

devices to determine pressurizer safety valve position, i.e., the

temperature sensor on the discharge line associated with the

inoperable acoustic monitor, and pressurizer relief tank level (NLA-

351), temperature (NTA-351) [[Page 6303]] and pressure (NPA-351)

indications. Also, the acoustic monitors are not used by the

operators in an emergency situation, as the operator relies on other

indications of loss of reactor coolant inventory per the emergency

operating procedures. In addition, previous experience with the

pressurizer safety valve position indicator acoustic monitoring

system has shown that, when any one of the pressurizer safety valves

opens, all three safety valve position indicator acoustic monitors

are actuated. Because of this, the operator receives no less

information regardless if only two or three channels are operable.

Based on the above, we believe that having an acoustic monitor

inoperable does not warrant reactor and plant shutdown. As the T/Ss

are currently stated, should one pressurizer safety valve position

indicator acoustic monitor become inoperable, it must be restored to

operable status within thirty days or the unit must be in hot

shutdown within the subsequent twelve hours. Thermal cycling from

unwarranted plant shutdowns increases the likelihood of reactor

vessel embrittlement and unnecessarily challenges the safety

systems. Because a signal from the pressurizer safety valve position

indicator acoustic monitors is not necessary nor used to ensure the

safe shutdown of the unit even if a pressurizer safety valve is

opened or stuck open during an emergency situation, we believe that

a plant shutdown due to an inoperable acoustic monitor would be

unwarranted.

We believe that the unit can be operated safely and that we

would still meet the intent of NUREG-0538 and NUREG-0737 with only

two out of three pressurizer safety valve position indicator

acoustic monitors operable.

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 requests involve no significant hazards consideration.

Local Public Document Room location: Maud Preston Palenske Memorial

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

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

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

NRC Project Director: John N. Hannon.

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

Atomic Power Station, Lincoln County, Maine

Date of amendment request: November 18, 1994.

Description of amendment request: The proposed amendment would

change the title of certain Plant Operation Review Committee (PORC)

members to reflect recent Maine Yankee organizational changes; update

training requirements to comply with 10 CFR 50.120, Training and

qualification of nuclear power plant personnel; and reporting frequency

requirements for the Radioactive Effluent Release and Estimated Dose

and Meteorological Summary Reports.

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. The NRC staff has reviewed the licensee's analysis

against the standards of 10 CFR 50.92(c). A summary of the licensee's

analysis is presented below:

1. The proposed amendment would not involve a significant increase

in the probability or consequences of an accident previously evaluated.

The changes proposed by this amendment request are administrative

in nature. Because the proposed changes do not involve any physical

alterations to plant equipment, operating setpoints, parameters or

conditions, the plant's response to previously evaluated accidents is

not affected.

The licensee therefore concludes that implementation of the

proposed change will not involve a significant increase in the

probability or consequences of an accident previously evaluated.

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

or different kind of accident from any accident previously evaluated.

The administrative nature of the proposed changes does not affect

the design, operation, maintenance or testing of the plant. Thus, no

new modes of failure are created.

The licensee therefore concludes that implementation of the

proposed change will not create the possibility of a new or different

kind of accident from any accident previously evaluated.

3. The proposed amendment would not involve a significant reduction

in a margin of safety.

The proposed change reflects an organizational change that does not

modify the qualification requirements or competence of the members of

the PORC. Thus, the capability of PORC to meet its responsibilities in

accordance with the plant Technical Specifications is unchanged.

Deleting the current training requirement for Shift Technical

Advisors eliminates duplicative training requirements and represents

conformance to 10 CFR 50.120, Training and qualification of nuclear

power plant personnel.

Elevating the responsibility for training the plant staff from the

Manager, Operations Department, to the Vice President of Operations,

does not represent a reduction in a margin of safety.

The proposed change to the Radioactive Effluent Release and

Estimated Dose and Meteorological Summary Reports is related to the

submittal schedule for statistical data and is administrative in

nature. The change in submittal frequency provides consistency between

the various required reports and also is administrative in nature.

The licensee therefore concludes that implementation of the

proposed change would not involve a significant reduction in a margin

of safety.

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: Wiscasset Public Library, High

Street, P.O. Box 367, Wiscasset, Maine 04578.

Attorney for licensee: Mary Ann Lynch, Esquire, Maine Yankee Atomic

Power Company, 329 Bath Road, Brunswick, Maine 04011.

NRC Project Director: Walter R. Butler.

Northeast Nuclear Energy Company, et al., Docket No. 50-336, Millstone

Nuclear Power Station, Unit No. 2, New London County, Connecticut

Date of amendment request: December 16, 1994.

Description of amendment request: The proposed change to the

Technical Specifications would require the wind direction and wind

speed sensors at the 142 foot elevation to identify the data to

determine action required to preclude flood damage to the Service Water

Pumps. Also, the proposed change would correct a typographical error in

the location of the sensors at the 374 foot elevation.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

* * * The proposed changes do not involve a significant hazards

consideration because the changes would not:

1. Involve a significant increase in the probability or

consequences of an accident previously analyzed.

NNECO [Northeast Nuclear Energy Company] is proposing to revise

LCOs [Limiting Conditions for Operation] 3.7.5.1.b.3 and 3.7.5.1.b.4

and Table 3.3-8 of the Millstone Unit No. 2 Technical

[[Page 6304]] Specifications by changing the elevation that the

average wind speed and average wind direction are measured and by

correcting a typographical error, respectively. The proposed changes

have no effect on any of the accidents analyzed in Chapter 14 of the

Millstone Unit No. 2 FSAR [Final Safety Analysis Report]. Site

flooding is considered in Section 2.5.4.2.1 of the FSAR. Utilizing

the wind speed indicator at the 142-foot elevation, in lieu of the

indicator on the 374-foot elevation will not significantly change

the ability of personnel to predict the potential for a major storm

with flooding.

The proposed changes do not alter the intent of the

surveillances, do not involve any physical changes to the plant, do

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

do not modify the manner in which the plant is operated.

Based on the above, the proposed changes do not involve a

significant increase in the probability or consequences of an

accident previously analyzed.

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

from any previously analyzed.

NNECO is proposing to revise LCOs 3.7.5.1.b.3 and 3.7.5.1.b.4

and Table 3.3-8 of the Millstone Unit No. 2 Technical Specifications

by changing the elevation that the average wind speed and average

wind direction are measured and by correcting a typographical error,

respectively. The proposed changes do not alter the intent of the

surveillances, do not involve any physical changes to the plant, do

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

do not modify the manner in which the plant is operated.

While the proposed changes to LCOs 3.7.5.1.b.3 and 3.7.5.1.b.4

do change the measurement location stipulated by the technical

specifications, this change is insignificant. Utilizing the wind

speed indicator at the 142-foot elevation, in lieu of the indicator

on the 374-foot elevation will not significantly change the ability

of personnel to predict the potential for a major storm with

flooding.

Based on the above, the proposed changes do not create the

possibility of a new or different kind of accident from any

previously analyzed.

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

NNECO is proposing to revise LCOs 3.7.5.1.b.3 and 3.7.5.1.b.4

and Table 3.3-8 of the Millstone Unit No. 2 Technical Specifications

by changing the elevation that the average wind speed and average

wind direction are measured and by correcting a typographical error,

respectively. The proposed changes will have no impact on the

physical protective boundaries (fuel matrix/cladding, reactor

coolant system pressure boundary, and containment). The proposed

changes do not alter the intent of the surveillances, do not involve

any physical changes to the plant, do not alter the way any

structure, system, or component functions, and do not modify the

manner in which the plant is operated.

While the proposed changes to LCOs 3.7.5.1.b.3 and 3.7.5.1.b.4

do change the manner in which potential flooding is predicted, this

change is insignificant. Utilizing the wind speed and direction

indicators at the 142-foot elevation, in lieu of the indicators at

the 374-foot elevation will not significantly change the ability of

personnel to predict the potential for a major storm with flooding.

Based on the above, the proposed changes do not involve a

significant reduction in a margin of safety.

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

this review, it appears that the three standards of 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 Resource Center,

Three Rivers Community-Technical College, Thames Valley Campus, 574 New

London Turnpike, Norwich, CT 06360.

Attorney for licensee: Ms. L. M. Cuoco, Senior Nuclear Counsel,

Northeast Utilities Service Company, Post Office Box 270, Hartford, CT

06141-0270.

NRC Project Director: Phillip F. McKee.

Northeast Nuclear Energy Company, et al., Docket No. 50-336, Millstone

Nuclear Power Station, Unit No. 2, New London County, Connecticut

Date of amendment request: December 21, 1994.

Description of amendment request: Proposed revision to License

Condition and Technical Specifications to relocate the Fire Protection

Requirements from the Technical Specifications to another controlled

document, the technical requirements manual (TRM).

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

* * * The proposed changes do not involve a significant hazards

consideration because the changes would not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed changes relocates the provisions of the Fire

Protection Program that are contained in the Technical

Specifications and places them in the TRM. No current requirements

are being added or deleted aside from removal of the special reports

section. Review of the Fire Protection Program and its revisions

will be the responsibility of the PORC [Plant Operations Review

Committee] and SORC [Station Operations Review Committee], just as

it has always been the responsibility of these groups to review

changes to the fire protection Limiting Condition for Operation and

Surveillance Requirements when they were part of the Technical

Specifications. In addition, no design basis accidents are affected

by this change, nor are safety systems adversely affected by the

changes. Therefore, there is no impact on the probability of

occurrence or the consequences of any design basis accidents.

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

from any previously evaluated.

The proposed changes relocate the provisions of the Fire

Protection Program that are contained in the Technical

Specifications and places them in the TRM. No current requirements

are being added or deleted aside from removal of the special report

section. There are no new failure modes associated with the proposed

changes. Since the plant will continue to operate as designed, the

proposed changes will not modify the plant response to the point

where it can be considered a new accident.

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

No change is being proposed for the Fire Protection Program

requirements themselves. The relevant Technical Specifications are

being relocated, and the requirements contained therein are being

incorporated into the TRM. Plant procedures will continue to provide

the specific instructions necessary for the implementation of the

requirements, just as when the requirements resided in the Technical

Specifications. Fire Protection Program changes will be governed by

the provisions of 10 CFR 50.59 and the current fire protection

license condition. As such, the changes do not directly affect any

protective boundaries nor does it impact the safety limits for the

boundary. Thus, there are no adverse impacts on the protective

boundaries, safety limits, or 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: Learning Resource Center,

Three Rivers Community-Technical College, Thames Valley Campus, 574 New

London Turnpike, Norwich, CT 06360.

Attorney for licensee: Ms. L. M. Cuoco, Senior Nuclear Counsel,

Northeast Utilities Service Company, Post Office Box 270, Hartford, CT

06141-0270.

NRC Project Director: Phillip F. McKee.

Northeast Nuclear Energy Company, et al., Docket No. 50-423, Millstone

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

Date of amendment request: December 2, 1994.

Description of amendment request: The proposed amendment modifies

the [[Page 6305]] surveillance requirements for the power range neutron

flux instrumentation to permit entering reactor operating modes 1 and 2

to perform necessary test for power range detectors.

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 (SHC), which is presented below:

* * * The proposed changes do not involve an SHC because the

changes would not:

1. Involve a significant increase in the probability or

consequence of an accident previously analyzed.

NNECO is proposing to modify Table 4.3-1 by adding Note 5 to

Functional Units 2b, 3, and 4. This note provides an exception from

the provisions of Technical Specification 4.0.4. Entry into Mode 2

or Mode 1, as appropriate, would allow for appropriate test

conditions to complete the channel calibration of power range

neutron detectors (i.e., Functional Units 2b, 3, and 4 of Table 4.3-

1). This will improve plant safety by performing tests at proper

conditions. The acceptance criteria, such as response times, test

frequency, or test methods, are not revised. Therefore, the power

range neutron detectors will perform their intended function when

called upon. Additionally, the proposed changes are consistent with

the new, improved STS for the Westinghouse plants (NUREG-1431).

Based on the above, the proposed changes to Functional Units 2b,

3, and 4 of Table 4.3-1 of the Millstone Unit No. 3 Technical

Specifications do not involve a significant increase in the

probability or consequences of an accident previously analyzed.

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

from any previously analyzed.

The proposed changes to Functional Units 2b, 3, and 4 of Table

4.3-1 do not make any physical or operational changes to existing

plant structures, systems, or components. The proposed changes do

not introduce any new failure mode. They simply allow tests to be

performed at appropriate conditions (e.g., Mode 2 or Mode 1 rather

than Mode 4 or Mode 3).

Additionally, the proposed changes do not modify the acceptance

criteria for the tests. The purpose of the tests is to ensure that

the power range neutron detectors can perform their intended

function.

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

new or different kind of accident from any previously analyzed.

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

The proposed changes to Functional Units 2b, 3, and 4 of Table

4.3-1 do not have any adverse impact on the design basis accident

analyses. The applicable acceptance criteria for the power range

neutron detectors will not be modified by the proposed changes. The

proposed changes will permit the tests to be conducted under the

proper conditions, so that the ability of the power range neutron

detectors to perform their intended safety function can be

confirmed.

Based on the above, there is no 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: Learning Resource Center,

Three Rivers Community-Technical College, Thames Valley Campus, 574 New

London Turnpike, Norwich, CT 06360.

Attorney for licensee: Ms. L. M. Cuoco, Senior Nuclear Counsel,

Northeast Utilities Service Company, Post Office Box 270, Hartford, CT

06141-0270.

NRC Project Director: Phillip F. McKee.

Northern States Power Company, Docket Nos. 50-282 and 50-306, Prairie

Island Nuclear Generating Plant, Unit Nos. 1 and 2, Goodhue County,

Minnesota

Date of amendment requests: August 30, 1994.

Description of amendment requests: The proposed amendments would

revise the Technical Specifications (TS) for Prairie Island Nuclear

Generating Plant as recommended by Generic Letter (GL) 93-05, ``Line-

Item Technical Specification Improvements to Reduce Surveillance

Requirements for Testing During Power Operation.'' The proposed

amendments would also revise testing and calibration requirements

associated with the containment hydrogen recombiners. The proposed TS

changes are as follows:

(1) TS Table 4.1-1C, ``Miscellaneous Instrumentation Surveillance

Requirements.'' Delete Item 14, ``Accumulator Level and Pressure'' and

corresponding frequency interval designations.

(2) TS Table 4.1-2A, ``Minimum Frequencies For Equipment Tests,''

Item 2. Revise the frequency for partial movement of all control rod

assemblies from every 2 weeks to once per quarter.

(3) TS 4.3, ``Primary Coolant System Pressure Isolation Values.''

Under Specification heading, extend the amount of time the plant can be

shut down before pressure isolation valve testing will be required from

72 hours to 7 days.

(4) TS SR 4.4.I, 4.4.I.a, 4.4.I.b, 4.4.I.b.1, 4.4.I.b.2, and

4.4.I.b.3, ``Electrical Hydrogen Recombiners.'' Revise the containment

hydrogen recombiner testing surveillance frequency from every 6 months

to every refueling interval. Delete the specific requirement to perform

CHANNEL CALIBRATION of recombiner instruments and control circuits.

Delete the requirement to sequentially perform the resistance to ground

test following the functional test.

(5) TS SR 4.5.A.2.b, ``Containment Spray System.'' Revise the

containment spray system nozzle testing surveillance frequency from

once every 5 years to once every 10 years.

(6) TS SR 4.8.A.1, 4.8.A.2, and Footnote, ``Auxiliary Feedwater

System.'' Revise the testing frequency for the auxiliary feedwater

pumps from intervals of 1 month to semi-quarterly on a staggered test

basis.

(7) BASES 4.8, ``Steam And Power Conversion Systems.'' Revise the

Bases to include testing frequency for the auxiliary feedwater pumps

from intervals of 1 month to semi-quarterly on a staggered test basis.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed amendment[s] will not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

Except for hydrogen recombiner changes to conform to Standard

Technical Specifications, the requested changes were extensively

reviewed by the NRC during the preparation of NUREG-1366 and Generic

Letter 93-05. For the sake of clarity each proposed change is

discussed separately in the order appearing in the Prairie Island

Technical Specifications.

A. This Technical Specification amendment removes the

accumulator water level and pressure channel surveillance from the

Technical Specifications and places them into a licensee controlled

test procedure. These changes are consistent with industry

recognition that accumulator instrumentation operability is not

directly related to the capability of the accumulators to perform

their safety function.

Relocating the instrumentation surveillance requirements is an

administrative change which will not affect equipment testing,

availability, or operation. Therefore, it will not have an effect on

the probability or consequences of an accident.

B. This Technical Specification amendment changes control rod

movement from every two weeks to once every quarter. Control rod

movement testing is performed to determine if the control rods are

immovable. Control rods may be electrically stuck due to a problem

in the control rod drive circuitry or mechanically stuck. Electrical

problems with the control rod drive system, in general, do not

prevent insertion of a control rod into [[Page 6306]] the core when

the reactor trip breakers are opened.

NUREG-1366 determined that control rod movement testing is not

effective in determining immovable control rods. Most of the

mechanically immovable control rods are discovered during plant

startup during initial pulling of the rods or during rod drop

testing. Extending the surveillance interval will not affect this

failure discovery method.

The accident analyses assume that the single highest worth rod

is struck while fully withdrawn and will not insert. One immovable

control rod will still bound this accident analysis. For these

reasons, the extension of the surveillance frequency from once every

two weeks to once every quarter will not involve a significant

increase in the probability or consequences of a previously

evaluated accident.

C. This Technical Specification amendment will require Reactor

Coolant Systems Pressure Isolation Valves (PIV) to be surveillance

tested after seven days at cold shutdown instead of after three days

at cold shutdown.

The PIVs are important in preventing over pressurization and

rupture of the Emergency Core Cooling System low pressure piping

which could result in a LOCA [loss-of-coolant accident] that

bypasses containment. Allowable leakage from any PIV is sufficiently

low to ensure early detection of possible in-series check valve

failure. This change will not change the refueling outage

surveillance, nor will it change the required testing to be

performed after maintenance, repair, or replacement. The proposed

level of surveillance is appropriate for these valves.

These valves have had very good operating performance and should

continue to have the same performance record with continuation of

the same maintenance and testing program. Furthermore, these valves

are backed by motor or air-operated valves which have performed

reliably.

For these reasons, the extension of the amount of time from

three days to seven days before pressure isolation valve testing is

required will not result in a significant increase in the

probability or consequences of a previously evaluated accident.

D. This Technical Specification amendment will revise the

containment hydrogen recombiner testing surveillance from every six

months to every refueling interval.

The two independent containment hydrogen recombiners provide

post-accident hydrogen control of the containment atmosphere. The

recombiners are designed to be passive until an accident occurs.

Industry experience and in particular, Prairie Island experience

has demonstrated that this equipment is highly reliable. Since the

recombiners are not required until after an accident, there would

likely be time to effect accessible repairs if the equipment were

not operable.

Relocation of the recombiner calibration is an administrative

change which will not affect recombiner operability. Deletion of

specific testing sequence will not affect the performance of

recombiner testing.

Equipment redundancy, reliability and time for repairs ensures

post-accident control. For these reasons, these changes will not

result in a significant increase in the probability or consequences

of a previously evaluated accident.

E. This Technical Specification amendment will revise the

containment spray system nozzle testing surveillance from once every

five years to once every ten years.

Two independent containment spray systems provide post-accident

cooling of the containment atmosphere and provide a mechanism for

removing iodine from the containment atmosphere. This surveillance

test verifies by air flow test that the spray nozzles are

unobstructed. The extension of the surveillance frequency does not

affect administrative controls that preclude entry of foreign

material into the nozzles.

At Prairie Island the piping headers and nozzles are fabricated

from austenitic stainless steel. There have been no reported in-

service problems noted with spray nozzle testing from plants with

stainless steel headers and nozzles and there is no indication that

the lines would corrode and become obstructed.

For these reasons, this change will not result in a significant

increase in the probability or consequences of a previously

evaluated accident.

F. This Technical Specification amendment will revise the

frequency for testing the Auxiliary Feedwater Pumps (AFWP) from

monthly to semi-quarterly on a STAGGERED TEST BASIS.

Two 100% redundant, diverse pumps provide an emergency source of

feedwater to the steam generators. The Prairie Island AFWPs have

performed reliably. However, frequent testing of the pumps and

associated equipment wears out the equipment resulting in equipment

unavailability. AFWP availability will be increased by semi-

quarterly surveillance testing on a STAGGERED TEST BASIS.

For these reasons, this change will not result in a significant

increase in the probability or consequences of previously evaluated

accident.

Therefore, the probability or consequences of an accident

previously evaluated are not affected by any of the proposed

amendments.

2. The proposed amendment[s] will not create the possibility of

a new or different kind of accident from any accident previously

analyzed.

The extension of facility surveillance intervals as discussed

previously will not result in changes in plant configuration or

operation. The changes in recombiner calibration and testing will

not result in changes in plant configuration or operation.

Therefore, the possibility of a new or different kind of accident

from any accident previously evaluated would not be created.

3. The proposed amendment[s] will not involve a significant

reduction in the margin of safety.

The amendments proposed in this License Amendment Request do not

reduce the ability of any system or component to perform its safety

related function. The basis of NUREG-1366, Generic Letter 93-05, and

the analysis performed in support of this License Amendment Request

is that the reduction in surveillance testing can improve safety by

reducing challenges to plant systems, personnel exposure, and

equipment wear or degradation. The proposed changes to surveillance

frequencies do not change the method of performing any surveillance.

The operation of systems and equipment remains unchanged. Therefore,

a significant reduction in the margin of safety would not be

involved.

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 requests involve 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: Jay Silberg, Esq., Shaw, Pittman, Potts, and

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

NRC Project Director: John N. Hannon.

Northern States Power Company, Docket Nos. 50-282 and 50-306, Prairie

Island Nuclear Generating Plant, Unit Nos. 1 and 2, Goodhue County,

Minnesota

Date of amendment requests: December 5, 1994.

Description of amendment requests: The proposed amendments would

revise Technical Specification 3.8 to allow containment airlock doors

to remain open during core alterations provided certain conditions are

met. This request is similar to the amendment for Calvert Cliffs

Nuclear Power Plant which the NRC approved on August 30, 1994. In

addition, these amendments would allow containment penetrations to

remain open during core alterations provided certain conditions are

met.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed amendment will not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

The proposed containment refueling integrity amendments do not

affect the probability of a fuel handling accident, they only deal

with the containment systems.

The containment is provided for the purpose of mitigating the

consequences of postulated accidents. For the fuel handling accident

in containment, the licensing basis analyses, including the NRC

safety [[Page 6307]] evaluation report transmitted February 2, 1982,

assumed that containment was completely abrogated and all

radioactive materials released from the containment refueling pool

are assumed to be released to the outside atmosphere. The requested

amendments to Technical Specification 3.8.A.1.a modify the use of

containment to mitigate the consequences of a fuel handling accident

in containment, however, since instantaneous offsite release of all

fuel handling accident materials released to containment has already

been considered, the probability and consequences of a loss of

containment accident are not increased.

Therefore, the probability or consequences of an accident

previously evaluated are not affected by any of the proposed

amendments.

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

new or different kind of accident from any accident previously

analyzed.

The requested amendments to Technical Specification 3.8.A.1.a

modify the use of containment to mitigate the consequences of a fuel

handling accident in containment. There are no new failure modes or

mechanisms associated with the proposed changes, nor do the proposed

changes involve any modification of plant equipment or changes in

plant operational limits. Previous analyses, including the NRC fuel

handling accident safety evaluation for Prairie Island, have already

assumed the containment is abrogated. The proposed license

amendments may affect the release path for fission products released

during a fuel handling accident in containment, but no new or

different kind of accident will result.

Therefore, the possibility of a new or different kind of

accident from any accident previously evaluated would not be

created.

3. The proposed amendment will not involve a significant

reduction in the margin of safety

The margin of safety as defined by the licensing bases fuel

handling accident analyses is not reduced. The previous analyses are

very conservative, assuming all radioactive material released from

[containment] by the fuel handling accident is immediately released

to the outside atmosphere, and bound any changes introduced by these

requested amendments.

Technical Specification 3.8.A.1.a exists to minimize the

consequences of a fuel handling accident in containment. However,

with the current Technical Specification 3.8.A.1.a, there will still

be releases due to the necessity to open the containment airlocks to

evacuate personnel. With implementation of this amendment, the

ability of the closed airlocks to contain the accident releases may

improve.

Some radioactive material could be released through containment

penetrations that are open at the time of the accident. Since it is

not likely that containment will be pressurized by a fuel handling

accident, the releases are expected to be minimal. This amendment

will maintain containment post-fuel handling accident offsite

releases well within the limits of 10CFR100 and the current license

basis releases.

Therefore, a significant reduction in the margin of safety would

not be involved.

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 requests involve 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: Jay Silberg, Esq., Shaw, Pittman, Potts, and

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

NRC Project Director: John N. Hannon.

Northern States Power Company, Docket Nos. 50-282 and 50-306, Prairie

Island Nuclear Generating Plant, Unit Nos. 1 and 2, Goodhue County,

Minnesota

Date of amendment requests: January 9, 1995.

Description of amendment requests: The proposed amendments would

revise Prairie Island Nuclear Generating Plant Technical Specification

(TS) 4.12, ``Steam Generator Tube Surveillance,'' to incorporate

revised acceptance criteria for steam generator tubes with degradation

in the tubesheet roll expansion region. These criteria for steam

generator tube acceptance were developed by Westinghouse Electric

Corporation and are known as F* (``F-Star'') and L* (``L-Star''). These

criteria would be utilized to avoid unnecessary plugging and sleeving

of steam generator tubes.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed amendment[s] will not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

The supporting technical and safety evaluations of the subject

criterion demonstrate that the presence of the tubesheet will

enhance the tube integrity in the region of the hardroll by

precluding tube deformation beyond its initial expanded outside

diameter. The resistance to both tube rupture and tube collapse is

strengthened by the presence of the tubesheet in that region. The

results of hardrolling of the tube into the tubesheet is an

interference fit between the tube and the tubesheet. Tube rupture

cannot occur because the contact between the tube and tubesheet does

not permit sufficient movement of tube material. The radial preload

developed by the rolling process will secure a postulated separated

tube end within the tubesheet during all plant conditions. In a

similar manner, the tubesheet does not permit sufficient movement of

tube material to permit buckling collapse of the tube during

postulated LOCA loadings.

The F* length of roll expansion is sufficient to preclude tube

pullout from tube degradation located below the F* distance,

regardless of the extent of the tube degradation. The existing

Technical Specification leakage rate requirements and accident

analysis assumptions remain unchanged in the unlikely event that

significant leakage from this region does occur. As noted above,

tube rupture and pullout is not expected for tubes using the F*

criterion. Any leakage out of the tube from within the tubesheet at

any elevation in the tubesheet is fully bounded by the existing

steam generator tube rupture analysis included in the Prairie Island

Plant USAR [Updated Safety Analysis Report]. For plants with partial

depth roll expansion like Prairie Island, a postulated tube

separation within the tube near the top of the roll expansion (with

subsequent limited tube axial displacement) would not be expected to

result in coolant release rates equal to those assumed in the USAR

for a steam generator tube rupture event due to the limited gap

between the tube and tubesheet. The proposed plugging criterion does

not adversely impact any other previously evaluated design basis

accident.

Leakage testing of roll expanded tubes indicates that for roll

lengths approximately equal to the F* distance, any postulated

faulted condition primary to secondary leakage from F* tubes would

be insignificant.

2. The proposed amendment[s] will not create the possibility of

a new or different kind of accident from any accident previously

analyzed.

Implementation of the proposed F* criterion does not introduce

any significant changes to the plant design basis. Use of the

criterion does not provide a mechanism to initiate an accident

outside of the region of the expanded portion of the tube. Any

hypothetical accident as a result of any tube degradation in the

expanded portion of the tube would be bounded by the existing tube

rupture accident analysis. Tube bundle structural integrity will be

maintained. Tube bundle leaktightness will be maintained such that

any postulated accident leakage from F* tubes will be negligible

with regards to offsite doses.

3. The proposed amendment[s] will not involve a significant

reduction in the margin of safety.

The use of the F* criterion has been demonstrated to maintain

the integrity of the tube bundle commensurate with the requirements

of Reg Guide 1.121 [``Bases for Plugging Degraded PWR Steam

Generator Tubes''] (intended for indications in the free span of

tubes) and the primary to secondary pressure boundary under normal

and postulated accident conditions. Acceptable tube degradation for

the F* criterion is any degradation indication in the tubesheet

region, more than the F* distance below the bottom of the transition

between the roll [[Page 6308]] expansion and the unexpanded tube.

The safety factors used in the verification of the strength of the

degraded tube are consistent with the safety factors in the ASME

Boiler and Pressure Vessel Code used in steam generator design. The

F* distance has been verified by testing to be greater than the

length of roll expansion required to preclude both tube pullout and

significant leakage during normal and postulated accident

conditions. Resistance to tube pullout is based upon the primary to

secondary pressure differential as it acts on the surface area of

the tube, which includes the tube wall cross-section, in addition to

the inner diameter based area of the tube. The leak testing

acceptance criteria are based on the primary to secondary leakage

limit in the Technical Specifications and the leakage assumptions

used in the USAR accident analysis.

Implementation of the tubesheet plugging criterion will decrease

the number of tubes which must be taken out of service with tube

plugs or repaired with sleeves. Both plugs and sleeves reduce the

RCS (reactor coolant system) flow margin; thus, implementation of

the F* criterion will maintain the margin of flow that would

otherwise be reduced in the event of increased plugging or sleeving.

Based on the above, it is concluded that the proposed change

does not result in a significant reduction in margin with respect to

plant safety as defined in the USAR or the Technical Specification

Bases.

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 requests involve 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: Jay Silberg, Esq., Shaw, Pittman, Potts, and

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

NRC Project Director: John N. Hannon.

Northern States Power Company, Docket Nos. 50-282 and 50-306, Prairie

Island Nuclear Generating Plant, Unit Nos. 1 and 2, Goodhue County,

Minnesota

Date of amendment requests: January 13, 1995.

Description of amendment requests: The proposed amendments would

revise Prairie Island Nuclear Generating Plant Technical Specification

4.4.D.1 to change the interval for the performance of the Residual Heat

Removal (RHR) System leakage test from once every 12 months to perform

the test during each refueling shutdown.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed amendment[s] will not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

The proposed changes to the RHR system leakage test interval

only involve the leak-tightness of the RHR system for postaccident

operation. As such, the proposed changes will have no impact on the

probability of an accident previously evaluated.

The extension of the RHR system leakage test interval could

increase the possibility of undetected RHR system leakage outside

the containment during post accident conditions. However, the

possible consequences of leakage from the RHR system outside

containment are minor relative to those of the design basis

accident. Therefore, because leakage from the RHR system has a minor

effect on offsite dose, and since previous testing on a 12 month

interval has not found significant RHR system leakage, the extension

of the test interval to refueling is not expected to significantly

impact the offsite dose consequences of an accident. In addition, it

is probable that RHR system leakage would be identified during the

normal quarterly functional testing and inspection of the RHR

system.

Therefore, for the reasons discussed above, the proposed changes

will not significantly affect the probability or consequences of an

accident previously evaluated.

2. The proposed amendment[s] will not create the possibility of

a new or different kind of accident from any accident previously

analyzed.

There are no new failure modes or mechanisms associated with the

proposed changes. The proposed changes do not involve any

modification of the plant equipment or any changes in operational

limits. The proposed changes only modify the interval for the

performance of the RHR system leakage test. The performance of the

RHR system leakage test on a refueling basis instead of every 12

months cannot create a new or different kind of accident.

Therefore, for the reasons discussed above, the proposed changes

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

from any previously evaluated, and the accident analyses presented

in the Updated Safety Analysis Report [USAR] will remain bounding.

3. The proposed amendment[s] will not involve a significant

reduction in the margin of safety.

The performance of the RHR system leakage test at power is more

complex than performing the test during refueling shutdown. It is

preferable, from an RHR system reliability and plant safety

standpoint, to perform the test during refueling shutdown when the

RHR system is already operating and when no changes to the RHR

system configuration are required. Any possible increase in the risk

to the public health and safety incurred by extending the RHR leak

test interval from 12 months to refueling shutdown will be off-set

by the reduction in risk obtained by not performing the RHR system

leakage test during power operation.

The extension of the test interval would mean that possible RHR

leakage could exist undetected for a longer period than allowed by

the current Technical Specifications. However, the possible

consequences of leakage from the RHR system outside containment are

minor relative to those of the design basis accident. In addition,

it is probable that RHR system leakage would be identified during

the normal quarterly functional testing and inspection of the RHR

system.

Based on the above, it is concluded that the proposed change

does not result in a significant reduction in margin with respect to

plant safety as defined in the USAR or the Technical Specification

Bases.

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 requests involve 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: Jay Silberg, Esq., Shaw, Pittman, Potts, and

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

NRC Project Director: John N. Hannon.

Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-311,

Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem County, New

Jersey

Date of amendment request: December 23, 1994.

Description of amendment request: The proposed amendment to the

Technical Specifications revises the surveillance requirement to

perform a visual inspection of containment areas affected by

containment entry when containment integrity is established. It is

consistent with Item 7.5 of Generic Letter 93-05, ``Line-Item Technical

Specifications Improvements to Reduce Surveillance Requirements for

Testing During Power Operation.''

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 not involve a significant increase in the probability or

consequences of an accident previously evaluated. [[Page 6309]]

The proposed change does not alter the assumptions, design

parameters or results of Updated Final Safety Analysis Report

(UFSAR) accidents analyzed. The proposed change does not involve a

hardware change, a change to the operation of any systems or

components, or a change to any existing structures. The proposed

change leads to a reduction in radiation exposure to plant personnel

and the elimination of an unnecessary burden on plant staff. The

revised visual inspection practice will not increase the probability

or consequences of an accident previously evaluated.

2. Does not create the possibility of a new or different kind of

accident from any previously evaluated.

The proposed change does not modify equipment, affect system

design bases or operability. This change does not alter parameters

utilized in the analyzed accident scenarios. The proposed change in

surveillance frequency is consistent with the guidance provided in

GL 93-05. The performance of a visual inspection of containment

areas affected by multiple containment entries on a daily bases

[basis] and at the completion of the final entry when containment

integrity is established will not create the possibility of a new or

different kind of accident from those previously evaluated.

3. Does not involve a significant reduction in a margin of

safety.

The proposed change only involves a decrease in surveillance

frequency when multiple entries are made in a single day and does

not alter the performance of the surveillance itself. System

equipment and operation remains unchanged. Operability and

reliability is still maintained by the required inspection. The

adaptation of the proposed surveillance frequency does not involve a

significant reduction in the margins 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: Salem Free Public Library, 112

West Broadway, Salem, New Jersey 08079.

Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston and

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

NRC Project Director: John F. Stolz.

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah

Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee.

Date of amendment request: December 16, 1994 (TS 94-06).

Description of amendment request: The proposed change would revise

the auxiliary feedwater system technical specifications and associated

Bases by incorporating the Westinghouse Standard Technical

Specification limits and format, extending the limiting condition for

operation to Mode 4, relaxing the achievement of hot shutdown from 6

hours to 12 hours, relaxing the verification of valve position

surveillance frequency from 7 days to 31 days, and verifying the

position of automatic valves every 31 days in lieu of valve

manipulation.

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:

TVA has evaluated the proposed technical specification (TS)

change and has determined that it does not represent a significant

hazards consideration based on criteria established in 10 CFR

50.92(c). Operation of Sequoyah Nuclear Plant (SQN) in accordance

with the proposed amendment will not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed TS change replaces SQN's auxiliary feedwater (AFW)

system specification and the associated bases section with improved

requirements that are modeled after the Westinghouse Standard

(NUREG-1431) Technical Specification (STS). The proposed change is

consistent with the STS for ensuring that three trains of AFW remain

operable in Modes 1, 2, and 3. In addition, the proposed change

provides a TS improvement by extending the limiting condition for

operation (LCO) applicability to Mode 4. This LCO requirement for

Mode 4 ensures that at least one motor-driven AFW pump remains

operable when steam generators are being used for decay heat

removal. The proposed 72 hour allowed outage time (for one

inoperable train of AFW) is consistent with the STS and remains

unchanged from SQN's current allowed outage time. One proposed

change to relax shutdown requirements from 6 hours to 12 hours for

achieving hot shutdown is considered to be acceptable. This

relaxation is based on shutdown times contained in the STS and the

operating experience to reach thus condition from full power in an

orderly manner without challenging plant systems. The proposed

surveillance requirements (SRs) provide test frequencies that are

consistent with the STS and are based on operating experience and

the design reliability of the equipment. The proposed relaxation in

surveillance frequency from 7 days to 31 days for verifying valve

position in the AFW flow path is considered acceptable based on

existing procedural controls for valve configuration. The proposed

change to include a STS SR for verifying automatic valves in the

flow path are in their correct position every 31 days (in lieu of

valve manipulation) is considered acceptable based on existing

surveillance that verify proper actuation of SQN's automatic AFW

valves.

The proposed changes provide TS improvements for SQN's AFW

system that ensure the system operates within the bounds of SQN's

AFW accident analysis as contained in the Final Safety Analysis

Report (FSAR). This change does not involve a physical modification

to SQN's AFW system. Accordingly, the proposed changes do not

involve an increase in the probability or consequences of an

accident previously evaluated.

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

from any previously analyzed.

The proposed TS change incorporates requirements that bound the

limiting design-basis accidents (DBAs) evaluated in SQN's FSAR. The

TS bases have been revised to reflect the limiting DBAs and provide

clarification with regard to the assumptions used in SQN's AFW

accident analysis. No new event initiator has been created, not

[sic] has any hardware been changed. This change does not involve a

physical change to SQN's AFW system or any other system. Therefore,

the proposed change will not create the possibility of a new or

different kind of accident from any previously analyzed.

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

TVA's proposed change replaces SQN's AFW system TS requirements

with TS requirements adopted from the Westinghouse STS. Because the

overall similarity in the requirements between SQN's current AFW

specification and the STS version, the TS requirements remain

essentially unchanged. The proposed 72-hour allowed outage time (for

one inoperable train of AFW) is consistent with the STS and remains

unchanged from SQN's current allowed outage time. One proposed

change to relax shutdown requirements from 6 hours to 12 hours for

achieving hot shutdown is considered to be acceptable. This

relaxation is based on shutdown times contained in the STS and the

operating experience to reach this condition from full power in an

orderly manner without challenging plant systems. The proposed SRs

provide test frequencies that are consistent with the STS and are

based on operating experience and the design reliability of the

equipment. The proposed relaxation in surveillance frequency from 7

days to 31 days for verifying valve position in the AFW flow path is

considered acceptable based on existing procedural controls for

valve configuration. The proposed relaxation in surveillance

frequency from 7 days to 31 days for verifying valve position in the

AFW flow path is considered acceptable based on existing procedural

controls for valve configuration. The proposed change to include a

STS SR for verifying automatic valves in the flow path are in their

correct position every 31 days (in lieu of valve manipulation) is

considered acceptable based on other existing surveillances that

verify proper actuation of SQN's automatic AFW valves.

The proposed changes provide TS improvements for SQN's AFW

System that ensure the system operates within the bounds of SQN's

AFW accident analysis as contained in the FSAR. This change does not

[[Page 6310]] involve a physical modification to SQN AFW system.

Accordingly, the margin of safety has not been reduced.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Chattanooga-Hamilton County

Library, 1101 Broad Street, Chattanooga, Tennessee 37402.

Attorney for licensee: General Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902.

NRC Project Director: Frederick J. Hebdon.

The Cleveland Electric Illuminating Company, Centerior Service Company,

Duquesne Light Company, Ohio Edison Company, Pennsylvania Power

Company, Toledo Edison Company, Docket No. 50-440, Perry Nuclear Power

Plant, Unit No. 1, Lake County, Ohio

Date of amendment request: December 16, 1994.

Description of amendment request: The proposed license amendment

would revise Technical Specification 6.3, ``Unit Staff

Qualifications.'' Currently, the Technical Specifications require that

the Operations Manager obtain a senior reactor operator (SRO) license

by August 1995. A change is proposed to relieve the requirement for the

Operations Manager to hold a Perry Nuclear Power Plant (PNPP) SRO

license if an Operations section middle manager holds a PNPP SRO

license.

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 affects an administrative control, which was

based on the guidance of ANSI N18.1-1971, ``Selection and Training

of Nuclear Power Plant Personnel.'' ANSI N18.1-1971 recommended that

the Operations Manager hold a senior reactor operator (SRO) license.

The current guidance in Section 4.2.2 of ANSI/ANS-3.1-1993,

``American National Standard for Selection, Qualification, and

Testing of Personnel for Nuclear Power Plants'' recommends, as one

alternative, that the Operations Manager have plant operational

knowledge consistent with the requirements of the Operations

Manager's position, providing an Operations middle manager holds an

SRO license. This individual (currently designated as the Operations

Superintendent) would be required to meet the criteria for, and

would have responsibilities as recommended in, ANSI/ANS-3.1-1993 for

the Operations Middle Manager position. The proposed change is

consistent with the recommendations of ANSI/ANS-3.1-1993.

The proposed change does not alter the design of any system,

structure or component, nor does it change the way plant systems are

operated. It does not reduce the knowledge, qualifications, or

skills of licensed operators, and does not affect the way the

Operations Section is managed by the Operations Manager. The

Operations Manager will continue to maintain the effective

performance of section personnel and ensure the plant is operated

safely and in accordance with the requirements of the operating

license. Additionally, the control room operators will continue to

be supervised by the licensed senior operators such as the Unit

Supervisors and the Shift Supervisors. For those areas of knowledge

that require an SRO license, the Operations Superintendent will

provide the appropriate technical guidance to the control room

staff.

In summary, the proposed change does not affect the ability of

the Operations Manager to provide the plant oversight required of

the position. Thus, it does not involve a significant increase in

the probability or consequences of an accident previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

The proposed change to Technical Specification 6.3.1 does not

affect the design or function of any plant system, structure, or

component, nor does it change the way plant systems are operated. It

does not affect the performance of NRC licensed operators. Operation

of the plant in conformance with the Technical Specifications and

other license requirements will continue to be supervised by

personnel who hold an NRC SRO license. The proposed change to

Technical Specifications 6.3.1 ensures that either the Operations

Manager or Operations Superintendent will be a knowledgeable and

qualified individual by requiring one of the individuals to hold an

SRO license for PNPP. Based on the above, the proposed change does

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

from any previously evaluated.

3. The proposed change does not result in a significant

reduction in the margin of safety.

The proposed change involves an administrative control which is

not related to the margin of safety as defined in the Technical

Specifications. The proposed change provides an alternative which

ensures that the level of knowledge and experience required of an

individual who fills the Operations Manager position is acceptable.

The proposed change does not affect the conservative manner in which

the plant is operated. The control room operators will continue to

be supervised by personnel who hold an SRO license. Thus, the

proposed change does not involve a significant reduction in a margin

of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Perry Public Library, 3753

Main Street, Perry, Ohio 44081.

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

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

NRC Project Director: Leif J. Norrholm.

The Cleveland Electric Illuminating Company, Centerior Service Company,

Duquesne Light Company, Ohio Edison Company, Pennsylvania Power

Company, Toledo Edison Company, Docket No. 50-440, Perry Nuclear Power

Plant, Unit No. 1, Lake County, Ohio

Date of amendment request: December 21, 1994.

Description of amendment request: The proposed license amendment

would revise Technical Specification 3/4.3.7.7, ``Traversing In-Core

Probe System,'' and its Bases to allow the use of substitute data

generated from the process computer, normalized with available

operating measurements, to replace data from inoperable local power

range monitor (LPRM) strings for up to 10 LPRM strings.

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 TIP [traversing in-core probe] system is not used to prevent

or mitigate the consequences of any previously analyzed accident or

transient. No assumptions are made in any accident analysis relative

to the operation of the TIP system. No other safety related system

is affected by this change.

The use of substitute values from calculations performed by the

on-line computer core monitoring system does not affect the

consequences of plant transients previously evaluated in the USAR

[Updated Safety Analysis Report] because the total core TIP reading

(nodal power) uncertainty remains less than 8.7%. Thus, the MCPR

[minimum critical power ratio] safety limit is not affected.

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

or different kind of [[Page 6311]] accident from any accident

previously evaluated.

The proposed change does not involve the installation of any new

equipment or the modification of any equipment designed to prevent

or mitigate the consequences of accidents or transients. Therefore,

the change has no effect on any accident initiator, and no new or

different type of accidents are postulated to occur.

3. The proposed change does not result in a significant

reduction in the margin of safety.

The total core TIP reading uncertainties will remain within the

assumptions of the licensing basis; thus, the margin of safety to

the MCPR safety limits is not reduced. The ability of the computer

to accurately represent nodal powers in the reactor core is not

compromised. The ability of the computer to accurately predict the

LHGR [linear heat generation rate], APLHGR [average planar linear

heat generation rate], MCPR, and its ability to provide for LPRM

calibration, are not compromised. Therefore, the margin of safety is

not significantly reduced.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Perry Public Library, 3753

Main Street, Perry, Ohio 44081.

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

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

NRC Project Director: Leif J. Norrholm.

TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak Steam

Electric Station, Units 1 and 2, Somervell County, Texas

Date of amendment request: December 6, 1994.

Brief description of amendments: The proposed amendment would

revise Technical Specifications to allow appropriate remedial action

for high particulate levels in the diesel generator fuel oil inventory

and other out-of-limit properties in new diesel generator fuel oil that

has been added to the existing diesel generator fuel oil storage

inventory.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

the probability or consequences of an accident previously evaluated.

The proposed changes allow 7 days to correct particulate

contamination in the stored fuel oil for the diesel generators and

30 days to confirm or restore the adequacy of the stored fuel oil if

certain properties of new fuel that has been added to the fuel oil

storage inventory have been discovered to exceed the specified

values. These changes do not affect plant operations and the only

equipment affected are the diesel generators. The ability of the

diesel generators to provide electrical power when needed is

directly dependent upon, in part, having fuel oil of adequate

quality. The only accident which is potentially initiated by a

diesel generator failure is the station blackout event. The

mitigation of many accidents is dependent upon the availability of

at least one train of electrical power from an emergency diesel

generator (EDG). With the proposed changes, the fuel oil should

continue to have sufficient quality to assure the operability of the

diesel generators until the particulate and other properties are

returned to within limits. This is due in part to the existing fuel

oil quality requirements that are more stringent than the vendor

requires for the EDG to operate and the system of filters installed

to insure good quality fuel actually reaches the EDG. Even though

the margin provided in the quality of the fuel oil may be affected

(see the response to question 3 below), adequate fuel oil quality is

being maintained to assure the operability of the diesel generators

and therefore, these changes do not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

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

or different kind of accident from any accident previously

evaluated.

There are no hardware changes and no changes in system

operations involved. These changes only affect the quality of the

stored fuel oil for the diesel generators. The availability of a

diesel generator has been addressed by the CPSES [Comanche Peak

Steam Electric Station] design and in particular by the analysis of

the station blackout event. These changes do not create the

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

previously evaluated.

3. The proposed changes do not involve a significant reduction

in a margin of safety.

The margin of safety of interest for these changes is the

quality of the stored fuel oil for the diesel generators as compared

to minimum quality which will support the diesel generators ability

to supply electrical power when needed. Particulate contamination

increases slowly over a period of time due to the chemical breakdown

of the fuel oil (or its additives or the surfaces on the tanks

themselves) or due to the introduction of foreign material during

refueling activities. When considered with the fact that the

existing limitation of 10 mg/L was developed for engines which

require much cleaner fuel oil (aircraft engines) and that the CPSES

diesel engines have in line duplex fuel oil filters which can be

switched while the engine is operating, the 7 days which are being

provided to restore the particulate levels do not involve a

significant reduction in the margin of safety. The levels of

particulate are expected to not exceed the specified value by a

significant amount and the specified value is already quite

conservative. Seven days is a reasonable time period in which to

restore the parameter but is short enough to ensure that the

contamination values do not exceed the vendors recommended fuel oil

tolerances required for the EDGs to run. In a similar manner, the

properties of the new fuel oil that has been added to the fuel oil

storage inventory are not expected to deviate significantly from the

allowed values. The testing for gravity, viscosity, flash point,

clarity, water and sediment prior to adding the new fuel oil

provides adequate assurance that the stored fuel oil will be of

sufficient quality to support diesel generator operation. The

quality of the stored fuel oil is further protected from problems

being introduced by new fuel oil that has been added to the fuel oil

storage inventory by the fact that the new fuel oil is generally

diluted by a factor of four or more when it is added to the storage

tanks by the fuel oil that is already in the tanks. Allowing 30 days

to confirm or restore the properties of the stored fuel oil when a

sample of new fuel that has been added to the fuel oil storage

inventory has properties which exceed their specified values 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: University of Texas at

Arlington Library, Government Publications/Maps, 702 College, P.O. Box

19497, Arlington, Texas 76019.

Attorney for licensee: George L. Edgar, Esq., Newman and

Holtzinger, 1615 L Street, N.W., Suite 1000, Washington, D.C. 20036.

NRC Project Director: William D. Beckner.

TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak Steam

Electric Station, Units 1 and 2, Somervell County, Texas

Date of amendment request: December 7, 1994.

Brief description of amendments: The proposed amendment to the

technical specifications (TSs) would: (1) revise the Comanche Peak

Steam Electric Station (CPSES), Technical Specification Limiting

Condition for Operation (LCO) for the main steam isolation valves

(MSIVs) to increase the allowed outage time (AOT) in Mode 1; (2)

relocate the MSIVs full closure time requirement to a program

administratively controlled by the TS; and (3) revise the associated

Bases to [[Page 6312]] adopt the expanded Bases format adding

information specific to CPSES.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Do the proposed changes involve a significant increase in the

probability or consequences of an accident previously evaluated?

The proposed changes are to (1) revise the CPSES Technical

Specification Limiting Condition for Operation (LCO) for the MSIVs

to increase the Allowed Outage Time (AOT) from 4 hours to 8 hours in

Mode 1; (2) modify the Mode 2 and 3 Action Statement to better

reflect the safety significance of these valves by requiring that

the valves be closed within 8 hours and verified at least every 7

days; (3) relocate the MSIVs full closure time requirement to a

program administratively controlled by the TS; and (4) revise the

associated Bases to adopt the expanded Bases format adding

information specific to CPSES.

The revision of the CPSES Technical Specification Limiting

Condition For Operation (LCO) for the MSIVs to increase the Allowed

Outage Time (AOT) from 4 hours to 8 hours in Mode 1 only affects the

time that a condition can exist and as such does not affect any of

the conditions that could initiate an accident; therefore the

probability of an accident is not affected. Likewise, no new

conditions are created that would affect the analyses of any

accident; therefore the consequences of the accidents postulated for

CPSES are not affected.

Modifying the Mode 2 and 3 Action Statement to better reflect

the safety significance of these valves by requiring that the valves

be closed within 8 hours and verified at least every 7 days provides

clarity and adds a new verification requirement. Again no new plant

conditions are established, time limits and verification

requirements are merely being established; therefore, no accident

initiators are affected and there is no impact on the probability of

any accident. Likewise no conditions are being altered which affect

the analyses of any accidents which are postulated at CPSES and thus

the consequences of those accidents are unaffected.

Relocating the MSIVs full closure time requirement to a program

administratively controlled by the TS is an administrative change

only. It has no impact on actual plant operation and thus there is

no impact on the probability of any accident or on the consequences

of any accident.

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

different kind of accident from any accident previously evaluated?

None of the changes in this request affect plant design or create

new operating configurations. The only things affected are the times

that certain conditions are allowed, how soon actions need be

performed, how often to verify conditions and the administrative

location of certain requirements. These items do not create the

possibility of a new type or different kind of accident.

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

margin of safety?

The Technical Specifications LCOs ensure that the assumptions of

the safety analyses are preserved. There are no substantive changes

to the LCO; therefore, the safety analyses are unaffected and there

is no affect on the margin of safety.

Revising the CPSES Technical Specification Limiting Condition

For Operation (LCO) for the MSIVs to increase the Allowed Outage

Time (AOT) from 4 hours to 8 hours in Mode 1 allows the unit to

operate with an inoperable MSIV for a longer period of time.

Although the unavailability of equipment required to mitigate or

assess the consequence of an accident is increased, a more

reasonable completion time is provided to diagnose the problem,

mobilize the corrective action, obtain administrative clearances,

complete the maintenance, restore the valve to an operable

condition, and perform post-maintenance verification, where

appropriate. The additional time would reduce the probability of

unnecessary plant transients and plant shutdowns, thus improving

plant safety and increasing plant availability, while a qualitative

assessment has concluded that the impact on Core Damage Frequency is

negligible. TU Electric has concluded based on the discussion above

that there is no significant impact on the overall margin of safety

due to this change.

Modifying the Mode 2 and 3 Action Statement to better reflect

the safety significance of these valves by requiring that the valves

be closed within 8 hours and verified at least every 7 days is

primarily a clarification and a new verification requirement.

Specifying that an inoperable valve be closed within 8 hours makes

the requirement specific where no time limit was provided before.

The 8 hours specified is the same as is allowed in Mode 1 which was

qualitatively assessed as noted above and thus is a logical

limitation. The new requirement to verify the valves closed on a

periodic basis will increase assurance that the valves remain closed

and will thus enhance the margin of safety. Overall, TU Electric

concludes that these Mode 2 and 3 changes do not significantly

affect the margin of safety.

Relocating the MSIVs full closure time requirement to a program

administratively controlled by the TS is an administrative change

only. There is no impact on the margin of safety.

Revising the associated Bases to adopt the expanded Bases format

adding information specific to CPSES enhances the useability of the

Technical Specification. Overall, this is considered an improvement

which will benefit both the operators and support personnel. There

is no significant impact on the margin of safety and if there is an

impact, it improves the margin by providing easy access to support

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: University of Texas at

Arlington Library, Government Publications/Maps, 702 College, P.O. Box

19497, Arlington, Texas 76019.

Attorney for licensee: George L. Edgar, Esq., Newman and

Holtzinger, 1615 L Street, N.W., Suite 1000, Washington, D.C. 20036.

NRC Project Director: William D. Beckner.

TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak Steam

Electric Station, Units 1 and 2, Somervell County, Texas

Date of amendment request: December 19, 1994.

Brief description of amendments: The proposed changes to the

Technical Specification Action Statements of Tables 3.3-1 and 3.3-2

would allow testing of the reactor protective system (RPS) and the

engineered safety features actuation system (ESFAS) with the channel

under test in bypass.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

(1) Do the proposed changes involve a significant increase in

the probability or consequences of an accident previously evaluated?

The proposed changes will revise those Action Statements which

limit the use of bypass while testing for Reactor Protection System

(RPS) and Engineered Safety Feature Actuation System (ESFAS)

functions. The Actions Statements concern testing with a channel

inoperable and will be revised to allow testing with either the

inoperable channel or the channel being tested (but not both) placed

in bypass.

Testing in a bypass condition when all channels are operable

will not introduce new operating configurations. The number [of]

available channels with one channel in bypass for testing will

remain the same as the minimum number of channels and is the same as

the number of channels available when testing in trip. The number of

channels to trip will be unchanged when testing in bypass while the

number of channels to trip is reduced to one when testing in trip.

Although there may be a sight [slight] increase in possibility that

the failure of a channel could prevent the actuation of a function

(because testing in bypass could result in two-out-of-two logic

while testing in trip would have resulted in one-out-of-two logic),

testing in bypass will reduce the vulnerability to inadvertent

actuation of a function while maintaining the normal channels to

trip and the minimum channels [[Page 6313]] operable requirements

per the current technical specifications. Overall TU Electric

concludes (and WCAP-10271 with its associate SER from the NRC

supports) that testing in bypass when all channel [s] are operable

does not involve a significant increase in the probability or

consequences of an accident previously evaluated.

Testing in bypass with one channel inoperable will not introduce

new configurations. The current Actions Statements for ESFAS already

allow testing in bypass if one channel is inoperable. Under the

current Technical Specifications for an RPS function, an inoperable

channel is placed in bypass (via leads and jumpers) while

surveillance testing another channel (the channel under test is

placed in trip). Under the proposed changes, either the inoperable

channel or the channel being tested may be bypassed.

In either case, the result is one channel in bypass and the

other in trip, which leaves one-out-of-two operable channels to

initiate the protective function (if the initial logic was two-out-

of-four) or one-out-of-one operable channels to initiate the

protective function (if the initial logic was two-out-of-three).

Thus, testing in bypass with one channel inoperable does not involve

a significant increase in the probability or consequences of an

accident previously evaluated.

The proposed technical specification changes will also allow

certain ESFAS functions to be tested with an inoperable channel in

bypass and the channel being tested in trip. The current technical

specifications require that the inoperable channel be in trip and

that the channel being tested be in bypass. Per the same logic

provided above on testing in bypass with an inoperable channel, this

change has no impact on the capability of the system to respond to

plant conditions and does increase the potential for inadvertent

actuation of a function.

In summary, the proposed changes to the technical specifications

and testing in bypass do not increase the probability or

consequences of an accident previously evaluated.

(2) Do the proposed changes create the possibility of a new or

different type of accident from any accident previously evaluated?

No new operating configurations and no new failure modes are

being introduced by testing in bypass or by the proposed technical

specification changes; therefore, no new or different type of

accident from any accident previously evaluated is being created.

(3) Do the proposed changes involve a significant reduction in

the margin of safety?

Testing in bypass does not affect accident configurations,

sequences, or response scenarios as modeled in the safety analyses.

Testing or maintenance in a bypass configuration does not cause any

design or analysis acceptance criteria to be exceeded, nor does it

affect the integrity of the fission product barriers. The severity

of any accident previously evaluated is not increased. Bypass

testing does not affect the functional integrity of the Reactor

Protection System (RPS) or the Engineered Safety Features Actuation

System (ESFAS). Bypass testing and the proposed technical

specification 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

amendment request involves no significant hazards consideration.

Local Public Document Room location: University of Texas at

Arlington Library, Government Publications/Maps, 702 College, P.O. Box

19497, Arlington, Texas 76019.

Attorney for licensee: George L. Edgar, Esq., Newman and

Holtzinger, 1615 L Street, N.W., Suite 1000, Washington, D.C. 20036.

NRC Project Director: William D. Beckner.

TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak Steam

Electric Station, Units 1 and 2, Somervell County, Texas

Date of amendment request: December 30, 1994

Brief description of amendments: The proposed amendments would

revise the technical specification for fuel storage to authorize use of

the high density fuel storage racks, to increase the spent fuel storage

capacity, and to adopt the wording, content, and format of the Improved

Standard Technical Specifications.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Do the proposed changes involve a significant increase in the

probability or consequence of an accident previously evaluated?

This proposed license amendment includes changes which clarify

the Technical Specifications, identify existing licensing basis

criteria, revise the wording and format to be consistent with the

Improved Standard Technical Specifications (NUREG-1431), and provide

the criteria for acceptable fuel storage in high density racks. The

clarification and the revised wording and format are purely

administrative changes and have no impact on the probability or

consequences of an accident. The criteria for acceptable fuel

storage in the high density racks are discussed below.

The high density racks differ from the low density racks in that

the center to center storage cell spacing is decreased from a

nominal 16 inches to a nominal 9 inches and the high density racks

are free standing whereas the low density racks are bolted to the

pool. The allowed storage pattern in the high density racks results

in a nominal 12.7 inch center to center spacing (measured

diagonally) with a two out of four storage pattern (high density (2/

4)). Administrative controls are used to maintain the specified

storage patterns and to assure storage of a fuel assembly in a

proper location based on initial U-235 enrichment and burnup. The

increased storage capacity results in added weight in the pools and

additional heat loads.

The only potential impact on the probability of an accident

concerns the potential insertion of a fuel assembly in an incorrect

location in the high density racks. TU Electric has used

administrative controls to move fuel assemblies from location to

location since the initial receipt of fuel on site. Through receipt

of fuel for two initial core loads and four refueling outages (each

of which includes a complete core offload), TU Electric has not

inserted a fuel assembly into an improper location. This record

demonstrates the adequacy of the administrative controls in place

and confirms that the use of such administrative controls will not

involve a significant increase in the probability of an accident

previously evaluated.

The consequences of all of these changes have been assessed and

the current acceptance criteria in the licensing basis of CPSES will

continue to be met. The nuclear criticality, thermal-hydraulic,

mechanical, material and structural designs will accommodate these

changes. Potentially affected analyses, including a dropped spent

fuel assembly, a loss of spent fuel pool cooling, a seismic event,

and a fuel assembly placed in a location other than a prescribed

location, continue to satisfy the CPSES licensing basis acceptance

criteria. The analysis methods used by TU Electric are consistent

with methods used by TU Electric in the past or methods used

elsewhere in the industry and accepted by the NRC.

Based on the acceptability of the methodology used and

compliance with the current CPSES licensing basis, TU Electric

concludes that the use of the high density racks and the increase in

storage capacity do not involve a significant increase in the

probability or consequences of an accident previously evaluated.

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

different kind of accident from any accident previously evaluated?

The administrative changes to the Technical Specifications have

no impact on plant hardware or operations and therefore cannot

create a new or different kind of an accident.

The spacing changes between fuel assemblies, the administrative

controls, the storage limitations, and the increased storage

capacity do not generate new failure modes that could create a new

or different kind of an accident. The change from bolted low density

racks to free standing high density racks will not create the

possibility of a new or different kind of an accident. Free standing

racks have been commonly used at nuclear power plants to provide for

high density storage of spent fuel, and their use [[Page 6314]] does

not entail any unproven or unusual design or technology. In this

regard, a number of plants have previously changed from bolted or

restrained racks to free standing racks, including Millstone 1

(amendment dated November 27, 1989) and San Onofre 2 and 3

(amendment dated May 1, 1990), and such changes have not been

classified as involving a significant hazards consideration.

Furthermore, CPSES is not located in an area subject to severe

seismic events. A seismic event at CPSES would result in little

movement of the free standing racks and would not cause the high

density racks to collide with each other or the spent fuel pool

walls. Therefore, use of the free standing high density racks would

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

accident.

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

margin of safety?

The proposed administrative changes to the Technical

Specifications have no impact on any acceptance criteria, plant

operations or the actual failure of any systems, components or

structure; therefore these administrative changes have no impact on

the margin of safety.

The NRC guidance [Nuclear Regulatory Commission, Letter to all

Power Reactor Licensees, from B. K. Grimes, April 14, 1978, ``OT

Position for Review and Acceptance of Spent Fuel Storage and

Handling Applications,'' as amended by the NRC Letter dated January

18, 1979] has established that an evaluation of margin of safety

should address the following areas:

(1) Nuclear criticality considerations.

(2) Thermal-Hydraulic considerations.

(3) Mechanical, material and structural consideration.

The established acceptance criterion for criticality is that the

neutron multiplication factor in the spent fuel pool storage racks

shall be less than or equal to 0.95, including uncertainties, under

all conditions. The keff for the high density racks for CPSES

is always less than 0.95, including uncertainties at a 95/95

probability confidence level. Because the existing acceptance

criterion is shown to be satisfied, the high density racks do not

involve a significant reduction in the margin of safety with respect

to criticality considerations.

The thermal-hydraulic evaluation demonstrates that the

temperature margin of safety will be maintained. Re-evaluation of

the spent fuel pool cooling system for the increased heat loads

shows, with minor modifications, that the spent fuel cooling system

will maintain the abnormal maximum temperature of the spent fuel

pool water within the limits of the existing licensing basis (i.e.,

below 212 deg.F). Additionally, it shows that, with minor

modifications, the normal maximum temperature will be within the

existing design basis temperatures for the high density racks,

liner, structure, and cooling system and will not have any

significant impact on the spent fuel pool demineralizers. Thus, the

existing licensing basis remains valid, and there is no significant

reduction in the margin of safety for the thermal-hydraulic design

or spent fuel cooling.

The main safety function of the spent fuel pool and the high

density racks is to maintain the spent fuel assemblies in a safe

configuration through normal and abnormal operating conditions. The

design basis floor responses of the Fuel Building were confirmed to

be adequate and conservative and the floor loading will not exceed

the capacity of the Fuel Building. The high density rack materials

used are compatible with the spent fuel pool and the spent fuel

assemblies. The structural considerations of the high density racks

maintain margin of safety against tilting and deflection or

movement, such that the high density racks do not impact each other

or the pool walls, damage spent fuel assemblies, or cause

criticality concerns. Thus, the margin of safety with respect to

mechanical, material and structural considerations are not

significantly reduced by the use of the high density racks.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: University of Texas at

Arlington Library, Government Publications/Maps, 702 College, P.O. Box

19497, Arlington, Texas 76019.

Attorney for licensee: George L. Edgar, Esq., Newman and

Holtzinger, 1615 L Street, N.W., Suite 1000, Washington, DC 20036.

NRC Project Director: William D. Beckner.

Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,

Callaway County, Missouri

Date of amendment request: December 9, 1994.

Description of amendment request: The proposed amendment would

revise Technical Specification (TS) 4.3.2.2, 4.7.1.2.1, and the Bases

for Specification 3/4.7.1.2. The changes would decrease the frequency

of testing auxiliary feedwater pumps, provide consistent testing

requirements for the steam turbine-driven auxiliary feedwater pump, and

clarify performance parameters in the Bases.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

The proposed revision does not involve a significant hazards

consideration because operation of Callaway Plant with this change

would not:

(1) Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The Callaway Final Safety Analysis Report has been reviewed and

been found to be unaffected by these proposed changes. The changes

proposed by this Technical Specification amendment do not affect the

performance parameters of the Auxiliary Feedwater System (AFWS). The

changes proposed involve a decrease in the frequency of pump testing

from once per 31 days to once per 92 days as recommended by NRC

Generic Letter 93-05 and reflected in NUREG-1431 (T/S 4.7.1.2.1.a).

This change will decrease the out-of-service time of the AFWS due to

testing. This change will also decrease the number of component

manipulations performed on the system and will therefore decrease

the probability of a restoration error rendering the system

incapable of performing its intended function.

The pumps will be required to meet the same acceptance criteria

and will continue to be monitored as required by ASME Section XI. As

stated earlier, the overall effect is a slight decrease in the CDF

for Callaway. These proposed changes will also eliminate an

inconsistency among Specifications 4.7.1.2.1.b.2 and 4.3.2.2 and

Specification 4.7.1.2.1.a.2 regarding an exception to Specification

4.0.4 for entry into Mode 3 for the TDAFP. The methodology and

acceptance criteria of surveillance testing will not be changed. The

ability of the AFWS to perform its intended function during accident

conditions will continue to be demonstrated via surveillance

testing. The proposed changes to the Technical Specifications do not

affect any accident initiators for any accident evaluated in the

Final Safety Analysis Report (FSAR). The Bases changes are

corrections to errors which have no effect on any accident

initiators nor equipment failure modes.

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

accident from any previously evaluated.

The proposed Technical Specification changes do not modify any

equipment nor create any potential accident initiators. The proposed

change herein of potential interest is the exception to

Specification 4.0.4 for entry into Mode 3 for TDAFP response time

testing and auto-start testing. This allowance is already recognized

via Specification 4.7.1.2.1.a.2 and NUREG-1431, Standard Technical

Specifications-Westinghouse Plants.

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

The Bases for Specification 3/4.7.1.2 are to be clarified to

correctly state the design flow and pressure parameters for the

AFWS. No plant design changes are involved in any of the proposed

changes and the method and manner of plant operation remain the

same. The specific surveillance test methodology and acceptance

criteria remain unchanged.

As discussed above, the proposed change does not involve a

significant increase in the probability or consequences of an

accident previously evaluated or create the possibility of a new or

different kind of accident from any previously evaluated. These

changes do not result in a significant reduction in a margin of

safety. Therefore, it has been determined that the proposed changes

do not involve a significant hazards consideration.

[[Page 6315]] 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: Callaway County Public

Library, 710 Court Street, Fulton, Missouri 65251.

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

& Trowbridge, 2300 N Street, N.W., Washington, DC 20037.

NRC Project Director: Leif J. Norrholm.

Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,

Callaway County, Missouri

Date of amendment request: December 9, 1994, as supplemented on

December 22, 1994.

Description of amendment request: The proposed amendment would

revise Technical Specification (TS) Surveillance Requirement

4.8.1.1.2f.7 to remove the requirement to perform the hot restart test

within 5 minutes of completing the 24-hour endurance test and place

that requirement in a separate 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:

The proposed revision does not involve a significant hazards

consideration because operation of Callaway Plant with this change

would not:

(1) Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed revision to the T/S will not adversely impact plant

safety since the requirement to perform the hot restart test will

still be implemented via a separate surveillance requirement that

demonstrates the hot restart functional capability of the diesel

generators.

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

accident from any previously evaluated.

There are no design changes being made that would create a new

type of accident or malfunction and the method and manner of plant

operation remain unchanged. The performance capability of the

emergency diesel generators will not be affected. The verification

of the hot restart capability of the diesel generators will still be

performed, only the timing of the performance will be changed to

give plant operators added flexibility and prevent critical path

complications during outages.

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

There are no changes being made to the safety limits or safety

system settings that would adversely impact plant safety. The diesel

generators will still perform their intended safety function

following a loss of offsite power, to achieve and maintain the plant

in a safe shutdown condition.

Based on the above discussions, it has been determined that the

requested Technical Specification change does not involve a

significant increase in the probability or consequences of an

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

accident or condition over previous evaluations; or involve a

significant reduction in a margin of safety. Therefore, the

requested license amendment does not involve a significant hazards

consideration.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Callaway County Public

Library, 710 Court Street, Fulton, Missouri 65251.

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

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

NRC Project Director: Leif J. Norrholm.

Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, Vermont

Yankee Nuclear Power Station, Vernon, Vermont

Date of amendment request: December 14, 1994.

Description of amendment request: The proposed amendment would

revise instrument identification for low reactor pressure instrument

trip cards in emergency core cooling system (ECCS) actuation to reflect

a design change to be installed during the 1995 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 proposed change to the identification numbers for certain

reactor pressure instrumentation as included in the Technical

Specifications for ECCS Actuation Instrumentation is only necessary

because the specific identification numbers (Tag Nos.) have been

listed in the [***]. This is considered an administrative type

change. Acceptable measurement of Low Reactor Pressure is still

assured. All automatic control or trip functions will continue to be

provided.

The proposed change does not result in any function or setpoint

change. The hardware changes which have resulted in a need to change

the Technical Specifications have removed instrumentation no longer

required to be installed in the circuitry for measuring ECCS Low

Reactor Pressure. The existing logic for Low Reactor Pressure will

remain the same. The only change applicable to implementation of the

design modification is the use of different trip cards to provide

the trip function for ECCS Low Reactor Pressure.

The requested change to ECCS Actuation Instrumentation Tables

does not impact any FSAR [Final Safety Analysis Report] safety

analysis involving the ECCS or Protection Systems. These measurement

functions are not contributors to the initiation of accidents.

The change in instrument Tag Nos. on Tables 3.2.1 and 4.2.1 will

have no affect on any safety limit setting or plant system operation

and, therefore, does not modify or add any initiating parameters

that would significantly increase the probability or consequences of

any previously analyzed accident.

The administrative change to correct a typographical error on

Table 4.2.1 will have no affect on plant hardware, plant design,

safety limit setting or plant system operation and, therefore, does

not modify or add any initiating parameters that would significantly

increase the probability or consequences of any previously analyzed

accident.

Therefore, it is concluded that there is not a significant

increase in the probability or consequence of an accident previously

evaluated.

2. The proposal to change instrument Tag Nos. does not result in

any function changes or changes to Technical Specification

requirements pertaining to these functions.

The proposed change does not involve any change in Technical

Specification trip setpoints, plant operation, redundancy,

protective function or design basis of the plant. There is no impact

on any existing safety analysis or safety design limits. Low Reactor

Pressure instrumentation functions do not initiate nuclear system

parameter variations which are considered potential initiating

causes of threats to the fuel and the nuclear system process barrier

or that would create any new or different kind of accident.

As discussed above, the proposed administrative change only

corrects a typographical error concerning equipment identification

numbers. This change does not affect any equipment and it does not

involve any potential initiating events that would create any new or

different kind of accident.

Therefore, the proposed changes do not create the possibility of

a new or different kind of accident from any accident previously

evaluated.

3. The proposal to change the identification numbers for certain

reactor pressure instrumentation as included in the Technical

Specifications for ECCS Actuation Instrumentation does not affect

any existing safety margins. The change by itself is administrative.

The hardware changes which have resulted in a need to change the

Technical Specifications have been reviewed per 10 CFR 50.59(a)(2)

and determined to not constitute an unreviewed safety question.

The change in Tag Nos. or the change in the instrumentation used

to measure low [[Page 6316]] reactor pressure does not preclude the

ability of the Core Spray (CS) or Low Pressure Coolant Injection

(LPCI) Systems to perform their safety function to mitigate the

consequences of accidents or of any other safety system to

accomplish its safety functions. Proper post-accident ECCS

functioning will still be provided by safety class instruments used

to measure reactor pressure.

The change to instrument Tag Nos. as listed in the Technical

Specifications has no affect on the bases of Protective

Instrumentation which is to operate to initiate required system

protective actions. The changes to be implemented which have

resulted in a need to change the Technical Specifications will

actually improve the accuracy of reactor pressure measuring loops.

[***]

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

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